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Sample records for superphenix fuel load

  1. The first Superphenix fuel load reliability analysis and validation

    International Nuclear Information System (INIS)

    Marbach, G.; Beche, M.; Pajot, J.

    1986-09-01

    The excellent behavior of PHENIX driver fuel and the burnup values currently reached suggest that the first SUPERPHENIX fuel load will meet the design lifetime. However, to ensure the reliability of the entire load, all the parameters affecting fuel behavior in reactor must be analyzed. For that purpose, we have taken into account all the results of the examination and verifications during the fabrication process of the first load subassemblies. These data concern geometrical parameters or oxide composition as well as the cladding tube and plug weld soundness tests. The objective is to determine the actual dispersion of all the parameters to ensure the absence of failure due to fabrication defects with very high statistical confidence limits. The influence of all the parameters has been investigated for the situations which can occur during power-up, steady-state operation and transients. The fabrication quality allows us to demonstrate that in all cases good behavior criteria for fuel and structure will be maintained. This demonstration is based on calculation code results as well as on validation by specific experiments

  2. Design and fabrication procedures of Super-Phenix fuel elements

    International Nuclear Information System (INIS)

    Leclere, J.; Vialard, J.-L.; Delpeyroux, P.

    1975-01-01

    For Super-Phenix fuel assemblies, Phenix technological arrangements will be used again, but they will be simplified as far as possible. The maximum fuel can temperature has been lowered in order to obtain a good behavior of hexagonal tubes and cans at high irradiation levels. An important experimental programme and the experience gained from Phenix operation will confirm the merits of the options retained. The fuel element fabrication is envisaged to take place in the plutonium workshop at Cadarache. Usual procedures will be employed and both reliability and automation will be increased [fr

  3. Sodium leak on the fuel storage drum of Superphenix

    International Nuclear Information System (INIS)

    Acket, C.; Marcon, J.P.; Michoux, H.

    1988-01-01

    SUPERPHENIX the world's largest fast breeder prototype reached its nominal power 1200 MWe in December 1986. In March 1987 a sodium leakage was detected on the 'barillet'. This is a large double walled cylindrical sodium tank (14 m high, 9 m in diameter) made of ferritic steel and filled with 700 tonnes of sodium at a temperature of 200 0 C. Located close to the primary pool it is used in the refuelling process of the plant. The leakage of sodium through the main vessel was confined in the guard vessel. This paper presents the different stages of the operations undertaken: to guarantee and improve the safety until the complete drainage of sodium; to drain the vessels and localize the leakage; to characterise the defect and the presence or not of other similar or different defects; to define the next step between several solutions including the local repair and complete reconstruction. (author)

  4. Superphenix: technical and scientific achievements

    International Nuclear Information System (INIS)

    Guidez, Joel; Prele, Gerard

    2016-04-01

    In this book, the authors propose a synthesis of technical and scientific achievements related to the design, fabrication and eleven-year operation of Superphenix, the most powerful fast breeder reactor ever built and operated. They had the opportunity to use various and important archives maintained by the different involved institutions, actors and companies, such as the CEA with its MADONA database, AREVA and EDF. They address all the different fields: construction, chemistry, exploitation, handling, small and large components, materials, fuel manufacturing, environmental assessment, thermal hydraulics, the sodium-water reaction, sodium fires, the release of residual power, in-service inspection, and dismantling operations. Moreover, a chapter addresses design studies for Superphenix 2 and for the European Fast Reactor (EFR) which should be the successors of Superphenix

  5. Manufacture of the first fuel charge for the SUPER-PHENIX 1 reactor

    International Nuclear Information System (INIS)

    Pajot, J.; Beche, M.; Heyraud, J.

    1988-01-01

    After summarizing same general points on the Super Phenix core, the performances of fuel essemblies, the remainder of this discussion will deal with the manufacture by the CFCa of the first charge of fuel assemblies. The following aspects are considered in sequence - contract - production facilities - manufacturing procedures finally a few assessments will be presented

  6. Fuel pellet loading apparatus

    International Nuclear Information System (INIS)

    1980-01-01

    Apparatus is described for loading a predetermined amount of nuclear fuel pellets into nuclear fuel elements and particularly for the automatic loading of fuel pellets from within a sealed compartment. (author)

  7. The Superphenix dismantling

    International Nuclear Information System (INIS)

    Carle, R.

    1999-01-01

    This document presents selected abstracts of Remy Carle's presentation on the dismantling of Superphenix (october 1998). The author wonders about the consequences of such a decision. After a chronological account of this fast reactor project, its cost and the scientific and technical contribution, the dismantling problem is considered. For EDF (Electricite De France) the dismantling dimension is considered at the same time of the design. The main problem is the liquid sodium reprocessing: a technical but also a financing problem. The end of the speech deals with the political aspects of Superphenix and the relations with the public. (A.L.B.)

  8. No rebirth for Superphenix?

    International Nuclear Information System (INIS)

    Anon.

    1992-01-01

    Superphenix had its wings clipped. On June 29, French Prime Minister Peirre Beregovoy announced that the country's commercial fast breeder reactor project Superphenix (SPX-1) will be halted indefinitely. Beregovoy based his decision on safety concerns raised by French nuclear regulators for improved fire protection. With SPX-1's restart being nixed, other fast breeder programs in Europe may not fly at all. This may even ground France's reprocessing and recycling programs as well as nuclear back-end projects in Germany and Japan. Has the Superphenix fallen back into its ashes by the push of public opposition? Anti-nuclear voices heard throughout Germany and France speak the same language despite differences in government, licensing procedures, and nuclear control policies. They're calling for an end to nuclear power and the politicians are taking notice. Will the Superphenix meet a fate similar to the German prototype fast breeder project SNR-300 among others? The following article examines the political and public perceptions of the nuclear power industry in Germany and France; the reasons why France put the hot SPX-1 on ice; the effects of German nuclear policy upon France; and the future of breeder technology in Europe

  9. Fuel element loading system

    International Nuclear Information System (INIS)

    Arya, S.P; s.

    1978-01-01

    A nuclear fuel element loading system is described which conveys a plurality of fuel rods to longitudinal passages in fuel elements. Conveyor means successively position the fuel rods above the longitudinal passages in axial alignment therewith and adapter means guide the fuel rods from the conveyor means into the longitudinal passages. The fuel elements are vibrated to cause the fuel rods to fall into the longitudinal passages through the adapter means

  10. Safety issues for LMFBR: important features drawn from the assessments of Superphenix

    International Nuclear Information System (INIS)

    Natta, M.

    2002-01-01

    Superphenix, which is built on the site of Creys-Malville, is still the biggest LMFBR plant that has been in operation. It is a pool type reactor, as Phenix and the RNR 1 500 and EFR projects. After the analysis of the preliminary safety (1974-1975), the construction was authorised by decree of the Prime Minister in 1977, the authorization for fuel loading and star-up to 3% was given by the minister of industry in July 1985 and full power was achieved in December 1986. The plant was operated until the end of December 1996, producing the equivalent of 320 EFPD, corresponding to half of the maximum barn-up of the first core. The plant was definitively stopped on the 20. of April 1998 by a decision of the French government. During this period of 25 years of licensing, construction and operation of Superphenix, others discussions and preliminary licensing procedures were started for new projects, mainly the RNR 1500 French project and the EFR European project. The operation of Superphenix was also marked by several incidents, which led to additional licensing procedures and important modifications. This period was also marked by an important work of research and development in the safety field, mostly related to the issues concerning hypothetical core disruptive accidents (HCDA) and sodium fires; further, this period was marked by the Three Mile Island accident in 1979 and the Chernobyl accident in 1986. The purpose of this paper is to present some items which were discussed during this period of 25 years and which should be of interest for future LMFBRs. In this presentation, we shall discuss the key issues concerning the safety criteria and options taken with respect to severe accidents, i.e. core melt accidents, giving details on some specific which are less known since they were assessed only lately for Superphenix, sometimes in connection with the on-going safety researches. (author)

  11. Fuel Load (FL)

    Science.gov (United States)

    Duncan C. Lutes; Robert E. Keane

    2006-01-01

    The Fuel Load method (FL) is used to sample dead and down woody debris, determine depth of the duff/ litter profile, estimate the proportion of litter in the profile, and estimate total vegetative cover and dead vegetative cover. Down woody debris (DWD) is sampled using the planar intercept technique based on the methodology developed by Brown (1974). Pieces of dead...

  12. Nuclear fuel pellet loading apparatus

    International Nuclear Information System (INIS)

    Gerkey, K.S.

    1979-01-01

    An automatic apparatus for loading a predetermined amount of nuclear fuel pellets into a nuclear fuel element to be used in a nuclear reactor is described. The apparatus consists of a vibratory bed capable of supporting corrugated trays containing rows of nuclear fuel pellets and arranged in alignment with the open ends of several nuclear fuel elements. A sweep mechanism is arranged above the trays and serves to sweep the rows of fuel pellets onto the vibratory bed and into the fuel element. A length detecting system, in conjunction with a pellet stopping mechanism, is also provided to assure that a predetermined amount of nuclear fuel pellets are loaded into each fuel element

  13. French military plans for Superphenix

    International Nuclear Information System (INIS)

    Albright, D.

    1984-01-01

    France refuses to rule out military use of the plutonium produced by the planned breeder reactor Superphenix, although other nations, including the US, have contributed nuclear materials to it. US policy has been to separate military and civilian nuclear programs to set an example. France has not stated an intention to use Superphenix for military purposes, but is reserving the right to do so. It does not separate the two kinds of nuclear materials for economic reasons. The Non-Proliferation Treaty (NPT) does not address the possibility that plutonium pledged to peaceful use might be commingled with plutonium for military use in a civilian facility within a weapons state. The US could work to strengthen the US-Euratom Agreement on the basis of the contamination principle. 11 references

  14. Superphenix set to rise again

    International Nuclear Information System (INIS)

    Dorozynski, A.

    1993-01-01

    Superphenix, France's seemingly jinxed fast breeder reactor, which has not produced a single kilowatt of energy in more than 3 years, looks set to rise up next year like the mythical bird it is named after. The $5 billion reactor, the largest fast breeder in the world, has just been given the seal of approval by a public commission ordered by the government to look at the pros and cons of restarting. It still has hoops to jump through: a safety check and approval from the ministries of industries and environment. But the consortium of French, Italian, and German power utilities that run the plant are confident they can get it running by next summer. The Superphenix that rises out of the ashes will, however, be a different species of bird from the one planned 20 years ago. The consortium plans to turn the reactor into a debreeder, one that will incinerate more plutonium than it produces and so eat into Europe's plutonium stockpile. Calculations by Superphenix staff and the Atomic Energy Commission indicate that a plutonivorous fast breeder could incinerate 15 to 25 kilograms of plutonium while producing 1 billion kilowatt-hours of electricity-scarcely enough to make a dent in the tonnes of plutonium produced by Electricite de France's reactors each year. The Superphenix consortium is anxious to get the reactor back on line. The annual cost of upkeep and repair of the idle plant and salaries for its 700 staff may reach $140 million this year, 20% more than if the plant was running normally. If restarted, the existing core and a second one ready on the shelf will generate electricity worth $1.3 billion

  15. Nuclear fuel rod loading apparatus

    International Nuclear Information System (INIS)

    King, H.B.

    1981-01-01

    A nuclear fuel loading apparatus, incorporating a microprocessor control unit, is described which automatically loads nuclear fuel pellets into dual fuel rods with a minimum of manual involvement and in a manner and sequence to ensure quality control and accuracy. (U.K.)

  16. Neutron characteristics of the Super-Phenix 1 reactor at Creys-Malville

    International Nuclear Information System (INIS)

    Giacometti, C.; Bouget, Y.H.; Hammer, P.; Lyon, F.; Salvatores, M.; Sicard, B.; Pipaud, J.Y.

    1980-01-01

    The paper describes the method used to determine the critical enrichments for the first loading of the Super-Phenix reactor and the correction factors (together with their uncertainties) applied to the data calculated from the CARNAVAL IV code. These enrichments must be chosen so as to conform to the planned operating conditions of the reactor: nominal power of the pressure vessels, lifetime of the in-pile assemblies. Allowance for uncertainties of neutronic origin and those associated with the fabrication of the fuel pins calls for an over-enrichment of the first loading by approximately 4 per cent. An analysis is made of the effects of this over-enrichment on the core characteristics, which have to remain compatible with the established limits. (author)

  17. Pattern fuel assembly loading system

    International Nuclear Information System (INIS)

    Ahmed, H.J.; Gerkey, K.S.; Miller, T.W.; Wylie, M.E.

    1986-01-01

    This patent describes an interactive system for facilitating preloading of fuel rods into magazines, which comprises: an operator work station adapted for positioning between a supply of fuel rods of predetermined types, and the magazine defining grid locations for a predetermined fuel assembly; display means associated with the work station; scanner means associated with the work station and adapted for reading predetermined information accompanying the fuel rods; a rectangular frame adapted for attachment to one end of the fuel assembly loading magazine; prompter/detector means associated with the frame for detecting insertion of a fuel rod into the magazine; and processing means responsive to the scanner means and the sensing means for prompting the operator via the display means to pre-load the fuel rods into desired grid locations in the magazine. An apparatus is described for facilitating pre-loading of fuel rods in predetermined grid locations of a fuel assembly loading magazine, comprising: a rectangular frame adapted for attachment to one end of the fuel assembly loading magazine; and means associated with the frame for detecting insertion of fuel rods into the magazine

  18. Nuclear fuel rod loading apparatus

    International Nuclear Information System (INIS)

    King, H.B.; Macivergan, R.; Mckenzie, G.W.

    1980-01-01

    An apparatus incorporating a microprocessor control is provided for automatically loading nuclear fuel pellets into fuel rods commonly used in nuclear reactor cores. The apparatus comprises a split ''v'' trough for assembling segments of fuel pellets in rows and a shuttle to receive the fuel pellets from the split ''v'' trough when the two sides of the split ''v'' trough are opened. The pellets are weighed while in the shuttle, and the shuttle then moves the pellets into alignment with a fuel rod. A guide bushing is provided to assist the transfer of the pellets into the fuel rod. A rod carousel which holds a plurality of fuel rods presents the proper rod to the guide bushing at the appropriate stage in the loading sequence. The bushing advances to engage the fuel rod, and the shuttle advances to engage the guide bushing. The pellets are then loaded into the fuel rod by a motor operated push rod. The guide bushing includes a photocell utilized in conjunction with the push rod to measure the length of the row of fuel pellets inserted in the fuel rod

  19. Fuel rod pellet loading head

    International Nuclear Information System (INIS)

    Howell, T.E.

    1975-01-01

    An assembly for loading nuclear fuel pellets into a fuel rod comprising a loading head for feeding pellets into the open end of the rod is described. The pellets rest in a perforated substantially V-shaped seat through which air may be drawn for removal of chips and dust. The rod is held in place in an adjustable notched locator which permits alignment with the pellets

  20. Automated fuel pin loading system

    Science.gov (United States)

    Christiansen, D.W.; Brown, W.F.; Steffen, J.M.

    An automated loading system for nuclear reactor fuel elements utilizes a gravity feed conveyor which permits individual fuel pins to roll along a constrained path perpendicular to their respective lengths. The individual lengths of fuel cladding are directed onto movable transports, where they are aligned coaxially with the axes of associated handling equipment at appropriate production stations. Each fuel pin can be be reciprocated axially and/or rotated about its axis as required during handling steps. The fuel pins are inerted as a batch prior to welding of end caps by one of two disclosed welding systems.

  1. Nuclear fuel pellet loading machine

    International Nuclear Information System (INIS)

    Dazen, J.R.; Denero, J.V.

    1976-01-01

    A nuclear fuel pellet loading machine is described including an inclined rack mounted on a base and having parallel spaced grooves on its upper surface arranged to support fuel rods. A fuel pellet tray is adapted to be placed on a table spaced from the rack, the tray having columns of fuel pellets which are in alignment with the open ends of fuel rods located in the rack grooves. A transition plate is mounted between the fuel rod rack and the fuel pellet tray to receive and guide the pellets into the open ends of the fuel rods. The pellets are pushed into the fuel rods by a number of mechanical fingers mounted on a motor operated block which is moved along the pellet tray length by a drive screw driven by the motor. To facilitate movement of the pellets in the fuel rods the rack is mounted on a number of spaced vibrators which vibrate the fuel rods during fuel pellet insertion. A pellet sensing device movable into an end of each fuel rod indicates to an operator when each rod has been charged with the correct number of pellets

  2. 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.

  3. Superphenix 1 intermediate heat exchanger fabrication

    International Nuclear Information System (INIS)

    Noel, H.; Granito, F.; Pouderoux, P.

    1985-01-01

    The eight Superphenix 375-MW (thermal) intermediate heat exchangers (IHXs) are similar in overall design to the Phenix components. Detailed design changes had to be made during fabrication on the following grounds: Due to seismic resistance, the support area was raised as high as possible to situate the component natural frequencies well out of the resonance peak range and remove thick plate-to-shell connections from heavy thermal load areas. Integration of lessons drawn from the Phenix incidents, due mainly to secondary sodium radial temperature disparities, resulted in the design of a more adaptable outlet header, together with a sodium mixing device, and in the reduction of temperature differences by heat insulation. To avoid circumferential temperature disparities, the iron shot biological shielding plug was replaced by stacked stainless steel plates within an outer shell, which in the new design, is not a supporting structure. The thermal-hydraulic and mechanical design of the component necessitated the elaboration of sophisticated computer codes, with validation of results on mock-ups. The detailed design studies and the actual manufacturing work had to adapt to both design developments and to inherent fabrication difficulties, mainly related to the very tight tolerances imposed for these exceptionally large components and to the welding of steel with an excessive boron content. The construction of the Creys-Malville IHXs afforded valuable industrial experience, which should provide a basis for the design of simpler and less costly IHX units for the forthcoming 1500-MW (electric) breeder

  4. Nuclear fuel pellet loading machine

    International Nuclear Information System (INIS)

    Kee, R.W.; Denero, J.V.

    1975-01-01

    An apparatus for loading nuclear fuel pellets on trays for transfer in a system is described. A conveyor supplies pellets from a source to a loading station. When the pellets reach a predetermined position at the loading station, a manual or automatically operated arm pushes the pellets into slots on a tray and this process is repeated until pellet sensing switches detect that the tray is full. Thereupon, the tray is lowered onto a belt or other type conveyor and transferred to other apparatus in the system, such as a furnace for sintering, and in some cases, reduction of UO 2 . 2 to UO 2 . The pellets are retained on the tray and subsequently loaded directly into fuel rods to be used in the reactor core. (auth)

  5. Fabrication and testing of main sodium pumps of Superphenix 1

    International Nuclear Information System (INIS)

    Noel, H.; Pasqualini, G.

    1985-01-01

    The complexity of the loads involved and the extremely fine analysis required necessitates extensive design calculations for the Superphenix 1 primary and secondary pumps and associated expansion tanks, aiming toward detailed design validation, after slight adjustments, mainly to the secondary pumps and expansion tanks. The component parts to be built were far larger than those for the previous pumps (Rapsodie, Phenix), with very low manufacturing tolerances, which led to precision machining and welding operations, together with numerous dimensional inspections and materials characterization tests to achieve the required quality standards

  6. Fuel loading and homogeneity analysis of HFIR design fuel plates loaded with uranium silicide fuel

    International Nuclear Information System (INIS)

    Blumenfeld, P.E.

    1995-08-01

    Twelve nuclear reactor fuel plates were analyzed for fuel loading and fuel loading homogeneity by measuring the attenuation of a collimated X-ray beam as it passed through the plates. The plates were identical to those used by the High Flux Isotope Reactor (HFIR) but were loaded with uranium silicide rather than with HFIR's uranium oxide fuel. Systematic deviations from nominal fuel loading were observed as higher loading near the center of the plates and underloading near the radial edges. These deviations were within those allowed by HFIR specifications. The report begins with a brief background on the thermal-hydraulic uncertainty analysis for the Advanced Neutron Source (ANS) Reactor that motivated a statistical description of fuel loading and homogeneity. The body of the report addresses the homogeneity measurement techniques employed, the numerical correction required to account for a difference in fuel types, and the statistical analysis of the resulting data. This statistical analysis pertains to local variation in fuel loading, as well as to ''hot segment'' analysis of narrow axial regions along the plate and ''hot streak'' analysis, the cumulative effect of hot segment loading variation. The data for all twelve plates were compiled and divided into 20 regions for analysis, with each region represented by a mean and a standard deviation to report percent deviation from nominal fuel loading. The central regions of the plates showed mean values of about +3% deviation, while the edge regions showed mean values of about -7% deviation. The data within these regions roughly approximated random samplings from normal distributions, although the chi-square (χ 2 ) test for goodness of fit to normal distributions was not satisfied

  7. Automated nuclear fuel rod pattern loading system

    International Nuclear Information System (INIS)

    Lambert, D.V.; Nyland, T.W.; Byers, J.W.; Haley, D.E. Jr.; Cioffi, J.V.

    1990-01-01

    This patent describes an apparatus for loading fuel rods in a desired pattern. It comprises: a carousel having a plurality of movable gondolas for stocking thereon fuel rods of known enrichments; an elongated magazine defining a matrix of elongated slots being open at their forward ends for receiving fuel rods; a workstation defining a fuel rod feed path; and a holder and indexing mechanism for movably supporting the magazine and being actuatable for moving the magazine along X-Y axes to successively align one at a time selected ones of the slots with the feed path for loading in the magazine the successive fuel rods in a desired enrichment pattern

  8. Fabrication of HTTR first loading fuel

    International Nuclear Information System (INIS)

    Kato, S.; Yoshimuta, S.; Hasumi, T.; Sato, K.; Sawa, K.; Suzuki, S.; Mogi, H.; Shiozawa, S.; Tanaka, T.

    2001-01-01

    This paper summarizes the fabrication of the first loading fuel for HTTR, High Temperature engineering Test Reactor constructed by JAERI, Japan Atomic Energy Research Institute. The fuel fabrication started at the HTR fuel facility of NFI, Nuclear Fuel Industries, Ltd., June 1995. 4,770 fuel rods were fabricated through the fuel kernel, coated fuel particle and fuel compaction process, then 150 fuel elements were assembled in the reactor building December 1997. Fabrication technology for the fuel was established through a lot of R and D activities and fabrication experience of irradiation examination samples spread over about 30 years. Most of all, very high quality and production efficiency of fuel were achieved by the development of the fuel kernel process using the vibration dropping technology, the continuous 4-layer coating process and the automatic compaction process. As for the inspection technology, the development of the automatic measurement equipment for coated layer thickness of a coated fuel particle and uranium content of a fuel compact contributed to the higher reliability and rationalization of the inspection process. The data processing system for the fabrication and quality control, which was originally developed by NFI, made possible not only quick feedback of statistical quality data to the fabrication processes, but also automatic document preparation, such as inspection certificates and accountability control reports. The quality of the first loading fuel fully satisfied the design specifications for the fuel. In particular, average bare uranium fraction and SiC defective fraction of fuel compacts were 2x10 -6 and 8x10 -5 , respectively. According to the preceding irradiation examinations being performed at JMTR, Japan Materials Testing Reactor of JAERI, the specimen sampled from the first loading fuel shows good irradiation performance. (author)

  9. Fuel tank crashworthiness : loading scenarios

    Science.gov (United States)

    2011-03-16

    The Federal Railroad Administrations Office of Research and Development is conducting research into fuel tank crashworthiness. The breaching of fuel tanks during passenger : rail collisions and derailments increases the potential of serious injury...

  10. Bayesian techniques for surface fuel loading estimation

    Science.gov (United States)

    Kathy Gray; Robert Keane; Ryan Karpisz; Alyssa Pedersen; Rick Brown; Taylor Russell

    2016-01-01

    A study by Keane and Gray (2013) compared three sampling techniques for estimating surface fine woody fuels. Known amounts of fine woody fuel were distributed on a parking lot, and researchers estimated the loadings using different sampling techniques. An important result was that precise estimates of biomass required intensive sampling for both the planar intercept...

  11. Automated nuclear fuel rod pattern loading system

    International Nuclear Information System (INIS)

    Lambert, D.V.; Nylund, T.W.; Byers, J.W.; Haley, D.E. Jr.; Cioffi, J.V.

    1991-01-01

    This patent describes a method for loading fuel rods in a desired pattern. It comprises providing a supply of fuel rods of known enrichments; providing a magazine defining a matrix of elongated slots open at their forward ends for receiving fuel rods; defining a fuel rod feed path; receiving successively one at a time along the feed path fuel rods selected from the supply thereof; verifying successively one at a time along the feed path the identity of the selected fuel rods, the verifying including blocking passage of each selected fuel rod along the feed path until the identity of each selected fuel rod is confirmed as correct; feeding to the magazine successively one at a time along the feed path the selective and verified fuel rods; and supporting and moving the magazine along X-Y axes to successively align one at a time selected ones of the slots with the feed path for loading in the magazine the successive fuel rods in a desired enrichment pattern

  12. Automated system for loading nuclear fuel pins

    International Nuclear Information System (INIS)

    Marshall, J.L.

    1983-10-01

    A completely automatic and remotely controlled fuel pin fabrication system is being designed by the Westinghouse Hanford Company. The Pin Operations System will produce fuel pins for the Fast Flux Test Facility (FFTF) and the Clinch River Breeder Reactor Plant (CRBRP). The system will assemble fuel pin components into cladding tubes in a controlled environment. After fuel loading, the pins are filled with helium, the tag gas capsules are inserted, and the top end cap welded. Following welding, the pins are surveyed to assure they are free of contamination and then the pins are helium leak tested

  13. Monitoring of pipe displacements in French LMFBR SUPERPHENIX

    International Nuclear Information System (INIS)

    Foucher, N.; Debaene, J.P.; Renault, Y.; Blin, B.

    1993-01-01

    In order to check that pipe supports work properly and that the locking of snubbers or the loss of supports do not put a pipe in unacceptable loading conditions, a monitoring of the behaviour of the main pipes of SUPERPHENIX is planned. This monitoring system consists in measuring the displacements at selected points of the pipe by means of measuring rods and checking that these displacements remain inside allowable domains. These allowable domains are defined so that, if the displacements of the pipe are inside all these domains, the plant operator is sure that the stresses verify the allowable limits and then no additional inspection is carried out. In the opposite case, the operator will inspect the pipe in detail in order to determine the consequences and repair if necessary before restarting. Selection of points for monitoring was done with the to minimize the number of measures to be carried out and to use as far as possible the measuring rods that were installed to check that pipe displacements were consistent with what has been obtained in design calculations. However, it appears necessary to ensure that any incident occurring at any point of the pipe can be detected and, if necessary, additional measuring rods may be installed. An incident is said detectable if it induces on at least one measuring rod a deviation with respect to expected displacement not lower than 5 mm. It has been chosen so that small normal changes in measured displacements are not mistaken as incidents. The incidents that are supposed likely to occur are: 1) loss of a support which induces mainly primary stresses, 2) locking of a snubber which induces mainly secondary stresses. Monitoring of pipe displacements is a simple and effective way of checking that no damaging perturbation has occurred on the pipe. Calculations carried out on the DHR loops of SUPERPHENIX show that allowable domains of acceptable size may be obtained using a relatively small number of measuring rods. The method

  14. Apparatus for loading fuel pellets in fuel rods

    International Nuclear Information System (INIS)

    Tedesco, R.J.

    1976-01-01

    An apparatus is disclosed for loading fuel pellets into fuel rods for a nuclear reactor including a base supporting a table having grooves therein for holding a multiplicity of pellets. Multiple fuel rods are placed in alignment with grooves in the pellet table and a guide member channels pellets from the table into the corresponding fuel rods. To effect movement of pellets inside the fuel rods without jamming, a number of electromechanical devices mounted on the base have arms connected to the lower surface of the fuel rod table which cyclically imparts a reciprocating arc motion to the table for moving the fuel pellets longitudinally of and inside the fuel rods. These electromechanical devices include a solenoid having a plunger therein connected to a leaf type spring, the arrangement being such that upon energization of the solenoid coil, the leaf spring moves the fuel rod table rearwardly and downwardly, and upon deenergization of the coil, the spring imparts an upward-forward movement to the table which results in physical displacement of fuel pellets in the fuel rods clamped to the table surface. 8 claims, 6 drawing figures

  15. Fuel cladding behavior under rapid loading conditions

    Science.gov (United States)

    Yueh, K.; Karlsson, J.; Stjärnsäter, J.; Schrire, D.; Ledergerber, G.; Munoz-Reja, C.; Hallstadius, L.

    2016-02-01

    A modified burst test (MBT) was used in an extensive test program to characterize fuel cladding failure behavior under rapid loading conditions. The MBT differs from a normal burst test with the use of a driver tube to simulate the expansion of a fuel pellet, thereby producing a partial strain driven deformation condition similar to that of a fuel pellet expansion in a reactivity insertion accident (RIA). A piston/cylinder assembly was used to pressurize the driver tube. By controlling the speed and distance the piston travels the loading rate and degree of sample deformation could be controlled. The use of a driver tube with a machined gauge section localizes deformation and allows for continuous monitoring of the test sample diameter change at the location of maximum hoop strain, during each test. Cladding samples from five irradiated fuel rods were tested between 296 and 553 K and loading rates from 1.5 to 3.5/s. The test rods included variations of Zircaloy-2 with different liners and ZIRLO, ranging in burn-up from 41 to 74 GWd/MTU. The test results show cladding ductility is strongly temperature and loading rate dependent. Zircaloy-2 cladding ductility degradation due to operational hydrogen pickup started to recover at approximately 358 K for test condition used in the study. This recovery temperature is strongly loading rate dependent. At 373 K, ductility recovery was small for loading rates less than 8 ms equivalent RIA pulse width, but longer than 8 ms the ductility recovery increased exponentially with increasing pulse width, consistent with literature observations of loading rate dependent brittle-to-ductile (BTD) transition temperature. The cladding ductility was also observed to be strongly loading rate/pulse width dependent for BWR cladding below the BTD temperature and Pressurized Water Reactor (PWR) cladding at both 296 and 553 K.

  16. Preserving the memory: a strategic issue. The case of Superphenix

    International Nuclear Information System (INIS)

    Anon.

    2015-01-01

    In 1998 when Superphenix was decommissioned, a specific know-how and a collective memory of the activities and events began to disappear, what remained was archive files in which a multitude of more or less useful documents were kept. In 2006 a large campaign was launched to locate and interview the ancient employees and ask them about technical issues and choices made years before. About 40 interviews were performed and pieces of oral information covering Superphenix construction and operation were collected, transcribed and added to a database. This know-how concerning sodium-cooled reactors will be extremely useful for the Astrid project. (A.C.)

  17. Study on the HTGR axial fuel loading

    International Nuclear Information System (INIS)

    Tanaka, Ryokichi

    1981-01-01

    In the nuclear and thermal design of reactor cores, it is one of the important targets for reactor safety to flatten fuel temperature distribution as far as possible to prevent local peaking. As a macroscopic method to prevent temperature peaking, it is considered to give exponential type power output distribution in coolant flow direction, while flattening radial power output distribution. Assuming rod-shaped fuel, the distribution of fuel heat generation is given by an exponential function under constant maximum fuel temperature condition in the direction of channel. By applying this function to neutron source distribution, and in a premise that U-235 loading can be changed continuously, the preliminary investigation on no-reflector core by one-dimensional one-group consideration, and then the analytical solution of the diffusion equation for a core with reflectors by two group one-dimensional approximation were carried out. The results of these investigations revealed that the U-235 concentration required for achieving exponential type power output distribution is necessary to have large concentration gradient up to the distance equivalent to the length of a few fuel elements from the core inlet, but it is sufficient to have constant concentration in downstream fuel elements, which is 0.8 to 0.9 times as much as the average value along the channel, except for large flow rate channel. (Wakatsuki, Y.)

  18. Data processing and data collection in Super-Phenix

    International Nuclear Information System (INIS)

    Josue, M.; Thegner, G.

    1978-01-01

    The data processing systems for the Super-Phenix power station have been developed from Phenix systems, the various tasks being specified on the basis of the origin of information (specific to the boiler or common to the whole power station) and of its nature, i.e. depending on whether it is used for protection or for operational purposes or whether it provides personnel with a better understanding of phenomena related to the reactor. The data processing systems specific to the boiler are as follows: (a) the core temperature processing system (TRTC) with which fuel assembly temperatures can be monitored and any abnormally high value discovered, in which case it can cause a trip to shut down the reactor. To this extent it can be seen as part of the station safety equipment. In the interest of channel separation and satisfactory availability, the system is made up of two identical units based on the use of mini-computers, some of which (for analog acquisition) are decentralized and placed near the measuring points in the dome; and (b) the core fault detection and diagnosis system (DDDC), which is a necessary complement to the TRTC in that it fulfils certain boiler operation tasks and supplies information if incidents occur. It is made up of three subsystems (acquisition and retrieval, reactivity comparison, noise analysis). Among the systems applicable to the overall operation of the station, there is: (c) the complementary information processing system (TCI) which provides overall control and is based on a large quantity of information connected with the facility as a whole. (author)

  19. Superphenix: Is the fast breeder dream over -- or over yonder?

    International Nuclear Information System (INIS)

    Anon.

    1997-01-01

    A detailed history of France's Superphenix commercial fast breeder reactor project is presented. Important project milestones are discussed from the project's conception in 1971 to its current status. Recommendations of the Castaing Commission on the project and future plans for use of the reactor are outlined. In addition, world wide fast breeder projects are listed and discussed

  20. The deadly sins of high technology: Superphenix, Eurotunnel, Ariane 5.

    International Nuclear Information System (INIS)

    Bell, R.; Jeanmougin, Ch.

    1998-01-01

    Based on a detailed analysis of civil or military high technology projects (Superphenix reactor, English Channel tunnel, Ariane 5 launcher, etc..), mainly of European origin, this book reveals seven systematic sins in the realization of these projects: an abolishment of controls, a premature construction, a manumission of suppliers, a no-share of risks attitude, political manipulations, fraudulence and secrecy. (J.S.)

  1. In-service inspection in the Superphenix 1 vessels interspace

    International Nuclear Information System (INIS)

    Asty, M.; Saglio, R.

    1983-03-01

    The design of Superphenix 1 reactor vessels allows their in-service inspection. A self-propelling engine, the MIR, has been concieved for this need: it can do a visual and ultrasonic inspection. The MIR can move in the whole vessels interspace. The operating conditions are specified and the principle characteristics of the MIR engine are presented [fr

  2. Review on Fuel Loading Process and Performance for Advanced Fuel Handling Equipment

    International Nuclear Information System (INIS)

    Chang, Sang-Gyoon; Lee, Dae-Hee; Kim, Young-Baik; Lee, Deuck-Soo

    2007-01-01

    The fuel loading process and the performance of the advanced fuel handling equipment for OPR 1000 (Optimized Power Plant) are analyzed and evaluated. The fuel handling equipment, which acts critical processes in the refueling outage, has been improved to reduce fuel handling time. The analysis of the fuel loading process can be a useful tool to improve the performance of the fuel handling equipment effectively. Some recommendations for further improvement are provided based on this study

  3. The integrity of CANDU fuel during load following

    International Nuclear Information System (INIS)

    Tayal, M.; Manzer, A.M.; Sejnoha, R.; Hains, A.J.

    1989-08-01

    This paper summarizes data and analyses of integrity and of physics of CANDU fuel during load following. Measurements of irradiated fuel show that power cycles do not enhance release of fission gas. Data from research reactors show that the power cycles cause cyclic strains in the sheath. Finite element analyses show that the cyclic strains give highly multiaxial stresses in the sheath. The stresses and the strains are well into the plastic range. The cyclic loads 'use up' some fraction of the sheath's resistance to environmentally-assisted cracking (EAC), depending on the details of the fuel design and of then power cycles. The balance of the sheath's resistance to EAC continues to be available to counteract static loads. Thousands of fuel bundles have experienced many power cycles in research and in commercial reactors. Overall integrity of fuel bundles is well over 99%. Thus, CANDU fuel continues to show good performance in both base-load and load-following reactors

  4. Forest fuel reduces the nitrogen load

    International Nuclear Information System (INIS)

    Lundborg, A.

    1993-03-01

    A study of the literature was made on the basis of the following hypothesis: ''If nitrogen-rich felling residues are removed from the forest, the nitrogen load on the forest ecosystem is decreased and the risk of nitrogen saturation also decreases''. The study was designed to provide information on how the nitrogen situation is influenced if felling residues are removed from nitrogen-loaded forests and used as fuel. Felling residues release very little nitrogen during the first years after felling. They can immobilize nitrogen from the surroundings, make up a considerable addition to the nitrogen store in the soil, but also release nitrogen in later stages of degradation. The slash has an influence on the soil climate and thus on soil processes. Often there is an increase in the mineralization of litter and humus below the felling residues. At the same time, nitrification is favoured, particularly if the slash is left in heaps. Felling residues contain easily soluble nutrients that stimulate the metabolization of organic matter that otherwise is rather resistant to degradation. The slash also inhibits the clear-cut vegetation and its uptake of nitrogen. These effects result in increased leaching of nitrogen and minerals if the felling residues are left on the site. (99 refs.)

  5. FFTF initial fuel loading, preanalyses, and comparison with preliminary results

    International Nuclear Information System (INIS)

    Rothrock, R.B.; Daughtry, J.W.; Zimmerman, B.D.; Petrowicz, N.E.; Bennett, R.A.; Ombrellaro, P.A.

    1980-02-01

    Disadvantages of conventional loading from the center out were circumvented by loading one trisector at a time, and connecting the control rod drivelines in each sector after it was loaded so that the rods could be operated during the loading of subsequent trisectors. This sequence was interrupted once during the loading of the final sector, to achieve initial criticality at an approximately minimum critical loading and to measure absolute subcriticality by the rod drop technique. An in-core detector was preferable to the standard FTR ex-core detectors for monitoring the initial fuel loading. Consequently, special fission chambers were installed in an instrument thimble near the core center to monitor the initial fuel loading

  6. Load-following performance and assessment of CANDU fuel

    Energy Technology Data Exchange (ETDEWEB)

    Tayal, M.; Floyd, M.; Rattan, D.; Xu, Z.; Manzer, A.; Lau, J. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada); Kohn, E. [Ontario Power Generation, Fuel and Fuel Channel Analysis Dept., Toronto, Ontario (Canada)

    1999-09-01

    Load following of nuclear reactors is now becoming an economic necessity in some countries. When nuclear power stations are operated in a load-following mode, the reactor and the fuel may be subjected to step changes in power on a weekly, daily, or even hourly basis, depending on the grid's needs. This paper updates the previous surveys of load-following capability of CANDU fuel, focusing mainly on the successful experience at the Bruce B station. As well, initial analytical assessments are provided that illustrate the capability of CANDU fuel to survive conditions other than those for which direct in-reactor evidence is available. (author)

  7. In service inspection of SUPERPHENIX 1 vessels: MIR

    International Nuclear Information System (INIS)

    Asty, M.; Viard, J.; Lerat, B.; Saglio, R.

    1985-01-01

    Although no in-service inspection constraints were imposed on the Phenix vessels, the Safety Authorities asked that the design of SUPERPHENIX 1 makes it possible to monitor throughout the lifetime of the reactor, surface and internal defects on the main vessel. A pool design and the presence of heat baffles inside the main vessel make access from the inside of the vessel impossible. Thus, an inspection can only be performed from the outside of the main vessel: the distance between the walls of the main and safety vessels is such that an inspection device can be introduced into the corresponding space. As the design of the reactor precludes radiographic inspection, the method which was selected for monitoring internal defects in the main vessel is ultrasonics. However, the anisotropic structure of austenitic stainless steel welds limits the performance of this technique. The authors present the in-service inspection device, MIR, which has been specially developed for the visual and ultrasonic examination of SUPERPHENIX 1 vessels

  8. Loads on pebble bed fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Teuchert, E.; Maly, V.

    1974-03-15

    A comparison is made of key parameters for multi-recycle pebbles and single-pass once-through (OTTO) pebbles. The parameters analyzed include heat transfer characteristics with burn-up, temperature profiles, power per element as a function of axial position in the core, and burn-up. For the OTTO-scheme, the comparisons addressed the use of the conventional fuel element and the advanced "shell ball" designed to reduce the peak fuel temperature in the center of the fuel element. All studies addressed the uranium-thorium fuel cycle.

  9. Loading procedures for shipment of irradiated fuel

    Energy Technology Data Exchange (ETDEWEB)

    Bates, E F; Feltz, D E; Sandel, P S; Schoenbucher, B [Texas A and M University (United States)

    1974-07-01

    The Nuclear Science Center at Texas A and M does not have proper equipment and facilities for transferring irradiated fuel from the reactor pool to the transport vehicle. To accomplish the transfer of 23 MTR type fuel elements procedures were developed using a modified fork lift and flex-lift obtained locally. The transfer was accomplished without incident and with negligible personnel exposure. (author)

  10. Loading procedures for shipment of irradiated fuel

    International Nuclear Information System (INIS)

    Bates, E.F.; Feltz, D.E.; Sandel, P.S.; Schoenbucher, B.

    1974-01-01

    The Nuclear Science Center at Texas A and M does not have proper equipment and facilities for transferring irradiated fuel from the reactor pool to the transport vehicle. To accomplish the transfer of 23 MTR type fuel elements procedures were developed using a modified fork lift and flex-lift obtained locally. The transfer was accomplished without incident and with negligible personnel exposure. (author)

  11. Studies and research relatives to the safety of the Super-Phenix project

    International Nuclear Information System (INIS)

    Anselin, F.; Penet, F.

    1978-01-01

    The analysis of safety reports concerning the Creys Malville power station (Superphenix) must be based on technical data supplied by the NERSA, responsible for the plant, and on results of research and development programme carried out in various establishments and at the CEA in particular. By virtue of the procedure laid down for the safety analysis, i.e. analysis by the barrier method, verification of reactor shut down rules at power, permanence of cooling, confinement of dangerous products in the event of hypothetical failure of the above two functions safety R and D programmes have a double aspect: accident prevention on the one hand and study of the development of accident, even the most hypothetical on the other. In the accident prevention field the studies deal with the resistance of barriers under normal and accidental working conditions, inspection systems and reactor safety functions allowing abnormal situations to be detected and the reactor shut down; whence the special emphasis placed on emergency shut-down and cooling systems. In the accident field the R and D activities cover a wide range of studies on phenomena liable to arise, independently of their probability of occurence during the lifetime of the reactor; heating in the mass or boiling of sodium, fuel, meeting, movements of fused materials, fuel-sodium thermal interaction, core deformation, resistance of confinement recovery of molten fuel, post-accident cooling, transfer of radioactivity and contamination outside the reactor, radiological consequences and means of confinement of dangerous products [fr

  12. Locomotive fuel tank structural safety testing program : passenger locomotive fuel tank jackknife derailment load test.

    Science.gov (United States)

    2010-08-01

    This report presents the results of a passenger locomotive fuel tank load test simulating jackknife derailment (JD) load. The test is based on FRA requirements for locomotive fuel tanks in the Title 49, Code of Federal Regulations (CFR), Part 238, Ap...

  13. Design of fuel loading for Bohunice V-1 Unit 2 reaktor for fuel cycle No.19

    International Nuclear Information System (INIS)

    Majercik, J.

    1998-01-01

    The report contains description of the design of fuel loading for the fuel cycle No. 19 in the V-1 Bohunice Unit 2 reactor. Input data and computer codes used for the development of the design are shown. The fuel loading is characterized by the assortment of the fuel loaded and by the scheme of re shuffling of assemblies in the core. An evaluation of basic neutronic core parameters as relates to the compliance with safety criteria is a part of the report as well

  14. Loading ion exchange resins with uranium for HTGR fuel kernels

    International Nuclear Information System (INIS)

    Notz, K.J.; Greene, C.W.

    1976-12-01

    Uranium-loaded ion exchange beads provide an excellent starting material in the production of uranium carbide microspheres for nuclear fuel applications. Both strong-acid (sulfonate) and weak-acid (carboxylate) resins can be fully loaded with uranium from a uranyl nitrate solution utilizing either a batch method or a continuous column technique

  15. Nonlinear analyses of spent-fuel racks for consolidated fuel loading

    International Nuclear Information System (INIS)

    Kabir, A.F.; Godha, P.C.; Malik, L.E.; Bolourchi, S.

    1987-01-01

    Storage racks for spent-fuel assemblies in nuclear power plants are designed to withstand various combinations of loads generated by gravity, seismic, thermal, and accidental fuel drops. Due to the need for storing increased amounts of spent fuel in the existing fuel pools, many nuclear power utilities are evaluating existing fuel racks to safely carry the additional loads. The current study presents the seismic analyses of existing fuel racks of Northeast Utility Company's Millstone Unit Number 1 (BWR Mark I) nuclear plant to accommodate a 2:1 fuel consolidation. This objective requires rigorous nonlinear analyses to establish the full available capacities of the racks and thereby avoid expensive modifications or minimize any needed upgrades

  16. Application of the robust design concept for fuel loading pattern

    International Nuclear Information System (INIS)

    Endo, Tomohiro; Ohori, Kazuma; Yamamoto, Akio

    2011-01-01

    Application of the robust design concept for fuel loading pattern design is proposed as a new approach to improve the prediction accuracy of core characteristics. The robust design is a design concept that establishes a resistant (robust) system for perturbations or noises, by properly setting design variables. In order to apply the concept of robust design to fuel loading pattern design, we focus on a theoretical approach based on the higher order perturbation method. This approach indicates that the eigenvalue separation is one of the effective indices to measure the robustness of a designed fuel loading pattern. In order to verify the effectiveness of the eigenvalue separation as an index of robustness, numerical analysis is carried out for typical 3-loop PWR cores, and we evaluated the correlation between the eigenvalue separation and the variation of relative assembly power due to the perturbation of the cross section. The numerical results show that the variation of relative power decreases as the eigenvalue separation increases; thus, it is confirmed that the eigenvalue separation is an effective index of robustness. Based on the eigenvalue separation of a fuel loading pattern, we discuss design guidelines of a fuel loading pattern to improve the robustness. For example, if each fuel assembly has independent uncertainty on its cross section, the robustness of the core can be enhanced by increasing the relative power at the center of the core. The proposed guidelines will be useful to design a loading pattern that has robustness for uncertainties due to cross section, calculation method, and so on. (author)

  17. A loading pattern optimization method for nuclear fuel management

    International Nuclear Information System (INIS)

    Argaud, J.P.

    1997-01-01

    Nuclear fuel reload of PWR core leads to the search of an optimal nuclear fuel assemblies distribution, namely of loading pattern. This large discrete optimization problem is here expressed as a cost function minimization. To deal with this problem, an approach based on gradient information is used to direct the search in the patterns discrete space. A method using an adjoint state formulation is then developed, and final results of complete patterns search tests by this method are presented. (author)

  18. Optimal load allocation of multiple fuel boilers.

    Science.gov (United States)

    Dunn, Alex C; Du, Yan Yi

    2009-04-01

    This paper presents a new methodology for optimally allocating a set of multiple industrial boilers that each simultaneously consumes multiple fuel types. Unlike recent similar approaches in the utility industry that use soft computing techniques, this approach is based on a second-order gradient search method that is easy to implement without any specialized optimization software. The algorithm converges rapidly and the application yields significant savings benefits, up to 3% of the overall operating cost of industrial boiler systems in the examples given and potentially higher in other cases, depending on the plant circumstances. Given today's energy prices, this can yield significant savings benefits to manufacturers that raise steam for plant operations.

  19. Diversification of fuel costs accounting for load variation

    International Nuclear Information System (INIS)

    Ruangpattana, Suriya; Preckel, Paul V.; Gotham, Douglas J.; Muthuraman, Kumar; Velástegui, Marco; Morin, Thomas L.; Uhan, Nelson A.

    2012-01-01

    A practical mathematical programming model for the strategic fuel diversification problem is presented. The model is designed to consider the tradeoffs between the expected costs of investments in capacity, operating and maintenance costs, average fuel costs, and the variability of fuel costs. In addition, the model is designed to take the load curve into account at a high degree of resolution, while keeping the computational burden at a practical level. The model is illustrated with a case study for Indiana's power generation system. The model reveals that an effective means of reducing the volatility of the system-level fuel costs is through the reduction of dependence on coal-fired generation with an attendant shift towards nuclear generation. Model results indicate that about a 25% reduction in the standard deviation of the generation costs can be achieved with about a 20–25% increase in average fuel costs. Scenarios that incorporate costs for carbon dioxide emissions or a moratorium on nuclear capacity additions are also presented. Highlights: ► We propose a fuel price risk management model for generation investments accounting for load shape. ► The formulation incorporates a highly refined load curve while maintaining tractability. ► We demonstrate the model for planning generation investments in the state of Indiana for 2025. ► Scenarios reflect charges for CO 2 emissions and a moratorium on new nuclear power.

  20. Fine scale vegetation classification and fuel load mapping for prescribed burning

    Science.gov (United States)

    Andrew D. Bailey; Robert Mickler

    2007-01-01

    Fire managers in the Coastal Plain of the Southeastern United States use prescribed burning as a tool to reduce fuel loads in a variety of vegetation types, many of which have elevated fuel loads due to a history of fire suppression. While standardized fuel models are useful in prescribed burn planning, those models do not quantify site-specific fuel loads that reflect...

  1. The deadly sins of high technology: Superphenix, Eurotunnel, Ariane 5..; Les peches capitaux de la haute technologie: Superphenix, Eurotunnel, Ariane 5..

    Energy Technology Data Exchange (ETDEWEB)

    Bell, R.; Jeanmougin, Ch

    1998-12-31

    Based on a detailed analysis of civil or military high technology projects (Superphenix reactor, English Channel tunnel, Ariane 5 launcher, etc..), mainly of European origin, this book reveals seven systematic sins in the realization of these projects: an abolishment of controls, a premature construction, a manumission of suppliers, a no-share of risks attitude, political manipulations, fraudulence and secrecy. (J.S.)

  2. Engine combustion control at low loads via fuel reactivity stratification

    Science.gov (United States)

    Reitz, Rolf Deneys; Hanson, Reed M; Splitter, Derek A; Kokjohn, Sage L

    2014-10-07

    A compression ignition (diesel) engine uses two or more fuel charges during a combustion cycle, with the fuel charges having two or more reactivities (e.g., different cetane numbers), in order to control the timing and duration of combustion. By appropriately choosing the reactivities of the charges, their relative amounts, and their timing, combustion can be tailored to achieve optimal power output (and thus fuel efficiency), at controlled temperatures (and thus controlled NOx), and with controlled equivalence ratios (and thus controlled soot). At low load and no load (idling) conditions, the aforementioned results are attained by restricting airflow to the combustion chamber during the intake stroke (as by throttling the incoming air at or prior to the combustion chamber's intake port) so that the cylinder air pressure is below ambient pressure at the start of the compression stroke.

  3. Engine combustion control at low loads via fuel reactivity stratification

    Energy Technology Data Exchange (ETDEWEB)

    Reitz, Rolf Deneys; Hanson, Reed M.; Splitter, Derek A.; Kokjohn, Sage

    2017-12-26

    A compression ignition (diesel) engine uses two or more fuel charges during a combustion cycle, with the fuel charges having two or more reactivities (e.g., different cetane numbers), in order to control the timing and duration of combustion. By appropriately choosing the reactivities of the charges, their relative amounts, and their timing, combustion can be tailored to achieve optimal power output (and thus fuel efficiency), at controlled temperatures (and thus controlled NOx), and with controlled equivalence ratios (and thus controlled soot). At low load and no load (idling) conditions, the aforementioned results are attained by restricting airflow to the combustion chamber during the intake stroke (as by throttling the incoming air at or prior to the combustion chamber's intake port) so that the cylinder air pressure is below ambient pressure at the start of the compression stroke.

  4. Burnup performance of rock-like oxide (ROX) fuel in small pebble bed reactor with accumulative fuel loading scheme

    International Nuclear Information System (INIS)

    Simanullang, Irwan Liapto; Obara, Toru

    2017-01-01

    Highlights: • Burnup performance using ROX fuel in PBR with accumulative fuel loading scheme was analyzed. • Initial excess reactivity was suppressed by reducing 235 U enrichment in the startup condition. • Negative temperature coefficient was achieved in all condition of PBR with accumulative fuel loading scheme using ROX fuel. • Core lifetime of PBR with accumulative fuel loading scheme using ROX fuel was shorter than with UO 2 fuel. • In PBR with accumulative fuel loading scheme using ROX fuel, achieved discharged burnup can be as high as that for UO 2 fuel. - Abstract: The Japan Atomic Energy Agency (JAEA) has proposed rock-like oxide (ROX) fuel as a new, once-through type fuel concept. Here, burnup performance using ROX fuel was simulated in a pebble bed reactor with an accumulative fuel loading scheme. The MVP-BURN code was used to simulate the burnup calculation. Fuel of 5 g-HM/pebble with 20% 235 U enrichment was selected as the optimum composition. Discharged burnup could reach up to 218 GWd/t, with a core lifetime of about 8.4 years. However, high excess reactivity occurred in the initial condition. Initial fuel enrichment was therefore reduced from 20% to 4.65% to counter the initial excess reactivity. The operation period was reduced by the decrease of initial fuel enrichment, but the maximum discharged burnup was 198 GWd/t. Burnup performance of ROX fuel in this reactor concept was compared with that of UO 2 fuel obtained previously. Discharged burnup for ROX fuel in the PBR with an accumulative fuel loading scheme was as high as UO 2 fuel. Maximum power density could be lowered by introducing ROX fuel compared to UO 2 fuel. However, PBR core lifetime was shorter with ROX fuel than with UO 2 fuel. A negative temperature coefficient was achieved for both UO 2 and ROX fuels throughout the operation period.

  5. Stress analysis of fuel assemblies under seismic load

    International Nuclear Information System (INIS)

    Kiselev, A.; Krutko, E.; Kiselev, I.; Tutnov, A.

    2011-01-01

    One of the important parts of fuel assemblies (FA) safety validation is their strength estimation under the dynamic loads, such as the vibration effects caused by the work of reactor units and the seismic exposure of an earthquake, leading to extreme inertia loads on all elements of the NPP. Taking into account structural features of FA and a very large mass, the exposure of seismic loads can lead to significant deformation of fuel assemblies. It is necessary to assess the magnitude of the force interaction between the FA in case of an earthquake to estimate the strength and performance of fuel assemblies. It is also necessary to compute FA bending forms and maximum values for further RPS control rods inserting time estimation, and for disassembly possibility justification of the core and individual FA after the earthquake. The problem of WWER-1000 core dynamic behavior modeling with TVS-2M fuel assemblies under the seismic loads exposure using the finite element method is described. Each fuel assembly is represented by equivalent rod finite element model. The reactor core is simulated by 163 fuel assemblies in accordance with the reactor core construction. Stiffness characteristics of fuel assemblies are determined on the results of a series of static and dynamic TVS-2M FA field tests. The special algorithm was developed to consider the fuel rod slippage effect during deformation. The special contact elements are introduced into the model of the core to take into account the interaction of fuel assemblies with their neighbors and with core barrel. Solution of the dynamic equilibrium equations system of finite element model is implemented by direct integration using the explicit scheme. Parallel algorithms for numerical integration on multiprocessor computers with graphics processing unit is developed to improve the efficiency of calculations. Values of nodes displacement in finite element model of reactor core as a function of seismic excitation time are obtained

  6. Reduction of repository heat load using advanced fuel cycles

    International Nuclear Information System (INIS)

    Preston, Jeff; Miller, L.F.

    2008-01-01

    With the geologic repository at Yucca Mountain already nearing capacity full before opening, advanced fuel cycles that introduce reprocessing, fast reactors, and temporary storage sites have the potential to allow the repository to support the current reactor fleet and future expansion. An uncertainty analysis methodology that combines Monte Carlo distribution sampling, reactor physics data simulation, and neural network interpolation methods enable investigation into the factor reduction of heat capacity by using the hybrid fuel cycle. Using a Super PRISM fast reactor with a conversion ratio of 0.75, burn ups reach up to 200 MWd/t that decrease the plutonium inventory by about 5 metric tons every 12 years. Using the long burn up allows the footprint of 1 single core loading of FR fuel to have an integral decay heat of about 2.5x10 5 MW*yr over a 1500 year period that replaces the footprint of about 6 full core loadings of LWR fuel for the number of years required to fuel the FR, which have an integral decay heat of about.3 MW*yr for the same time integral. This results in an increase of a factor of 4 in repository support capacity from implementing a single fast reactor in an equilibrium cycle. (authors)

  7. Fuel element load/unload machine for the PEC reactor

    International Nuclear Information System (INIS)

    Clayton, K.F.

    1984-01-01

    GEC Energy Systems Limited are providing two fuel element load/unload machines for use in the Italian fast reactor programme. One will be used in the mechanism test facility (IPM) at Casaccia, to check the salient features of the machine operating in a sodium environment prior to the second machine being installed in the PEC Brasimone Reactor. The machine is used to handle fuel elements, control rods and other reactor components in the sodium-immersed core of the reactor. (U.K.)

  8. Apparatus for feeding nuclear fuel pellets to a loading tray

    International Nuclear Information System (INIS)

    Huggins, T.B.

    1979-01-01

    Apparatus for feeding nuclear fuel pellets at a uniform predetermined rate between pellet centering and grinding apparatus and a tray for loading pellets into nuclear fuel rod. Pellets discharged from the grinding apparatus are conveyed by a belt to a drive wheel forcing the pellets in engagement with the belt. The pellets under the drive wheel are capable of pushing a line of about 36 pellets onto a pellet dumping mechanism. As the dumping mechanism is actuated to dump the pellets on to a loading tray, the pellets moving toward the mechanism are stopped and the drive wheel is simultaneously lifted off the pellets until the pellet dumping process is completed. (U.K.)

  9. Optimal fuel loading pattern design using artificial intelligence techniques

    International Nuclear Information System (INIS)

    Kim, Han Gon; Chang, Soon Heung; Lee, Byung Ho

    1993-01-01

    The Optimal Fuel Shuffling System (OFSS) is developed for optimal design of PWR fuel loading pattern. OFSS is a hybrid system that a rule based system, a fuzzy logic, and an artificial neural network are connected each other. The rule based system classifies loading patterns into two classes using several heuristic rules and a fuzzy rule. A fuzzy rule is introduced to achieve more effective and fast searching. Its membership function is automatically updated in accordance with the prediction results. The artificial neural network predicts core parameters for the patterns generated from the rule based system. The back-propagation network is used for fast prediction of core parameters. The artificial neural network and the fuzzy logic can be used as the tool for improvement of existing algorithm's capabilities. OFSS was demonstrated and validated for cycle 1 of Kori unit 1 PWR. (Author)

  10. Maximum thermal loading test of BWR fuel assembly

    International Nuclear Information System (INIS)

    Nakajima, Yoshitaka; Yoshimura, Kunihiro; Nakamura, Satoshi; Ishizuka, Takao.

    1987-01-01

    Various proving tests on the reliability of nuclear power plants have been conducted at the Nuclear Power Engineering Test Center and at the Japan Power Plant Engineering and Inspection Corporation. The tests were initiated at the request of the Ministry of International Trade and Industry (MITI). Toshiba undertook one of the proving tests on the reliability of nuclear fuel assembly; the maximum thermal loading test of BWR fuel assembly from the Nuclear Power Engineering Test Center. These tests are part of the proving tests mentioned above, and their purpose is to confirm the reliability of the thermal hydraulic engineering techniques. Toshiba has been engaged for the past nine years in the design, fabrication and testing of the equipment. For the project, a test model fuel assembly was used to measure the critical power of the BWR fuel assembly and the void and fluidity of the coolant. From the test results, it has been confirmed that the heat is transferred safely from the fuel assembly to the coolant in the BWR nuclear power plant. In addition, the propriety and reliability of the thermal hydraulic engineering techniques for the fuel assembly have been proved. (author)

  11. Study on ant colony optimization for fuel loading pattern problem

    International Nuclear Information System (INIS)

    Kishi, Hironori; Kitada, Takanori

    2013-01-01

    Modified ant colony optimization (ACO) was applied to the in-core fuel loading pattern (LP) optimization problem to minimize the power peaking factor (PPF) in the modeled 1/4 symmetry PWR core. Loading order was found to be important in ACO. Three different loading orders with and without the adjacent effect between fuel assemblies (FAs) were compared, and it was found that the loading order from the central core is preferable because many selections of FAs to be inserted are available in the core center region. LPs were determined from pheromone trail and heuristic information, which is a priori knowledge based on the feature of the problem. Three types of heuristic information were compared to obtain the desirable performance of searching LPs with low PPF. Moreover, mutation operation, such as the genetic algorithm (GA), was introduced into the ACO algorithm to avoid searching similar LPs because heuristic information used in ACO tends to localize the searching space in the LP problem. The performance of ACO with some improvement was compared with those of simulated annealing and GA. In conclusion, good performance can be achieved by setting proper heuristic information and mutation operation parameter in ACO. (author)

  12. Spring retainer apparatus and method for facilitating loading of fuel rods into a fuel assembly grid

    International Nuclear Information System (INIS)

    De Mario, E.E.

    1988-01-01

    For use with a fuel assembly having at least one grid formed of interleaved straps defining hollow cells for respectively receiving fuel rods, at least some of the straps being disposed in pairs thereof so as to form springs in pairs therof being positioned in back-to-back relationships between adjacent ones of the cells, the springs in each pair thereof being configured to normally assume expanded positions in which they are displaced away from one another to engage fuel rods received in the respective cells and being deflectible to retracted positions in which they are displaced toward one another to allow loading of the fuel rods in the respective cells without engaging the springs, a spring retainer apparatus for facilitating the loading of the fuel rods into the cells of the fuel assembly grid is described comprising: (a) elongated holder bars, each holder bar being alignable with one of the pairs of the straps of the grid which defines the pairs of springs and extendible along, and in spaced relation from, the one strap pair and between and spaced from positions occupied by fuel rods when received in the cells of the grid; and (b) supported by each of the holder bars corresponding to the pairs of springs defined by the pair of straps aligned with the holder bar

  13. Effects of salvage logging and pile-and-burn on fuel loading, potential fire behaviour, fuel consumption and emissions

    Science.gov (United States)

    Morris C. Johnson; Jessica E. Halofsky; David L. Peterson

    2013-01-01

    We used a combination of field measurements and simulation modelling to quantify the effects of salvage logging, and a combination of salvage logging and pile-and-burn fuel surface fuel treatment (treatment combination), on fuel loadings, fire behaviour, fuel consumption and pollutant emissions at three points in time: post-windstorm (before salvage logging), post-...

  14. Deterministic methods for multi-control fuel loading optimization

    Science.gov (United States)

    Rahman, Fariz B. Abdul

    We have developed a multi-control fuel loading optimization code for pressurized water reactors based on deterministic methods. The objective is to flatten the fuel burnup profile, which maximizes overall energy production. The optimal control problem is formulated using the method of Lagrange multipliers and the direct adjoining approach for treatment of the inequality power peaking constraint. The optimality conditions are derived for a multi-dimensional multi-group optimal control problem via calculus of variations. Due to the Hamiltonian having a linear control, our optimal control problem is solved using the gradient method to minimize the Hamiltonian and a Newton step formulation to obtain the optimal control. We are able to satisfy the power peaking constraint during depletion with the control at beginning of cycle (BOC) by building the proper burnup path forward in time and utilizing the adjoint burnup to propagate the information back to the BOC. Our test results show that we are able to achieve our objective and satisfy the power peaking constraint during depletion using either the fissile enrichment or burnable poison as the control. Our fuel loading designs show an increase of 7.8 equivalent full power days (EFPDs) in cycle length compared with 517.4 EFPDs for the AP600 first cycle.

  15. Remediation of a former fuel loading site using phytoremediation

    Energy Technology Data Exchange (ETDEWEB)

    Kotecha, P [Jacques Whitford Environment Ltd., Mont-Royal, PQ (Canada)

    2001-07-01

    The degradation and/or removal of pollutants from a contaminated medium is caused, mediated, and/or assisted by vegetation is defined as phytoremediation. It is a method widely used for the degradation, removal, and/or stabilisation of soils, sludges, sediments, or wastewaters. Some of the substances that can be cleaned up using phytoremediation are heavy metals, radionucleotides, petroleum hydrocarbons, energetics, chlorinated hydrocarbons, biocides, metalloids, nutrients, salts, and volatile organic contaminants. A former fuel loading site currently owned by Ultramar was remediated by Jacques Whitford Environment Limited using phytoremediation. A gasoline loading facility, a fuel loading facility, and a berm along an adjacent creek were all located at the site. All buildings and petroleum equipment had been removed in the mid-1980s, and the site is now vacant. The first phase involved the revegetation of the site with a phytoremediation grass cover and hybrid poplars, then the tree roots were allowed to infiltrate the ground to act as intake paths for contaminated water, the tree roots acted as a barrier to the contaminants headed to the river. Some of the advantages of phytoremediation are: low cost technique that can be applied using solar-powered ecotechnology in situ, wide applicability, involves minimum site disruption, has wide public acceptance, produces were few by-products requiring disposal, if any, and the harvested material can be easily disposed of in cases involving plant harvesting. The outcome of the project was also presented.

  16. Common lessons drawn from different laboratories analyses of super-phenix start-up experiments

    International Nuclear Information System (INIS)

    Cabrillat, J.C.; Salvatores, M.; Carta, M.; D'Angelo, A.; Giese, H.; De Wouters, R.; Newton, T.; Harrison, P.; Sztark, H.; Wehmann, U.

    1990-01-01

    Measurements issued from the SUPER-PHENIX start-up experiments have been analysed by the different partners within the European Community with their own data and methods. Common lessons can be drawn from the different analyses and recommendations made on the definition of the characteristics of a common European formulaire and in the actions in support of its qualification

  17. Apparatus and method for loading fuel rods into grids of a fuel assembly

    International Nuclear Information System (INIS)

    De Mario, E.E.; Burman, D.L.; Olson, C.A.; Secker, J.R.

    1987-01-01

    This patent describes a fuel assembly having fuel rods and at least one grid formed of interleaved straps and yieldable springs, the interleaved straps defining hollow cells aligned in rows and columns thereof for receiving the respective fuel rods. A pair of the springs are disposed within each of the cells for engaging and supporting one of the fuel rods when received in the cell. An apparatus is described for facilitating the loading of the fuel rods into the grid of the fuel assembly, comprising: (a) first mean insertable concurrently into the cells of the grid for engaging and moving the springs from respective first positions in which each pair of springs will engage a respective fuel rod when disposed within the grid cell to respective second positions in which each pair of springs is disengaged from the respective fuel rod when disposed within the grid cell; (b) a pair of second means, one of the pair of the second means being insertable concurrently into the rows of the cells of the grid and the other of the pair of second means being insertable concurrently into the column of the cells

  18. Nonlinear analysis and evaluation of a reinforced concrete spent fuel storage pool for accidental thermal loads

    International Nuclear Information System (INIS)

    Kabir, A.F.; Bolourchi, S.

    1991-01-01

    A feasibility study was conducted for addition of consolidated fuel racks to an existing reinforced concrete spent fuel storage pool of a Mark I BWR plant. Nonlinear analysis of a detailed three-dimensional model of the fuel pool, considering cracking in concrete under gravity and thermal load conditions, showed that the pool has reserve capacities to carry the additional loads. (author)

  19. Estimating grass fuel loads with a disc pasture meter in the Kruger ...

    African Journals Online (AJOL)

    Reports the results of a study conducted to assess the efficiency of a new calibration procedure for the disc pasture meter, used for estimating the fuel load available for combustion during fires; The major portion of the fuel load in the savanna areas comprises surface fuels in the form of the standing grass sward. The disc ...

  20. Contamination transfers during fuel transport cask loading. A concrete situation

    International Nuclear Information System (INIS)

    Fournel, B.; Turchet, J.P.; Faure, S.; Allinei, P.G.; Briquet, L.; Baubet, D.

    2002-01-01

    In 1998, a number of contamination cases detected during fuel shipments have been pointed out by the french nuclear safety authority. Wagon and casks external surfaces were partly contaminated upon arrival in Valognes railway terminal. Since then, measures taken by nuclear power plants operators in France and abroad solved the problem. In Germany, a report analyzing the situation in depth has been published in which correctives actions have been listed. In France, EDF launched a large cleanliness program (projet proprete radiologique) in order to better understand contamination transfers mechanisms during power plants exploitation and to list remediation actions to avoid further problems. In this context, CEA Department for Wastes Studies at Cadarache (CEA/DEN/DED) was in charge of a study about contamination transfers during fuel elements loading operations. It was decided to lead experiments for a concrete case. The loading of a transport cask at Tricastin-PWR-1 was followed in november 2000 and different analysis comprising water analysis and smear tests analysis were carried out and are detailed in this paper. Results are discussed and qualitatively compared to those obtained in Philippsburg-BWR, Germany for a similar set of tests. (authors)

  1. Forest fuel reduces the nitrogen load - calculations of nitrogen flows

    International Nuclear Information System (INIS)

    Burstroem, F.; Johansson, Jan.

    1995-12-01

    Nitrogen deposition in Sweden has increased strongly during recent decades, particularly in southern Sweden. Nitrogen appears to be largely accumulated in biomass and in the soil. It is therefore desirable to check the accumulation of nitrogen in the forest. The most suitable way of doing this is to remove more nitrogen-rich biomass from the forest, i.e., increase the removal of felling residues from final fellings and cleanings. An ecological condition for intensive removal of fuel is that the ashes are returned. The critical load for nitrogen, CL(N), indicates the level of nitrogen deposition that the forest can withstand without leading to ecological changes. Today, nitrogen deposition is higher than the CL(N) in almost all of Sweden. CL(N) is calculated in such a manner that nitrogen deposition should largely be balanced by nitrogen losses through harvesting during a forest rotation. The value of CL(N) thus largely depends on how much nitrogen is removed with the harvested biomass. When both stems and felling residues are harvested, the CL(N) is about three times higher than in conventional forestry. The increase is directly related to the amount of nitrogen in the removed biofuel. Use of biofuel also causes a certain amount of nitrogen emissions. From the environmental viewpoint there is no difference between the sources of the nitrogen compounds. An analysis of the entire fuel chain shows that, compared with the amount of nitrogen removed from the forest with the fuel, about 5 % will be emitted as nitrogen oxides or ammonia during combustion, and a further ca 5 % during handling and transports. A net amount of about 90 % of biomass nitrogen is removed from the system and becomes inert nitrogen (N 2 ). 60 refs, 3 figs, 4 tabs, 11 appendices

  2. The dynamic behavior of the SUPER-PHENIX reactor under unprotected transient

    International Nuclear Information System (INIS)

    Gouriou, A.; Francillon, E.; Kayser, G.; Malenfer, G.; Languille, A.

    1982-01-01

    Due to design changes and progress on the knowledge of feed-back effects, a reactualization of the dynamic behavior of SUPER-PHENIX under unprotected transients was undertaken. We present the main data on feed-back characteristics and the results of dynamic calculations. With the present state of knowledge, the former conclusion is confirmed: the dynamic evolution is very slow and no irreversible phenomena happen in the short term

  3. Synthesis method validation for Super-Phenix 1 start-up core studies

    International Nuclear Information System (INIS)

    Pipaud, J.Y.; Gastaldo, G.; Giacometti, C.

    1980-09-01

    This paper aims at presenting the systematic studies performed in order to check and to improve the synthesis method wich is used to optimize the configuration of the SUPER-PHENIX 1 start-up core versus the diluent subassembly location and the control rod ring insertion. A special attention is paid to the choice of the trial functions when the two rod rings have different insertion depths. Present limits of the synthesis method are given and further improvements are indicated

  4. Measurements of negative reactivity in Masurca and Phenix control rods: Prospects for Superphenix

    International Nuclear Information System (INIS)

    Gauthier, J.C.; Petiot, R.; Coulon, P.; Giese, H.; West, J.P.

    1986-01-01

    Experimental assessment of the negative reactivity of the control rods in an industrial reactor has recently been the subject of numerous studies conducted in the light of forthcoming startup tests on the core of Superphenix. Representative tests have been carried out both on Phenix and on the Masurca critical mockup, and a test programme for Superphenix has been drawn up. Subcritical measurements (source multiplication technique) have been carried out on Phenix without absolute measurement of a standard. However, a precise relative interpretation using two counters demonstrates good agreement following the correction of spatial effects. The chief value of the rod drop measurements conducted on Masurca was that it provided a means of cross-checking the kinetic method to be validated against a standard source multiplication method. The results demonstrate complete agreement between the two methods. The acceptability of the rod drop method is therefore considered to be established. The programme foreseen for startup of Superphenix and the objectives which have been set are briefly indicated. The calculation methods to be used in respect of the startup tests have been established on the basis of experience gained through interpreting the experiments conducted in the course of the Racine (Masurca) programme. An analysis of these experiments included, among other things, a parametric study that has made it possible to devise a standard calculation method for predicting Superphenix rod worth values. The main feature is a scattering calculation with three energy groups and three dimensions. Two-dimensional scattering and transport calculations are therefore necessary in order to define the corrective factors to be applied to this initial result. The final result of this analysis is thus made equivalent to a 25-energy-group transport calculation with an extremely small spatial mesh

  5. SUPERPHENIX: Reactor core temperatures survey by minicomputers - original aspects related to safety

    International Nuclear Information System (INIS)

    Berlin, C.; Josue, M.; Pinoteau, J.

    1986-01-01

    The system for core temperatures fast processing (TRIC) utilized in SUPERPHENIX is part of the reactor protection system. Due to the number of temperature measurements taken into account, to the specific data processing and to the rapidity required in the treatment, the use of digital computing devices is justified. The present paper describes the conception of the system in order to satisfy the special requirements for the computers used in power reactors protection systems

  6. Operational method for demonstrating fuel loading integrity in a reactor having accessible 235U fuel

    International Nuclear Information System (INIS)

    Ward, D.R.

    1979-07-01

    The Health Physics Research Reactor is a small pulse reactor at the Oak Ridge National Laboratory. It is desirable for the operator to be able to demonstrate on a routine basis that all the fuel pieces are present in the reactor core. Accordingly, a technique has been devised wherein the control rod readings are recorded with the reactor at delayed critical and corrections are made to compensate for the effects of variations in reactor height above the floor, reactor power, core temperature, and the presence of any massive neutron reflectors. The operator then compares these readings with the values expected based on previous operating experience. If this routine operational check suggests that the core fuel loading might be deficient, a more rigorous follow-up may be made

  7. Radial power distribution shaping within a PWR fuel assembly utilizing asymmetrically loaded gadolinia-bearing fuel pins

    International Nuclear Information System (INIS)

    Stone, I.Z.

    1992-01-01

    As in-core fuel management designs evolve to meet the demands of increasing energy output, more innovative methods are developed to maintain power peaking within acceptable thermal margin limits. In-core fuel management staff must utilize various loading pattern strategies such as cross-core movement of fuel assemblies, multibatch enrichment schemes, and burnable absorbers as the primary means of controlling the radial power distribution. The utilization of fresh asymmetrically loaded gadolinia-bearing assemblies as a fuel management tool provides an additional means of controlling the radial power distribution. At Siemens Nuclear Power Corporation (SNP), fresh fuel assemblies fabricated with asymmetrically loaded gadolinia-bearing fuel rods have been used successfully for several cycles of reactor operation. Asymmetric assemblies are neutronically modeled using the same tools and models that SNP uses to model symmetrically loaded gadolinia-bearing fuel assemblies. The CASMO-2E code is used to produce the homogenized macroscopic assembly cross sections for the nodal core simulator. Optimum fuel pin locations within the asymmetrical assembly are determined using the pin-by-pin PDQ7 assembly core model for each new assembly design. The optimum pin location is determined by the rod loading that minimizes the peak-to-average pin power

  8. Superphenix 1 primary handling system fabrication and testing

    International Nuclear Information System (INIS)

    Branchu, J.; Ebbinghaus, K.; Gigarel, C.

    1985-01-01

    Primary handling covers the operations performed for spent fuel removal, new fuel insertion, and the insodium storage outside the new or spent fuel vessel. This equipment typifies many of the difficulties encountered with the project as a whole: fabrication coordination when several countries are involved and design and construction of very large, relatively complex components. Detailed design studies were mainly influenced by thermal and seismic requirements, as applicable to sodium-immersed structures. Where possible, well-tried mechanical solutions were used, but widely differing techniques were involved, ranging from the high precision fabrication of structures and mechanisms comprising numerous component parts, implying complex machining operations. No particular problems were encountered during the sodium testing of the primary handling equipment. Trends for the 1500-MW (electric) breeder include investigation of the advisability of fuel storage in the core lattice and the possibility of handling system simplification

  9. Simulations of Lithium-Based Neutron Coincidence Counter for Gd-Loaded Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Cowles, Christian C. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Kouzes, Richard T. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Siciliano, Edward R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2014-10-01

    The Department of Energy Office of Nuclear Safeguards and Security (NA-241) is supporting the project Lithium-Based Alternative Neutron Detection Technology Coincidence Counting for Gd-loaded Fuels at Pacific Northwest National Laboratory for the development of a lithium-based neutron coincidence counter for nondestructively assaying Gd loaded nuclear fuel. This report provides results from MCNP simulations of a lithium-based coincidence counter for the possible measurement of Gd-loaded nuclear fuel. A comparison of lithium-based simulations and UNCL-II simulations with and without Gd loaded fuel is provided. A lithium-based model, referred to as PLNS3A-R1, showed strong promise for assaying Gd loaded fuel.

  10. Experimental study on the potential of higher octane number fuels for low load partially premixed combustion

    NARCIS (Netherlands)

    Wang, S.; van der Waart, K.; Somers, B.; de Goey, P.

    2017-01-01

    The optimal fuel for partially premixed combustion (PPC) is considered to be a gasoline boiling range fuel with an octane number around 70. Higher octane number fuels are considered problematic with low load and idle conditions. In previous studies mostly the intake air temperature did not exceed 30

  11. Combustion and emissions characteristics of diesel engine fueled by biodiesel at partial load conditions

    International Nuclear Information System (INIS)

    An, H.; Yang, W.M.; Chou, S.K.; Chua, K.J.

    2012-01-01

    Highlights: ► Impact of engine load on engine’s performance, combustion and emission characteristics. ► The brake specific fuel consumption (BSFC) increases significantly at partial load conditions. ► The brake thermal efficiency (BTE) drops at lower engine loads, and increases at higher loads. ► The partial load also influences the trend of CO emissions. -- Abstract: This paper investigated the performance, combustion and emission characteristics of diesel engine fueled by biodiesel at partial load conditions. Experiments were conducted on a common-rail fuel injection diesel engine using ultra low sulfur diesel, biodiesel (B100) and their blend fuels of 10%, 20%, 50% (denoted as B10, B20 and B50 respectively) under various loads. The results show that biodiesel/blend fuels have significant impacts on the engine’s brake specific fuel consumption (BSFC) and brake thermal efficiency (BTE) at partial load conditions. The increase in BSFC for B100 is faster than that of pure diesel with the decrease of engine load. A largest increase of 28.1% in BSFC is found at 10% load. Whereas for BTE, the results show that the use of biodiesel results in a reduced thermal efficiency at lower engine loads and improved thermal efficiency at higher engine loads. Furthermore, the characteristics of carbon monoxide (CO) emissions are also changed at partial load conditions. When running at lower engine loads, the CO emission increases with the increase of biodiesel blend ratio and the decrease of engine speed. However, at higher engine loads, an opposite trend is obtained.

  12. Effects of fuel load and moisture content on fire behaviour and heating in masticated litter-dominated fuels

    Science.gov (United States)

    Jesse K. Kreye; Leda N. Kobziar; Wayne C. Zipperer

    2013-01-01

    Mechanical fuels treatments are being used in fire-prone ecosystems where fuel loading poses a hazard, yetlittle research elucidating subsequent fire behaviour exists, especially in litter-dominated fuelbeds. To address this deficiency, we burned constructed fuelbeds from masticated sites in pine flatwoods forests in northern Florida...

  13. Random hydrodynamic loads and the vibration of fuel elements in the turbulent coolant flow in WWER fuel assembly

    International Nuclear Information System (INIS)

    Perevezentsev, V.V.

    2012-01-01

    The generalizing empirical dependences of vibration movements on the random hydrodynamic loads have been obtained. Two characteristic regions of the influence of random hydrodynamic loads on the vibration movements have been discovered. With the values of random hydrodynamic loads more than 80 N/m, a considerable increase in the intensity of vibrations has been observed. It can be explained by the slippage of fuel element in the cell of the spacing lattice [ru

  14. In service inspection of superphenix 1 vessels: MIR

    International Nuclear Information System (INIS)

    Asty, M.; Viard, J.; Lerat, B.; Saglio, R.

    1985-02-01

    Presentation of the in-service inspection device, MIR, which has been specially developed for the visual and ultrasonic examination of Super Phenix 1 vessels (surface and internal defects). The inspections take place during fuel handling operations. The inspection device is a robot with a four-wheel drive vehicle which guidance along the welds is achieved by eddy-current devices; visual examination is performed by a television camera and ultrasonic probes are specially resistent to high temperatures

  15. The Fabrication Problem Of U3Si2-Al Fuel With Uranium High Loading

    International Nuclear Information System (INIS)

    Supardjo

    1996-01-01

    The quality of U 3 Si 2 -Al dispersion fuel product is the main aim for each fabricator. Low loading of uranium fuel element is easily fabricated, but with the increased, uranium loading, homogeneity of uranium distribution is difficult to achieve and it always formed white spots, blister, and dogboning in the fuel plates. The problem can be eliminated by the increasing treatment of the fuel/Al powder. The precise selection of fuel/Al particles diameter is needed indeed to make easier in the homogeneous process of powder and the porosities arrangement in the fuel plates. The increasing of uranium loading at constant meat thickness will increase the meat hardness, therefore to withdraw the dogboning forming, the use of harder cladding materials is necessity

  16. An automatic procedure for optimizing fuel loading in consideration of the effect of burnup nonuniformity in assembly

    International Nuclear Information System (INIS)

    Wang Guoli.

    1988-01-01

    The effect of burnup nonuniformity across the assembly on optimizing fuel loading in core is investigated. Some new rules which can be used for optimizing fuel loading in the core are proposed. New automatic procedure for optimizing fuel loading in the core is described

  17. MIR: an in-service inspection device for Superphenix 1 vessels

    International Nuclear Information System (INIS)

    Asty, M.; Ceccato, S.; Lerat, B.; Viard, J.

    1986-06-01

    The main and safety vessels of SUPERPHENIX 1 were designed to allow in-service inspections. The remote controlled inspection device MIR was developed for this purpose. It allows both visual and ultrasonic examinations to be performed. Basically, MIR consists of a tetrahedral structure provided with four steering and traction wheels, two for each vessel. A computer assisted control system enables it to be driven to any position on either the main or safety vessels. Operating conditions are briefly reviewed and the main features of MIR presented

  18. In service inspection for Superphenix vessels development of ultrasonic techniques available at high temperature

    International Nuclear Information System (INIS)

    Gondard, C.

    1983-12-01

    The main and safety vessels of SUPERPHENIX 1 were designed to allow in-service inspections. The remote controlled inspection device MIR was developped for this purpose. The ultrasonic examination has required the development of all new transducers fitted with severe operating conditions prevailing in intervessels interval. A list of problems to be resolved and technological solutions which were found is given. Measurements of acoustical properties on actual probes are compared with theoretical values. It appears that concordance is good and that an in-service inspection using high temperature transducers is possible with a good spatial resolution and signal to noise ratio

  19. Study on erbium loading method to improve reactivity coefficients for low radiotoxic spent fuel HTGR

    Energy Technology Data Exchange (ETDEWEB)

    Fukaya, Y., E-mail: fukaya.yuji@jaea.go.jp; Goto, M.; Nishihara, T.

    2015-11-15

    Highlights: • We attempted and optimized erbium loading methods to improve reactivity coefficients for LRSF-HTGR. • We elucidated the mechanism of the improvements for each erbium loading method by using the Bondarenko approach. • We concluded the erbium loading method by embedding into graphite shaft is preferable. - Abstract: Erbium loading methods are investigated to improve reactivity coefficients of Low Radiotoxic Spent Fuel High Temperature Gas-cooled Reactor (LRSF-HTGR). Highly enriched uranium is used for fuel to reduce the generation of toxicity from uranium-238. The power coefficients are positive without the use of any additive. Then, the erbium is loaded into the core to obtain negative reactivity coefficients owing to the large resonance the peak of neutron capture reaction of erbium-167. The loading methods are attempted to find the suitable method for LRSF-HTGR. The erbium is mixed in a CPF fuel kernel, loaded by binary packing with fuel particles and erbium particles, and embedded into the graphite shaft deployed in the center of the fuel compact. It is found that erbium loading causes negative reactivity as moderator temperature reactivity, and from the viewpoint of heat transfer, it should be loaded into fuel pin elements for pin-in-block type fuel. Moreover, the erbium should be incinerated slowly to obtain negative reactivity coefficients even at the End Of Cycle (EOC). A loading method that effectively causes self-shielding should be selected to avoid incineration with burn-up. The incineration mechanism is elucidated using the Bondarenko approach. As a result, it is concluded that erbium embedded into graphite shaft is preferable for LRSF-HTGR to ensure that the reactivity coefficients remain negative at EOC.

  20. Evaluation of subcritical hybrid systems loaded with reprocessed fuel

    International Nuclear Information System (INIS)

    Velasquez, Carlos E.; Barros, Graiciany de P.; Pereira, Claubia; Veloso, Maria Auxiliadora F.; Costa, Antonella L.

    2015-01-01

    Highlights: • Accelerator driven systems (ADS) and fusion–fission systems are investigated for transmutation and fuel regeneration. • The calculations were performed using Monteburns code. • The results indicate the most suitable system for achieve transmutation. - Abstract: Two subcritical hybrid systems containing spent fuel reprocessed by Ganex technique and spiked with thorium were submitted to neutron irradiation of two different sources: ADS (Accelerator-driven subcritical) and Fusion. The aim is to investigate the nuclear fuel evolution using reprocessed fuel and the neutronic parameters under neutron irradiation. The source multiplication factor and fuel depletion for both systems were analysed during 10 years. The simulations were performed using MONTEBURNS code (MCNP/ORIGEN). The results indicate the main differences when irradiating the fuel with different neutron sources as well as the most suitable system for achieving transmutation

  1. Discharge Burnup Evaluation of Natural Uranium Loaded CANFLEX-43 Fuel Bundle

    International Nuclear Information System (INIS)

    Roh, Gyu Hong; Kim, Yong Hee; Kim, Won Young; Park, Joo Hwan

    2009-11-01

    Using WIMS-AECL code, which is 2-dimensional lattice core used in CANDU physics calculation, the discharge burnup of the natural uranium loaded CANFLEX-43 fuel bundle was evaluated by comparing the discharge burnup of standard 37 element fuel bundle. When the discharge burnup of the standard 37 element fuel is 7,200 MWd/MTU, that of the CANFLEX 43 fuel bundle was evaluated as 7,077 MWd/MTU, by applying the same lattice conditions for both fuel bundles

  2. Fuel loading and control rod patterns optimization in a BWR using tabu search

    International Nuclear Information System (INIS)

    Castillo, Alejandro; Ortiz, Juan Jose; Montes, Jose Luis; Perusquia, Raul

    2007-01-01

    This paper presents the QuinalliBT system, a new approach to solve fuel loading and control rod patterns optimization problem in a coupled way. This system involves three different optimization stages; in the first one, a seed fuel loading using the Haling principle is designed. In the second stage, the corresponding control rod pattern for the previous fuel loading is obtained. Finally, in the last stage, a new fuel loading is created, starting from the previous fuel loading and using the corresponding set of optimized control rod patterns. For each stage, a different objective function is considered. In order to obtain the decision parameters used in those functions, the CM-PRESTO 3D steady-state reactor core simulator was used. Second and third stages are repeated until an appropriate fuel loading and its control rod pattern are obtained, or a stop criterion is achieved. In all stages, the tabu search optimization technique was used. The QuinalliBT system was tested and applied to a real BWR operation cycle. It was found that the value for k eff obtained by QuinalliBT was 0.0024 Δk/k greater than that of the reference cycle

  3. Sensitivity Analysis of Depletion Parameters for Heat Load Evaluation of PWR Spent Fuel Storage Pool

    International Nuclear Information System (INIS)

    Kim, In Young; Lee, Un Chul

    2011-01-01

    As necessity of safety re-evaluation for spent fuel storage facility has emphasized after the Fukushima accident, accuracy improvement of heat load evaluation has become more important to acquire reliable thermal-hydraulic evaluation results. As groundwork, parametric and sensitivity analyses of various storage conditions for Kori Unit 4 spent fuel storage pool and spent fuel depletion parameters such as axial burnup effect, operation history, and specific heat are conducted using ORIGEN2 code. According to heat load evaluation and parametric sensitivity analyses, decay heat of last discharged fuel comprises maximum 80.42% of total heat load of storage facility and there is a negative correlation between effect of depletion parameters and cooling period. It is determined that specific heat is most influential parameter and operation history is secondly influential parameter. And decay heat of just discharged fuel is varied from 0.34 to 1.66 times of average value and decay heat of 1 year cooled fuel is varied from 0.55 to 1.37 times of average value in accordance with change of specific power. Namely depletion parameters can cause large variation in decay heat calculation of short-term cooled fuel. Therefore application of real operation data instead of user selection value is needed to improve evaluation accuracy. It is expected that these results could be used to improve accuracy of heat load assessment and evaluate uncertainty of calculated heat load.

  4. Fuel load modeling from mensuration attributes in temperate forests in northern Mexico

    Science.gov (United States)

    Maricela Morales-Soto; Marín Pompa-Garcia

    2013-01-01

    The study of fuels is an important factor in defining the vulnerability of ecosystems to forest fires. The aim of this study was to model a dead fuel load based on forest mensuration attributes from forest management inventories. A scatter plot analysis was performed and, from explanatory trends between the variables considered, correlation analysis was carried out...

  5. Spatial and temporal variability of guinea grass (Megathyrsus maximus) fuel loads and moisture on Oahu, Hawaii

    Science.gov (United States)

    Lisa M. Ellsworth; Creighton M. Litton; Andrew D. Taylor; J. Boone Kauffman

    2013-01-01

    Frequent wildfires in tropical landscapes dominated by non-native invasive grasses threaten surrounding ecosystems and developed areas. To better manage fire, accurate estimates of the spatial and temporal variability in fuels are urgently needed. We quantified the spatial variability in live and dead fine fuel loads and moistures at four guinea grass (...

  6. Dwarf mistletoe effects on fuel loadings in ponderosa pine forests in northern Arizona

    Science.gov (United States)

    Chad Hoffman; Robert Mathiasen; Carolyn Hull Sieg

    2007-01-01

    Southwestern dwarf mistletoe (Arceuthobium vaginatum (Willd.) J. Presl ssp. cryptopodum) infests about 0.9 million ha in the southwestern United States. Several studies suggest that dwarf mistletoes affect forest fuels and fire behavior; however, few studies have quantified these effects. We compared surface fuel loadings and...

  7. Realization of an Electronic Load for Testing Low Power PEM Fuel Cells

    Directory of Open Access Journals (Sweden)

    Djordje Šaponjić

    2011-06-01

    Full Text Available A realized electronic load system intended for testing and characterization of hydrogen fuel sells is described. The system is based on microcontroller PIC16F877 by applying the concept of virtual instrumentation. The accomplished accuracy of the developed electronic system allows performing efficiently investigations of the electro-chemical phenomena involved in the process of designing hydrogen fuel cells.

  8. The control system adopted for Super-Phenix. Reasons for choice and evaluation of performance

    International Nuclear Information System (INIS)

    Decuyper, J.; Skull, G.; Hery, M.; Hennebicq, J.P.

    1978-01-01

    The paper reviews all the research done in working out the control system for the fast-neutron Super-Phenix power station, which is now under construction at Creys-Malville, France. The purpose of the system is to provide a balance between the power produced by the reactor and that taken by the electricity-generating plant. After an introductory section on the structure of the power station and the operating conditions imposed, the following main stages in design work are described: development of the system simulation model and corrobaration on the basis of test results; specification of possible control system layouts (i.e. the various possible connections between regulating variables and regulated variables), optimization of control coefficients of each layout, comparison of performance and choice of layout; detailed study of the layout chosen. Special reference is made to the following typical aspects of Super-Phenix operating technology: response of the power station to primary frequency control; stability of steam generators operating in parallel; establishment of the sodium temperature value. The final part is a summary of the research carried out and a description of the performance of the computer codes. (author)

  9. High-uranium-loaded U3O8--Al fuel element development program

    International Nuclear Information System (INIS)

    Martin, M.M.

    1978-01-01

    The High-Uranium-Loaded U 3 O 8 --Al Fuel Development Program supports Argonne National Laboratory efforts to develop high-uranium-density research and test reactor fuel to accommodate use of low-uranium enrichment. The goal is to fuel most research and test reactors with uranium of less than 20% enrichment for the purpose of lowering the potential for diversion of highly-enriched material for nonpeaceful usages

  10. Fatigue analysis of CANFLEX-NU fuel elements subjected to power-cyclic loads

    International Nuclear Information System (INIS)

    Sim, Ki Seob; Suk, Ho Chun.

    1997-08-01

    This report describes the fatigue analysis of the CANDU advanced fuel, so-called CANFLEX-NU, subjected to power-cyclic loads more than 1,000. The CANFLEX-NU bundle is composed of 43 elements with natural uranium fuel. As a result, the CANFLEX-NU fuel elements will maintain good integrity under the condition of 1,500 power-cycles. (author). 4 refs., 19 figs

  11. Expert system for assisting the diagnostic and localisation of breakdowns on the fuel elements loading machine

    International Nuclear Information System (INIS)

    Merlin, J.; Pradal, B.

    1990-01-01

    An expert system is developed in order to minimize the time lost through breakdowns of the fuel loading device. The expert system developed by FRAMATOME uses MAINTEX software. The expert systems MACHA and SEDMAC were designed respectively for use on 1300 MWe and 900 MWe loading machines [fr

  12. Improvement of burnup analysis for pebble bed reactors with an accumulative fuel loading scheme

    International Nuclear Information System (INIS)

    Simanullang, Irwan Liapto; Obara, Toru

    2015-01-01

    Given the limitations of natural uranium resources, innovative nuclear power plant concepts that increase the efficiency of nuclear fuel utilization are needed. The Pebble Bed Reactor (PBR) shows some potential to achieve high efficiency in natural uranium utilization. To simplify the PBR concept, PBR with an accumulation fuel loading scheme was introduced and the Fuel Handling System (FHS) removed. In this concept, the pebble balls are added little by little into the reactor core until the pebble balls reach the top of the reactor core, and all pebble balls are discharged from the core at the end of the operation period. A code based on the MVP/MVP-BURN method has been developed to perform an analysis of a PBR with the accumulative fuel loading scheme. The optimum fuel composition was found using the code for high burnup performance. Previous efforts provided several motivations to improve the burnup performance: First, some errors in the input code were corrected. This correction, and an overall simplification of the input code, was implemented for easier analysis of a PBR with the accumulative fuel loading scheme. Second, the optimum fuel design had been obtained in the infinite geometry. To improve the optimum fuel composition, a parametric survey was obtained by varying the amount of Heavy Metal (HM) uranium per pebble and the degree of uranium enrichment. Moreover, an entire analysis of the parametric survey was obtained in the finite geometry. The results show that improvements in the fuel composition can lead to more accurate analysis with the code. (author)

  13. Skidder load capacity and fuel consumption HP-41C program

    Science.gov (United States)

    Ross A. Phillips

    1983-01-01

    This program gives the log weight that the skidder can move and gives fuel consumption either in liters or gallons per turn. Slope of the skid trail, skidder weight, and skid distance must be entered into the program.

  14. Genetic algorithm for the optimization of the loading pattern for reactor core fuel management

    International Nuclear Information System (INIS)

    Zhou Sheng; Hu Yongming; zheng Wenxiang

    2000-01-01

    The paper discusses the application of a genetic algorithm to the optimization of the loading pattern for in-core fuel management with the NP characteristics. The algorithm develops a matrix model for the fuel assembly loading pattern. The burnable poisons matrix was assigned randomly considering the distributed nature of the poisons. A method based on the traveling salesman problem was used to solve the problem. A integrated code for in-core fuel management was formed by combining this code with a reactor physics code

  15. Comparative analysis of different methods of modelling of most loaded fuel pin in transients

    International Nuclear Information System (INIS)

    Ovdiyenko, Y.; Khalimonchuk, V.; Ieremenko, M.

    2007-01-01

    Different methods of modeling of most loaded fuel pin are presented at the work. Calculation studies are performed on example of accident related to WWER-1000 cluster rod ejection with using of spatial kinetic code DYN3D that uses nodal method to calculate distribution of neutron flux in the core. Three methods of modeling of most loaded fuel pin are considered - flux reconstruction in fuel macrocell, pin-by-pin calculation by using of DYN3D/DERAB package and by introducing of additional 'hot channel'. Obtained results of performed studies could be used for development of calculation kinetic models during preparing of safety analysis report (Authors)

  16. Fuel rod behavior of a PWR during load following

    International Nuclear Information System (INIS)

    Perrotta, J.A.; Andrade, G.G. de

    1982-01-01

    The behavior of a PWR fuel rod when operating in normal power cycles, excluding in case of accidents, is analysed. A computer code, that makes the mechanical analysis of the cladding using the finite element method was developed. The ramps and power cycles were simulated suposing the existence of cracks in pellets when the cladding-pellet interaction are done. As a result, an operation procedure of the fuel rod in power cycle is recommended. (E.G.) [pt

  17. An automated optimization of core fuel loading pattern for pressurized water reactors

    International Nuclear Information System (INIS)

    Chen Renji

    1988-11-01

    An optimum method was adopted to search for an optimum fuel loading pattern in pressurized water reactors. A radial power peak factor was chosen as the objective function of the optimum loading. The direct search method with shuffling rules is used to find optimum solution. The search for an optimum loading pattern with the smallest radial power peak by exchanging fuel assemblies was made. The search process is divided into two steps. In the first step fresh fuels or high reactivity fuels are arranged which are placed in core interior to have a reasonable fuel loading pattern. To further reduce the radial power peak factor, the second step will be necessary to rearrange the exposed lower reactivity fuel around the assemblies which has the radial power peak. In optimum process 1.5 group coarse mesh diffusion theory or two group nodal Green function diffusion theory is utilized to calculate the two dimensional power distribution after each shuffle. Also, above two methods are combinatively utilized to calculate the state quantity. It is not only true to save CPU time, but also can obtian exact results. Besides above function, the code MSOFEL is used to search critical boron concentration and to predict burn-up. The code has been written with FORTRAN-4. The optimum loading pattern was chosen for OCONEE and QINSHAN nuclear power plants as reference examples. The validity and feasibility of MSOFEL was demonstrated

  18. Validation of the Nuclear Design Method for MOX Fuel Loaded LWR Cores

    International Nuclear Information System (INIS)

    Saji, E.; Inoue, Y.; Mori, M.; Ushio, T.

    2001-01-01

    The actual batch loading of mixed-oxide (MOX) fuel in light water reactors (LWRs) is now ready to start in Japan. One of the efforts that have been devoted to realizing this batch loading has been validation of the nuclear design methods calculating the MOX-fuel-loaded LWR core characteristics. This paper summarizes the validation work for the applicability of the CASMO-4/SIMULATE-3 in-core fuel management code system to MOX-fuel-loaded LWR cores. This code system is widely used by a number of electric power companies for the core management of their commercial LWRs. The validation work was performed for both boiling water reactor (BWR) and pressurized water reactor (PWR) applications. Each validation consists of two parts: analyses of critical experiments and core tracking calculations of operating plants. For the critical experiments, we have chosen a series of experiments known as the VENUS International Program (VIP), which was performed at the SCK/CEN MOL laboratory in Belgium. VIP consists of both BWR and PWR fuel assembly configurations. As for the core tracking calculations, the operating data of MOX-fuel-loaded BWR and PWR cores in Europe have been utilized

  19. Optimization programs for reactor core fuel loading exhibiting reduced neutron leakage

    International Nuclear Information System (INIS)

    Darilek, P.

    1991-01-01

    The program MAXIM was developed for the optimization of the fuel loading of WWER-440 reactors. It enables the reactor core reactivity to be maximized by modifying the arrangement of the fuel assemblies. The procedure is divided into three steps. The first step includes the passage from the three-dimensional model of the reactor core to the two-dimensional model. In the second step, the solution to the problem is sought assuming that the multiplying properties, or the reactivity in the zones of the core, vary continuously. In the third step, parameters of actual fuel assemblies are inserted in the ''continuous'' solution obtained. Combined with the program PROPAL for a detailed refinement of the loading, the program MAXIM forms a basis for the development of programs for the optimization of fuel loading with burnable poisons. (Z.M.). 16 refs

  20. Completion of UO2 pellets production and fuel rods load for the RA-8 critical facility

    International Nuclear Information System (INIS)

    Marajofsky, Adolfo; Perez, Lidia E.; Thern, Gerardo G.; Altamirano, Jorge S.; Benitez, Ana M.; Cardenas, Hugo R.; Becerra, Fabian A.; Perez, Aldo E.; Fuente, Mariano de la

    1999-01-01

    The Advanced Fuels Division produced fuel pellets of 235 U with 1.8% and 3.6% enrichment and Zry-4 cladding loads for the RA-8 reactor at Pilcaniyeu Technological Unit. For economical and availability reasons, the powder acquired was initially UO 2 with 3.4% enrichment in 235 U, therefore the 235 U powder with 1.8% enrichment was produced by mechanical mixture. The production of fuel pellets for both enrichments was carried out by cold pressing and sintering processes in reducing atmosphere. The load of Zry-4 claddings was performed manually. The production stages can be divided into setup, qualification and production. This production allows not only to fulfill satisfactorily the new fuel rods supply for the RA-8 reactor but also to count with a new equipment and skilled personnel as well as to meet quality and assurance control methods for future pilot-scale production and even new fuel elements production. (author)

  1. Development and validation of a nuclear data and calculation system for Superphenix with steel reflectors; Developpement et qualification d`un formulaire adapte a superphenix avec reflecteurs

    Energy Technology Data Exchange (ETDEWEB)

    Bosq, J Ch

    1998-11-09

    This thesis concerns the definition and the validation of the ERANOS neutronic calculation system for steel reflected fast reactors. The calculation system uses JEF2.2 evaluated nuclear data, the ECCO cell code and the BISTRO and VARIANT transport codes. After a description of the physical phenomena induced by the existence of the these sub-critical media, an inventory of the past studies related to steel reflectors is reported. A calculational scheme taking into account the important physical phenomena (strong neutronic slowing-down, presence of broad resonances of the structural materials and spatial variation of the spectrum in the reflector) is defined. This method is validated with the TRIPOLI4 reference Monte-Carlo code. The use of this upgraded calculation method for the analysis of the part of the CIRANO experimental program devoted to the study of steel reflected configurations leads to discrepancies between the calculated and measured values. These remaining discrepancies obtained for the reactivity and the fission rate traverses are due to inaccurate nuclear data for the structural materials. The adjustment of these nuclear data in order to reduce these discrepancies id demonstrated. The additional uncertainty associated to the integral parameters of interest for a nuclear reactor (reactivity and power distribution) induced by the replacement of a fertile blanket by a steel reflector is determined for the Superphenix reactor and is proved to be small. (author) 86 refs.

  2. Canopy Fuel Load Mapping of Mediterranean Pine Sites Based on Individual Tree-Crown Delineation

    Directory of Open Access Journals (Sweden)

    Giorgos Mallinis

    2013-12-01

    Full Text Available This study presents an individual tree-crown-based approach for canopy fuel load estimation and mapping in two Mediterranean pine stands. Based on destructive sampling, an allometric equation was developed for the estimation of crown fuel weight considering only pine crown width, a tree characteristic that can be estimated from passive imagery. Two high resolution images were used originally for discriminating Aleppo and Calabrian pines crown regions through a geographic object based image analysis approach. Subsequently, the crown region images were segmented using a watershed segmentation algorithm and crown width was extracted. The overall accuracy of the tree crown isolation expressed through a perfect match between the reference and the delineated crowns was 34.00% for the Kassandra site and 48.11% for the Thessaloniki site, while the coefficient of determination between the ground measured and the satellite extracted crown width was 0.5. Canopy fuel load values estimated in the current study presented mean values from 1.29 ± 0.6 to 1.65 ± 0.7 kg/m2 similar to other conifers worldwide. Despite the modest accuracies attained in this first study of individual tree crown fuel load mapping, the combination of the allometric equations with satellite-based extracted crown width information, can contribute to the spatially explicit mapping of canopy fuel load in Mediterranean areas. These maps can be used among others in fire behavior prediction, in fuel reduction treatments prioritization and during active fire suppression.

  3. Analytical Dancoff factor evaluations for reactor designs loaded with TRISO particle fuel

    International Nuclear Information System (INIS)

    Ji, Wei; Liang, Chao; Pusateri, Elise N.

    2014-01-01

    Highlights: • The Dancoff factors for randomly distributed TRISO fuel particles are evaluated. • A new “dual-sphere” model is proposed to predict Dancoff factors. • The new model accurately accounts for the coating regions of fuel particles. • High accuracy is achieved over a broad range of design parameters. • The new model can be used to analyze reactors with double heterogeneity. - Abstract: A new mathematical model, the dual-sphere model, is proposed to analytically evaluate Dancoff factors of TRISO fuel kernels based on the chord method. The accurate evaluation of fuel kernel Dancoff factors is needed when one analyzes nuclear reactors loaded with TRISO particle fuel. In these reactor designs, fuel kernels are randomly distributed and shield each other, causing a shadowing effect. The Dancoff factor is a quantitative measure of this effect and is determined by the spatial distribution of fuel kernels. A TRISO fuel particle usually consists of four layers that form a coating region outside the fuel kernel. When fuel particles are loaded in the reactor, the spatial distribution of fuel kernels can be affected by the thickness of the coating region. Therefore, the coating region should be taken into account in the calculation of Dancoff factors. However, the previous model, the single-sphere model, assumes no coating regions in the Dancoff factor predictions. To address this model deficiency, the dual-sphere model is proposed by deriving a new chord length distribution function between two fuel kernels that explicitly accounts for coating regions. The new model is employed to derive analytical solutions of infinite medium, intra-fuel pebble and intra-fuel compact/pin Dancoff factors over a wide range of volume packing fractions of TRISO fuel particles, varying from 2% to 60%. Comparisons are made with the predictions from the single-sphere model and reference Monte Carlo simulations. A significant improvement of the accuracy, over the ranges of

  4. Backfitting Superphenix

    International Nuclear Information System (INIS)

    Montane, C.; Ballenberger, L.

    1993-01-01

    The fast breeder reactor with an installed power of 1242 MWe on the site of Creys-Malville is down for backfitting activities in the galleries enclosing the four secondary systems in order to ensure nuclear safety under conditions of a very large sodium spray fire. The measures carried out comprise preventive fire protection measures and steps limiting the consequences of such a fire. The nuclear generating unit is scheduled for recommissioning in summer 1994. (orig.) [de

  5. Report of Inquiry Commission (1) on Superphenix and the fast neutron reactor system. Vol. 2. Hearings

    International Nuclear Information System (INIS)

    Galley, Robert; Bataille, Christian

    1998-01-01

    This document is a two-volume report, made on behalf of the Inquiry Commission of French National Assembly, concerning the issue of Superphenix and the fast neutron reactor system. The first volume contains the report while the second presents the accounts of 27 hearings in the Inquiry Commission. Questions concerning the technical aspects, costs of decommissioning operations, environment and social impacts, etc, are addressed and discussed with officials implied in nuclear safety, environment protection, science and technology, trade unions, education, atomic energy agency, military applications, industry and commerce. The conclusions drawn from these hearings were synthesized in the volume one of the report submitted to the French National Assembly by the Inquiry Commission

  6. Inter-vessels in-service inspection of Super-Phenix

    International Nuclear Information System (INIS)

    Asty, M.; Saglio, R.; Viard, J.; Lerat, B.

    1984-01-01

    The vessels design of fast breeder reactor Super-Phenix enables inspection during operating time. A self-moving machine -MIR- has been built up especially for that purpose. It is able to carry out visual and ultrasonorous inspection. MIR structure is that of a tetrahedron, all tops of which are fitted with two wheels, as for traction and direction. The wheels are leaning on booth the two vessels. Thanks to a computer-assisted control system, MIR is able to move along in every part of the inter-vessels space. Studies have been carried on at the French Commissariat a l'Energie Atomique, by two Sections of the advanced technologies Service. After outlining MIR working conditions, its main characteristics are described [fr

  7. Direct methanol feed fuel cell with reduced catalyst loading

    Science.gov (United States)

    Kindler, Andrew (Inventor)

    1999-01-01

    Improvements to direct feed methanol fuel cells include new protocols for component formation. Catalyst-water repellent material is applied in formation of electrodes and sintered before application of ionomer. A membrane used in formation of an electrode assembly is specially pre-treated to improve bonding between catalyst and membrane. The improved electrode and the pre-treated membrane are assembled into a membrane electrode assembly.

  8. Departure fuel loads in time-minimizing migrating birds can be explained by the energy costs of being heavy

    NARCIS (Netherlands)

    Klaassen, M.R.J.; Lindstrom, A.

    1996-01-01

    Lindstrom & Alerstam (1992 Am. Nat. 140, 477-491) presented a model that predicts optimal departure fuel loads as a function of the rate of fuel deposition in time-minimizing migrants. The basis of the model is that the coverable distance per unit of fuel deposited, diminishes with increasing fuel

  9. Use of Pd-Pt loaded graphene aerogel on nickel foam in direct ethanol fuel cell

    Science.gov (United States)

    Tsang, Chi Him A.; Leung, D. Y. C.

    2018-01-01

    A size customized binder-free bimetallic Pd-Pt loaded graphene aerogel deposited on nickel foam plate (Pd-Pt/GA/NFP) was prepared and used as an electrode for an alkaline direct ethanol fuel cell (DEFC) under room temperature. The effect of fuel concentration and metal composition on the output power density of the DEFC was systematically investigated. Under the optimum fuel concentration, the cell could achieve a value of 3.6 mW cm-2 at room temperature for the graphene electrode with Pd/Pt ratio approaching 1:1. Such results demonstrated the possibility of producing a size customized metal loaded GA/NFP electrode for fuel cell with high performance.

  10. Determination of thermal reactivity coefficients for the first fuel loading of MO34

    International Nuclear Information System (INIS)

    Lueley, J.; Vrban, B.; Farkas, G.; Hascik, J.; Hinca, R.; Petriska, M.; Slugen, V.

    2012-01-01

    The article introduces determination of thermal reactivity coefficients, especially summarized (isothermal) and moderator (density) reactivity coefficients between 200 grad C and 260 grad C with 2 grad C step, - in compliance with the assignment - for the first fuel loading into the RC of NP Mochovce units using 2 nd generation fuel during the start-up using calculation code MCNP5 1.60. (authors)

  11. Preliminary analysis of a large 1600 MWe PWR core loaded with 30% MOX fuel

    International Nuclear Information System (INIS)

    Polidoro, Franco; Corsetti, Edoardo; Vimercati, Giuliano

    2011-01-01

    The paper presents a full-core 3-D analysis of the performances of a large 1600 MWe PWR core, loaded with 30% MOX fuel, in accordance with the European Utility Requirements (EUR). These requirements state that the European next generation power plants have to be designed capable to use MOX (UO 2 - PuO 2 ) fuel assemblies up to 50% of the core, together with UO 2 fuel assemblies. The use of MOX assemblies has a significant impact on key physic parameters and on safety. A lot of studies have been carried out in the past to explore the feasibility of plutonium recycling strategies by loading LWR reactors with MOX fuel. Many of these works were based on lattice codes, in order to perform detailed analyses of the neutronic characteristics of MOX assemblies. With the aim to take into account their interaction with surrounding UO 2 fuel elements, and the global effects on the core at operational conditions, an integrated approach making use of a 3-D core simulation is required. In this light, the present study adopts the state-of-art numerical models CASMO-5 and SIMULATE-3 to analyze the behavior of the core fueled with 30% MOX and to compare it with that of a large PWR reference core, fueled with UO 2 . (author)

  12. Impact of nuclear library difference on neutronic characteristics of thorium-loaded light water reactor fuel

    International Nuclear Information System (INIS)

    Unesaki, H.; Isaka, S.; Nakagome, Y.

    2006-01-01

    Impact of nuclear library difference on neutronic characteristics of thorium-loaded light water reactor fuel is investigated through cell burnup calculations using SRAC code system. Comparison of k ∞ and nuclide composition was made between the results obtained by JENDL-3.3, ENDF/B-VI.8 and JEFF3.0 for (U, Th)O 2 fuels as well as UO 2 fuels, with special interest on the burnup dependence of the neutronic characteristics. The impact of nuclear data library difference on k ∞ of (U, Th)O 2 fuels was found to be significantly large compared to that of UO 2 fuels. Notable difference was also found in nuclide concentration of TRU nuclides. (authors)

  13. Criticality safety study of dry spent fuel cask loaded with increased enrichment fuel

    International Nuclear Information System (INIS)

    Bznuni, S.; Baghdasaryan, N.; Amirjanyan, A.

    2013-01-01

    Existing Dry Spent Fuel Casks (DSC) for transporting and storing of Armenian NPP fuel was licensed for WWER-440 fuel assemblies with 3.6% enrichment. Having in mind that ANPP introduced new fuel assemblies with increased enrichment (3.82 %) re-assessment of criticality safety analysis for DSC is required. Criticality safety analysis of DSC was performed by KENO-VI program using 238-GROUP ENDF/B-VII.0 LIBRARY (V7-238). Results of analysis showed that additional 8 borated racks for fuel assemblies should be included in the design of DSC. In addition feasibility study was performed to find out level of burnup-credit approach implementation to keep current design of DSC unchanged. Burnup-credit analysis was performed by STARBUCS program using axial burnup profiles from Armenian NPP neutronics analysis carried out by BIPR code. (authors)

  14. Apparatus for loading fuel rods into grids of nuclear fuel assemblies

    International Nuclear Information System (INIS)

    Shallenberger, J.M.; Ferlan, S.J.

    1989-01-01

    For use with a nuclear fuel assembly including support grids having cells for receiving fuel rods and with detents disposed within the respective cells for resiliently engaging and laterally supporting the fuel rods received therein, an apparatus is described for facilitating scratchless insertion of each fuel rod into cells of the support rids. The apparatus consists of: a thin-walled metallic tubular member which is long enough to extend through at least a majority of support grids, and is positionable so as to have its thin wall interposed, during insertion of each fuel rod, between the latter and the detents within the cells receiving it, the thin-walled tubular member having a substantially uniform wall thickness of not more than about 0.008 inch, an as-formed inner diameter substantially equal to the outer diameter of the fuel rod, and a longitudinal slit formed in the wall of the tubular member so as to render the wall resiliently deflectable in a diameter-reducing sense, the longitudinal slit having a width sufficient to preclude overlapping of the edges of the wall along the slit, and insufficient for any of the detents to enter the slit when the wall of the tubular member is in position between the detents and the fuel rod

  15. Fire frequency effects on fuel loadings in pine-oak forests of the Madrean Province

    Science.gov (United States)

    Francisco J. Escobedo; Peter F. Ffolliott; Gerald J. Gottfried; Florentino Garza

    2001-01-01

    Loadings of downed woody fuels in pine-oak forests of the Madrean Province are heavier on sites in southeastern Arizona with low fire frequencies and lower on sites in northeastern Sonora, Mexico, with high fire frequencies. Low fire frequencies in southeastern Arizona are attributed largely to past land uses and the fire suppression policies of land management...

  16. Effects of prescribed burning on vegetation and fuel loading in three east Texas state parks

    Science.gov (United States)

    Sandra Rideout; Brian P. Oswald

    2002-01-01

    This study was conducted to evaluate the initial effectiveness of prescribed burning in the ecological restoration of forests within selected parks in east Texas. Twenty-four permanent plots were installed to monitor fuel loads, overstory, sapling, seedling, shrub and herbaceous layers within burn and control units of Mission Tejas, Tyler and Village Creek state parks...

  17. Loading pattern optimization with maximum utilization of discharging fuel employing adaptively constrained discontinuous penalty function

    International Nuclear Information System (INIS)

    Park, T. K.; Joo, H. G.; Kim, C. H.

    2010-01-01

    In order to find the most economical loading pattern (LP) considering multi-cycle fuel loading, multi-objective fuel LP optimization problems are examined by employing an adaptively constrained discontinuous penalty function (ACDPF) method. This is an improved method to simplify the complicated acceptance logic of the original DPF method in that the stochastic effects caused by the different random number sequence can be reduced. The effectiveness of the multi-objective simulated annealing (SA) algorithm employing ACDPF is examined for the reload core LP of Cycle 4 of Yonggwang Nuclear Unit 4. Several optimization runs are performed with different numbers of objectives consisting of cycle length and average burnup of fuels to be discharged or reloaded. The candidate LPs obtained from the multi-objective optimization runs turn out to be better than the reference LP in the aspects of cycle length and utilization of given fuels. It is note that the proposed ACDPF based MOSA algorithm can be a practical method to obtain an economical LP considering multi-cycle fuel loading. (authors)

  18. A load factor based mean-variance analysis for fuel diversification

    Energy Technology Data Exchange (ETDEWEB)

    Gotham, Douglas; Preckel, Paul; Ruangpattana, Suriya [State Utility Forecasting Group, Purdue University, West Lafayette, IN (United States); Muthuraman, Kumar [McCombs School of Business, University of Texas, Austin, TX (United States); Rardin, Ronald [Department of Industrial Engineering, University of Arkansas, Fayetteville, AR (United States)

    2009-03-15

    Fuel diversification implies the selection of a mix of generation technologies for long-term electricity generation. The goal is to strike a good balance between reduced costs and reduced risk. The method of analysis that has been advocated and adopted for such studies is the mean-variance portfolio analysis pioneered by Markowitz (Markowitz, H., 1952. Portfolio selection. Journal of Finance 7(1) 77-91). However the standard mean-variance methodology, does not account for the ability of various fuels/technologies to adapt to varying loads. Such analysis often provides results that are easily dismissed by regulators and practitioners as unacceptable, since load cycles play critical roles in fuel selection. To account for such issues and still retain the convenience and elegance of the mean-variance approach, we propose a variant of the mean-variance analysis using the decomposition of the load into various types and utilizing the load factors of each load type. We also illustrate the approach using data for the state of Indiana and demonstrate the ability of the model in providing useful insights. (author)

  19. Sudden oak death-caused changes to surface fuel loading and potential fire behavior in Douglas-fir-tanoak forests

    Science.gov (United States)

    Y.S. Valachovic; C.A. Lee; H. Scanlon; J.M. Varner; R. Glebocki; B.D. Graham; D.M. Rizzo

    2011-01-01

    We compared stand structure and fuel loading in northwestern California forests invaded by Phytophthora ramorum, the cause of sudden oak death, to assess whether the continued presence of this pathogen alters surface fuel loading and potential fire behavior in ways that may encumber future firefighting response. To attempt to account for these...

  20. Fuel assemblies mechanical behaviour improvements based on design changes and loading patterns computational analyses

    International Nuclear Information System (INIS)

    Marin, J.; Aullo, M.; Gutierrez, E.

    2001-01-01

    In the past few years, incomplete RCCA insertion events (IRI) have been taking place at some nuclear plants. Large guide thimble distortion caused by high compressive loads together with the irradiation induced material creep and growth, is considered as the primary cause of those events. This disturbing phenomenon is worsened when some fuel assemblies are deformed to the extent that they push the neighbouring fuel assemblies and the distortion is transmitted along the core. In order to better understand this mechanism, ENUSA has developed a methodology based on finite element core simulation to enable assessments on the propensity of a given core loading pattern to propagate the distortion along the core. At the same time, the core loading pattern could be decided interacting with nuclear design to obtain the optimum response under both, nuclear and mechanical point of views, with the objective of progressively attenuating the core distortion. (author)

  1. New load cycling strategy for enhanced durability of high temperature proton exchange membrane fuel cell

    DEFF Research Database (Denmark)

    Thomas, Sobi; Jeppesen, Christian; Steenberg, Thomas

    2017-01-01

    The objective of this paper is to develop a new operational strategy to increase the lifetime of a high temperature proton exchange membrane (HT-PEMFCs) fuel cell system by using load cycling patterns to reduce the phosphoric acid loss from the fuel cell. Four single cells were operated under.......8 Acm-2 for the higher end, were selected for the load cycling operation. The relaxation time, which is the period of time spent at low current density operation, is varied to understand how the performance over prolonged period behaves. The duration of the high current density operation is selected...... based on the relaxation time in order to have the same average current density of (0.55 Acm-2 ) for all the cells. Cell 5, with a relaxation time of 2 min performs best and shows lower degradation rate of 36 μVh-1 compared to other load cycling cells with smaller relaxation times. The cell operated...

  2. Sources of variance in BC mass measurements from a small marine engine: Influence of the instruments, fuels and loads

    Science.gov (United States)

    Jiang, Yu; Yang, Jiacheng; Gagné, Stéphanie; Chan, Tak W.; Thomson, Kevin; Fofie, Emmanuel; Cary, Robert A.; Rutherford, Dan; Comer, Bryan; Swanson, Jacob; Lin, Yue; Van Rooy, Paul; Asa-Awuku, Akua; Jung, Heejung; Barsanti, Kelley; Karavalakis, Georgios; Cocker, David; Durbin, Thomas D.; Miller, J. Wayne; Johnson, Kent C.

    2018-06-01

    Knowledge of black carbon (BC) emission factors from ships is important from human health and environmental perspectives. A study of instruments measuring BC and fuels typically used in marine operation was carried out on a small marine engine. Six analytical methods measured the BC emissions in the exhaust of the marine engine operated at two load points (25% and 75%) while burning one of three fuels: a distillate marine (DMA), a low sulfur, residual marine (RMB-30) and a high-sulfur residual marine (RMG-380). The average emission factors with all instruments increased from 0.08 to 1.88 gBC/kg fuel in going from 25 to 75% load. An analysis of variance (ANOVA) tested BC emissions against instrument, load, and combined fuel properties and showed that both engine load and fuels had a statistically significant impact on BC emission factors. While BC emissions were impacted by the fuels used, none of the fuel properties investigated (sulfur content, viscosity, carbon residue and CCAI) was a primary driver for BC emissions. Of the two residual fuels, RMB-30 with the lower sulfur content, lower viscosity and lower residual carbon, had the highest BC emission factors. BC emission factors determined with the different instruments showed a good correlation with the PAS values with correlation coefficients R2 >0.95. A key finding of this research is the relative BC measured values were mostly independent of load and fuel, except for some instruments in certain fuel and load combinations.

  3. Data acquisition and management system for Superphenix fuel element, manufacturing and inspection results

    International Nuclear Information System (INIS)

    Beche, M.; Bertothy, G.; Nougues, B.; Oswald, M.

    1982-04-01

    The system was designed within the scope of our quality assurance organization. The underlying concepts are: 1) traceability - each product or component can be identified and located at any time during the fabrication process and all relevant background information can be called up when required; 2) quality assessment - this concept covers all operations designed to evaluate the level of product compliance with specifications and tolerance limits; 3) product management - this system is a significant aid in assuring an accurate and continually updated product inventory at all stages of the fabrication process (input, process operations, scraps and wastes, retention etc...)

  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. Full Load Performance of a Spark Ignition Engine Fueled with Gasoline-Isobutanol Blends

    Directory of Open Access Journals (Sweden)

    Adrian Irimescu

    2009-10-01

    Full Text Available With fossil fuels reserves coming ever closer to depletion and the issue of air pollution caused by automotive transport becoming more and more important, mankind has looked for various solutions in the field of internal combustion engines. One of these solutions is using biofuels, and while the internal combustion engine will most likely disappear along with the last fossil fuel source, studying biofuels and their impact on automotive power-trains is a necessity even if only on a the short term basis. While engines built to run on alcohol-gasoline blends offer good performance levels even at high concentrations of alcohol, unmodified engines fueled with blends of biofuels and fossil fuels can exhibit a drop in power. The object of this study is evaluating such phenomena when a spark ignition engine is operated at full load.

  6. The Analysis of the Effect of Coolant Channel Width on Fuel Loading of the RSG-GAS Core

    International Nuclear Information System (INIS)

    Surbakti; Tukiran

    2004-01-01

    The RGS-GAS using uranium silicide fuel, plate type and 250 g U of loading is planned to increase the fuel loading to 300 g U even to 400 g U. The silicide fuel has advantages when increase the fuel loading in the same volume. Because of that case, it is necessary to analyze the effect of coolant channel width on fuel loading of the RSG-GAS core. Analyzing the effect the work which done is to generate cell and core calculation using WIMSD/4 and Batan-2DIFF codes. The WIMSD/4 code is used to generate cross section of core material and Batan-2DIFF is used to calculate the effective multiplication factor. The model that used in this calculation there are three kind of fuel loading namely, 250 g U, 300 g U and 400 g U. The coolant channel width is simulated from 1.75 mm to 2.55 mm. From that fuel loadings, it is analyzed which coolant channel width gave the best effective multiplication factor. From result of analysis showed that the best effective multiplication factor is on the coolant channel width of 2.55 mm for third of fuel loadings. (author)

  7. Analysis of proton exchange membrane fuel cell catalyst layers for reduction of platinum loading at Nissan

    International Nuclear Information System (INIS)

    Ohma, Atsushi; Mashio, Tetsuya; Sato, Kazuyuki; Iden, Hiroshi; Ono, Yoshitaka; Sakai, Kei; Akizuki, Ken; Takaichi, Satoshi; Shinohara, Kazuhiko

    2011-01-01

    The biggest issue that must be addressed in promoting widespread use of fuel cell vehicles (FCVs) is to reduce the cost of the fuel cell system. Especially, it is of vital importance to reduce platinum (Pt) loading of catalyst layers (CLs) in the membrane electrode assembly (MEA) of a proton exchange membrane fuel cell (PEMFC). In order to lower the Pt loading of the MEA, mass transport of reactants related to the performance in high current density should be enhanced significantly as well as kinetics of the catalyst, which can result in the better Pt utilization and effectiveness. In this study, we summarized our analytical approach and methods for reduction of Pt loading in CLs. Microstructure, mass transport properties of the reactants, and their relation in CLs were elucidated by applying experimental analyses and computational methods. A simple CL model for I–V performance prediction was then established, where experimentally elucidated parameters of the microstructure and the properties in CLs were taken into account. Finally, we revealed the impact of lowering the Pt loading on the transport properties, polarization, and the I–V performance.

  8. Increasing the flexibility of base-load generating units in operation on fossil fuel

    Energy Technology Data Exchange (ETDEWEB)

    Girshfel' d, V Ya; Khanaev, V A; Volkova, E D; Gorelov, V A; Gershenkroi, M L

    1979-01-01

    Increasing the flexibility of base-load generating units operating on fossil fuel by modifying them is a necessary measure. The highest economic effect is attained with modification of gas- and oil-fired generating units in the Western United Power Systems of the European part of the SPSS. On the basis of available experience, 150- and 200-MW units can be extensively used to regulate the power in the European part of the SPSS through putting them into reserve for the hours of the load dip at night. The change under favorable conditions of 150- and 200-MW units operating on coal to a district-heating operating mode does not reduce the possibilities for flexible operation of these units because it is possible greatly to unload the turbines while the minimum load level of the pulverized fuel fired boiler is retained through transferring a part of the heat load to the desuperheater. It is necessary to accumulate and analyze experience with operation of generating units (especially of supercritical units) with regular shutdowns and starts of groups of units and to solve the problems of modification of generating units, with differentiation with respect to types of fuel and to the united power supply system.

  9. An Electronic Measurement Instrumentation of the Impedance of a Loaded Fuel Cell or Battery.

    Science.gov (United States)

    Aglzim, El-Hassane; Rouane, Amar; El-Moznine, Reddad

    2007-10-17

    In this paper we present an inexpensive electronic measurement instrumentationdeveloped in our laboratory, to measure and plot the impedance of a loaded fuel cell orbattery. Impedance measurements were taken by using the load modulation method. Thisinstrumentation has been developed around a VXI system stand which controls electroniccards. Software under Hpvee ® was developed for automatic measurements and the layout ofthe impedance of the fuel cell on load. The measurement environment, like the ambienttemperature, the fuel cell temperature, the level of the hydrogen, etc..., were taken withseveral sensors that enable us to control the measurement. To filter the noise and theinfluence of the 50Hz, we have implemented a synchronous detection which filters in a verynarrow way around the useful signal. The theoretical result obtained by a simulation underPspice ® of the method used consolidates the choice of this method and the possibility ofobtaining correct and exploitable results. The experimental results are preliminary results ona 12V vehicle battery, having an inrush current of 330A and a capacity of 40Ah (impedancemeasurements on a fuel cell are in progress, and will be the subject of a forthcoming paper).The results were plotted at various nominal voltages of the battery (12.7V, 10V, 8V and 5V)and with two imposed currents (0.6A and 4A). The Nyquist diagram resulting from theexperimental data enable us to show an influence of the load of the battery on its internalimpedance. The similitude in the graph form and in order of magnitude of the valuesobtained (both theoretical and practical) enables us to validate our electronic measurementinstrumentation. One of the future uses for this instrumentation is to integrate it with several control sensors, on a vehicle as an embedded system to monitor the degradation of fuel cell membranes.

  10. An Electronic Measurement Instrumentation of the Impedance of a Loaded Fuel Cell or Battery

    Directory of Open Access Journals (Sweden)

    Reddad El-Moznine

    2007-10-01

    Full Text Available In this paper we present an inexpensive electronic measurement instrumentationdeveloped in our laboratory, to measure and plot the impedance of a loaded fuel cell orbattery. Impedance measurements were taken by using the load modulation method. Thisinstrumentation has been developed around a VXI system stand which controls electroniccards. Software under Hpvee® was developed for automatic measurements and the layout ofthe impedance of the fuel cell on load. The measurement environment, like the ambienttemperature, the fuel cell temperature, the level of the hydrogen, etc..., were taken withseveral sensors that enable us to control the measurement. To filter the noise and theinfluence of the 50Hz, we have implemented a synchronous detection which filters in a verynarrow way around the useful signal. The theoretical result obtained by a simulation underPspice® of the method used consolidates the choice of this method and the possibility ofobtaining correct and exploitable results. The experimental results are preliminary results ona 12V vehicle battery, having an inrush current of 330A and a capacity of 40Ah (impedancemeasurements on a fuel cell are in progress, and will be the subject of a forthcoming paper.The results were plotted at various nominal voltages of the battery (12.7V, 10V, 8V and 5Vand with two imposed currents (0.6A and 4A. The Nyquist diagram resulting from theexperimental data enable us to show an influence of the load of the battery on its internalimpedance. The similitude in the graph form and in order of magnitude of the valuesobtained (both theoretical and practical enables us to validate our electronic measurementinstrumentation. One of the future uses for this instrumentation is to integrate it with several control sensors, on a vehicle as an embedded system to monitor the degradation of fuel cell membranes.

  11. Large Hybrid Energy Systems for Making Low CO2 Load-Following Power and Synthetic Fuel

    International Nuclear Information System (INIS)

    Cherry, Robert S.; Boardman, Richard D.; Aumeier, Steven

    2012-01-01

    Hybrid energy systems using nuclear heat sources can economically produce load-following electrical power by exploiting the surplus generation capacity available at night or seasonally to make synthetic fuel. Vehicle fuel is the only current energy use large enough to absorb all the energy capacity that might be diverted from the power industry, and its ease of storage obviates problems with discontinuous synfuel production. The potential benefits and challenges of synfuels integration are illustrated by the production of methanol from natural gas (as a source of carbon) using steam from a light water nuclear power reactor which is assumed to be available in accord with a year's worth of power demand data. Methanol's synthesis process is easily adapted to using 300 C heat from a light water reactor and this simple compound can be further processed into gasoline, biodiesel, or dimethyl ether, fuels which can be used with the current vehicle fleet. A supplemental feed to the methanol process of natural gas (for energy) allows operation at constant full rate when the nuclear heat is being used to produce electrical power. The higher capital costs of such a system are offset by a lower cost of heat and power production from a large base load type of plant and by reduced costs associated with much lower CO2 emissions. Other less tangible economic benefits of this and similar hybrid systems include better use of natural resource for fuels and greater energy services security from the domestic production of vehicle fuel.

  12. IMPROVEMENT OF PERFORMANCE OF DUAL FUEL ENGINE OPERATED AT PART LOAD

    Directory of Open Access Journals (Sweden)

    N. Kapilan

    2010-12-01

    Full Text Available Rising petroleum prices, an increasing threat to the environment from exhaust emissions, global warming and the threat of supply instabilities has led to the choice of inedible Mahua oil (MO as one of the main alternative fuels to diesel oil in India. In the present work, MO was converted into biodiesel by transesterification using methanol and sodium hydroxide. The cost of Mahua oil biodiesel (MOB is higher than diesel. Hence liquefied petroleum gas (LPG, which is one of the cheapest gaseous fuels available in India, was fumigated along with the air to reduce the operating cost and to reduce emissions. The dual fuel engine resulted in lower efficiency and higher emissions at part load. Hence in the present work, the injection time was varied and the performance of the dual fuel engine was studied. From the engine tests, it is observed that an advanced injection time results in higher efficiency and lower emissions. Hence, advancing the injection timing is one of the ways of increasing the efficiency of LPG+MOB dual fuel engine operated at part load.

  13. Numerical study of radial stepwise fuel load reshuffling traveling wave reactor

    International Nuclear Information System (INIS)

    Zhang Dalin; Zheng Meiyin; Tian Wenxi; Qiu Suizheng; Su Guanghui

    2015-01-01

    Traveling wave reactor is a new conceptual fast breeder reactor, which can adopt natural uranium, depleted uranium and thorium directly to realize the self sustainable breeding and burning to achieve very high fuel utilization fraction. Based on the mechanism of traveling wave reactor, a concept of radial stepwise fuel load reshuffling traveling wave reactor was proposed for realistic application. It was combined with the typical design of sodium-cooled fast reactors, with which the asymptotic characteristics of the inwards stepwise fuel load reshuffling were studied numerically in two-dimension. The calculated results show that the asymptotic k_e_f_f parabolically varies with the reshuffling cycle length, while the burnup increases linearly. The highest burnup satisfying the reactor critical condition is 38%. The power peak shifts from the fuel discharging zone (core centre) to the fuel uploading zone (core periphery) and correspondingly the power peaking factor decreases along with the reshuffling cycle length. In addition, at the high burnup case the axial power distribution close to the core centre displays the M-shaped deformation. (authors)

  14. Advances in neutronics calculation of fast neutron reactors - Demonstration on Super-Phenix reactor

    International Nuclear Information System (INIS)

    Czernecki, Sebastien

    1998-01-01

    The fast reactor european neutronics calculations system, ERANOS, has integrated recent improvements both in nuclear data, with the use of the adjusted nuclear library ERALIB 1 from the JEF2.2 library, and calculation methods, with the use of the new european cell code, ECCO, and the deterministic code, TGV/VARIANT. This code performs full 3-D reactor calculation in the transport theory with variational method. The aim of this work is to create and validate a new calculational scheme for fast spectrum systems offering good compromise between accuracy and running time. The new scheme is based on these improvements plus a special procedure accounting for control rod heterogeneity, which uses a reactivity equivalence homogenization. The new scheme has been validated by means of experiment/calculation comparisons, using the extensive start-up program measurements performed in Super-Phenix reactor. The validation uses also recent measurements performed in the Phenix reactor. The results are very satisfactory and show a significant improvement for almost all core parameters, especially for critical mass, control rod worth and radial subassembly power distribution. A detailed analysis of the discrepancies between the old scheme and the new one for this parameter allows to understand the separate effects of methods and nuclear data on the radial power distribution shape. (author) [fr

  15. Development and validation of a nuclear data and calculation system for Superphenix with steel reflectors

    International Nuclear Information System (INIS)

    Bosq, J.Ch.

    1998-01-01

    This thesis concerns the definition and the validation of the ERANOS neutronic calculation system for steel reflected fast reactors. The calculation system uses JEF2.2 evaluated nuclear data, the ECCO cell code and the BISTRO and VARIANT transport codes. After a description of the physical phenomena induced by the existence of the these sub-critical media, an inventory of the past studies related to steel reflectors is reported. A calculational scheme taking into account the important physical phenomena (strong neutronic slowing-down, presence of broad resonances of the structural materials and spatial variation of the spectrum in the reflector) is defined. This method is validated with the TRIPOLI4 reference Monte-Carlo code. The use of this upgraded calculation method for the analysis of the part of the CIRANO experimental program devoted to the study of steel reflected configurations leads to discrepancies between the calculated and measured values. These remaining discrepancies obtained for the reactivity and the fission rate traverses are due to inaccurate nuclear data for the structural materials. The adjustment of these nuclear data in order to reduce these discrepancies id demonstrated. The additional uncertainty associated to the integral parameters of interest for a nuclear reactor (reactivity and power distribution) induced by the replacement of a fertile blanket by a steel reflector is determined for the Superphenix reactor and is proved to be small. (author)

  16. Reliability of the spent fuel identification for flask loading procedure used by COGEMA for fuel transport to La Hague

    International Nuclear Information System (INIS)

    Eid, M.; Zachar, M.; Pretesacque, P.

    1991-01-01

    The Spent Fuel Identification for Flask Loading (SFIFL) procedure designed by COGEMA is analysed and its reliability calculated. The reliability of the procedure is defined as the probability of transporting only approved fuel elements for a given number of shipments. The procedure describes a non-coherent system. A non-coherent system is the one in which two successive failures could result in a success, from the system mission point of view. A technique that describes the system with the help of its maximal cuts (states) is used for calculations. A maximal cut contains more than one failure which can split into two cuts (sub-states). Cuts splitting will enable us to analyse, in a systematic way, non-coherent systems with independent basic components. (author)

  17. Reliability of the spent fuel identification for flask loading procedure used by COGEMA for fuel transport to La Hague

    International Nuclear Information System (INIS)

    Eid, M.; Zachar, M.; Pretesacque, P.

    1990-01-01

    The Spent Fuel Identification for Flask Loading, SFIFL, procedure designed by COGEMA is analysed and its reliability is calculated. The reliability of the procedure is defined as the probability of transporting only approved fuel elements for a given number of shipments. The procedure describes a non-coherent system. A non-coherent system is the one in which two successive failures could result in a success, from the system mission point of view. A technique that describes the system with the help of its maximal cuts (states), is used for calculations. A maximal cut contains more than one failure can split into two cuts, (sub-states). Cuts splitting will enable us to analyse, in a systematic way, non-coherent systems with independent basic components. (author)

  18. HTGR fuel development: investigations of breakages of uranium-loaded weak acid resin microspheres

    International Nuclear Information System (INIS)

    Carpenter, J.A. Jr.

    1977-11-01

    During the HTGR fuel development program, a high percentage of uranium-loaded weak acid resin microspheres broke during pneumatic transfer, carbonization, and conversion. One batch had been loaded by the UO 3 method; the other by the ammonia neutralization method. To determine the causes of failure, samples of the two failed batches were investigated by optical microscopy, scanning electron microscopy, electron beam microprobe, and other techniques. Causes of failure are postulated and methods are suggested to prevent recurrence of this kind of failure

  19. High-Uranium-Loaded U3O8-Al fuel element development program. Part 1

    International Nuclear Information System (INIS)

    Martin, M.M.

    1993-01-01

    The High-Uranium-Loaded U 3 O 8 -Al Fuel Element Development Program supports Argonne National Laboratory efforts to develop high-uranium-density research and test reactor fuel to accommodate use of low-uranium enrichment. The goal is to fuel most research and test reactors with uranium of less than 20% enrichment for the purpose of lowering the potential for diversion of highly-enriched material for nonpeaceful usages. The specific objective of the program is to develop the technological and engineering data base for U 3 O 8 -Al plate-type fuel elements of maximal uranium content to the point of vendor qualification for full scale fabrication on a production basis. A program and management plan that details the organization, supporting objectives, schedule, and budget is in place and preparation for fuel and irradiation studies is under way. The current programming envisions a program of about four years duration for an estimated cost of about two million dollars. During the decades of the fifties and sixties, developments at Oak Ridge National Laboratory led to the use of U 3 O 8 -Al plate-type fuel elements in the High Flux Isotope Reactor, Oak Ridge Research Reactor, Puerto Rico Nuclear Center Reactor, and the High Flux Beam Reactor. Most of the developmental information however applies only up to a uranium concentration of about 55 wt % (about 35 vol % U 3 O 8 ). The technical issues that must be addressed to further increase the uranium loading beyond 55 wt % U involve plate fabrication phenomena of voids and dogboning, fuel behavior under long irradiation, and potential for the thermite reaction between U 3 O 8 and aluminum

  20. Impact of Bulldozer's Engine Load Factor on Fuel Consumption, CO2 Emission and Cost

    OpenAIRE

    V. Kecojevic; D. Komljenovic

    2011-01-01

    Problem statement: Bulldozers consume a large amount of diesel fuel and consequently produce a significant quantity of CO2. Environmental and economic cost issues related to fuel consumption and CO2 emission represent a substantial challenge to the mining industry. Approach: Impact of engine load conditions on fuel consumption and the subsequent CO2 emission and cost was analyzed for Caterpillar bulldozers. Results were compared with the data on bulldozers' fuel consu...

  1. Automatic control of load increases power and efficiency in a microbial fuel cell

    Energy Technology Data Exchange (ETDEWEB)

    Premier, Giuliano C.; Kim, Jung Rae; Michie, Iain [Sustainable Environment Research Centre (SERC), Faculty of Advanced Technology, University of Glamorgan, Pontypridd, Mid-Glamorgan CF37 1DL (United Kingdom); Dinsdale, Richard M.; Guwy, Alan J. [Sustainable Environment Research Centre (SERC), Faculty of Health, Sport and Science, University of Glamorgan, Pontypridd, Mid-Glamorgan CF37 1DL (United Kingdom)

    2011-02-15

    Increasing power production and coulombic efficiency (CE) of microbial fuel cells (MFCs) is a common research ambition as the viability of the technology depends to some extent on these measures of performance. As MFCs are typically time varying systems, comparative studies of controlled and un-controlled external load impedance are needed to show if control affects the biocatalyst development and hence MFC performance. The application of logic based control of external load resistance is shown to increase the power generated by the MFC, when compared to an equivalent system which has a static resistive load. The controlled MFC generated 1600 {+-} 400 C, compared to 300 {+-} 10 C with an otherwise replicate fixed load MFC system. The use of a parsimonious gradient based control was able to increase the CE to within the range of 15.1-22.7%, while the CE for a 200 {omega} statically loaded MFC lay in the range 3.3-3.7%. The controlled MFC improves the electrogenic anodic biofilm selection for power production, indicating that greater power and substrate conversion can be achieved by controlling load impedance. Load control ensured sustainable current demand, applied microbial selection pressures and provided near-optimal impedance for power transference, compared to the un-controlled system. (author)

  2. Fuel-assembly behavior under dynamic impact loads due to dry-storage cask mishandling

    International Nuclear Information System (INIS)

    1991-07-01

    Continued operation of nuclear power plants is contingent on the ability to provide adequate storage of spent fuel. Until recently, utilities have been able to maintain interim in-pool spent fuel storage. However, many facilities have reached their capacity and are now faced with shipping their spent fuel in dry casks to alternate storage facilities. The objective of this report is to provide estimates of the structural integrity of irradiated LWR fuel rods subjected to impact loads resulting from postulated cask handling accidents. This is accomplished in five stages: (1) Material properties for irradiated fuel are compiled for use in the structural analyses. (2) Results from parametric analyses of representative assembly designs are used to determine the most limiting case for end and side drop postulated handling accidents. (3) Detailed structural analysis results are presented for these critical designs. The detailed analyses include the coupling of assembly interaction with the cask and cask internals. (4) Criteria for both ultimate stress and brittle fracture failure modes of fuel rod cladding are established. (5) Safe cask handling drop height limits are computed based on items 2 through 4 above. 44 figs., 18 tabs

  3. Fuel element cladding state change mathematical model for a WWER-1000 plant operated in the mode of varying loading

    Directory of Open Access Journals (Sweden)

    S. N. Pelykh

    2010-09-01

    Full Text Available Main features of a fuel element cladding state change mathematical model for a WWER-1000 reactor plant operated in the mode of varying loading are listed. The integrated model is based on the energy creep theory, uses the finite element method for imultaneous solution of the fuel element heat conduction and mechanical deformation equa-tions. Proposed mathematical model allows us to determine the influence of the WWER-1000 regime parameters and fuel assembly design characteristics on the change of cladding properties under different loading conditions of normal operation, as well as the cladding limiting state at variable loading depending on the length, depth and number of cycles.

  4. Using Airborne LIDAR Data for Assessment of Forest Fire Fuel Load Potential

    Science.gov (United States)

    İnan, M.; Bilici, E.; Akay, A. E.

    2017-11-01

    Forest fire incidences are one of the most detrimental disasters that may cause long terms effects on forest ecosystems in many parts of the world. In order to minimize environmental damages of fires on forest ecosystems, the forested areas with high fire risk should be determined so that necessary precaution measurements can be implemented in those areas. Assessment of forest fire fuel load can be used to estimate forest fire risk. In order to estimate fuel load capacity, forestry parameters such as number of trees, tree height, tree diameter, crown diameter, and tree volume should be accurately measured. In recent years, with the advancements in remote sensing technology, it is possible to use airborne LIDAR for data estimation of forestry parameters. In this study, the capabilities of using LIDAR based point cloud data for assessment of the forest fuel load potential was investigated. The research area was chosen in the Istanbul Bentler series of Bahceköy Forest Enterprise Directorate that composed of mixed deciduous forest structure.

  5. On behaviour of fuel elements subject to combined cyclic thermomechanical loads

    International Nuclear Information System (INIS)

    Hsu, T.R.

    1980-01-01

    This paper presents detailed finite element formulations on the kinematic hardening rule of plasticity included in an existing thermoelastoplastic stress analysis code primarily designed to predict the thermomechanical behaviour of nuclear reactor fuel elements. The kinematic hardening rule is considered to be important for structures subject to repeated (or cyclic) loads. The start-up/operation/shut-down and various power excursions in a reactor all can be classified as cyclic loadings. In addition to the shifting of material yield surfaces as usually handled by the kinematic hardening rule, the thermal effect and temperature-dependent material properties have also been included in the present work for the first time. A case study related to an in-reactor experiment on a single fuel element indicated that significantly higher cumulative sheath residual strains after two load cycles was obtained by the present scheme than those calculated by the usual isotropic hardening rule. This observation may alert fuel modellers to select proper hardening rules in their analyses. (orig.)

  6. Substantiation of strength of TVSA-ALPHA fuel assembly under dynamic seismic loads

    International Nuclear Information System (INIS)

    Tutnov, A.; Kiselev, A.; Kiselev, A.; Krutko, E.; Kiselev, I.; Samoilov, O.; Kaydalov, V.

    2009-01-01

    A special place in the substantiation of the safe operation of fuel assemblies is the assessment their operating capability under seismic loads, leading to short-term (several seconds or tens of seconds) the dynamic effects on the reactor core. The level of acceleration of various elements of the reactor installation can be higher than 1,5 g (g - acceleration of gravity) and depends on the height of these elements relatively the ground, which movement causes an earthquake. This dynamic load cause significant deformation of the active zones design element, in particular of the fuel assemblies (FA), which could lead to a contact (or impact) interaction between them. The report presents the results of studies of stress-strain state of FA of TVSA-ALPHA type under the influence of seismic loads of the 8th level on Richter scale using standard approach. According to a normative approach the natural frequencies and modes of FA are calculated in the preliminary stage. The obtained results are conservative from the point of view that in the real FA design the most loaded SG in the middle of the fuel assemblies are made in a combined with mixing grid variant, which are joint by a common rim. This increases the overall carrying capacity of SG as compared with the calculation SG model. It is also necessary to bear in mind that the dynamic (impact) loading the basic mechanical properties of the material may have a significant difference from static (standard) values. This refers in particular to the yield limit, the value of which can be several times higher than specified in the calculation

  7. Control rod ejection accident analysis for a PWR with thorium fuel loading

    Energy Technology Data Exchange (ETDEWEB)

    Da Cruz, D.F. [Nuclear Research and Consultancy Group NRG, Westerduinweg 3, P.O. Box 25, 1755 ZG Petten (Netherlands)

    2010-07-01

    This paper presents the results of 3-D transient analysis of a pressurized water reactor (PWR) core loaded with 100% Th-Pu MOX fuel assemblies. The aim of this study is to evaluate the safety impact of applying a full loading of this innovative fuel in PWRs of the current generation. A reactivity insertion accident scenario has been simulated using the reactor core analysis code PANTHER, used in conjunction with the lattice code WIMS. A single control rod assembly, with the highest reactivity worth, has been considered to be ejected from the core within 100 milliseconds, which may occur due to failure of the casing of the control rod driver mechanism. Analysis at both hot full power and hot zero power reactor states have been taken into account. The results were compared with those obtained for a representative PWR fuelled with UO{sub 2} fuel assemblies. In general the results obtained for both cores were comparable, with some differences associated mainly to the harder neutron spectrum observed for the Th-Pu MOX core, and to some specific core design features. The study has been performed as part of the LWR-DEPUTY project of the EURATOM 6. Framework Programme, where several aspects of novel fuels are being investigated for deep burning of plutonium in existing nuclear power plants. (authors)

  8. Performance of an Active Micro Direct Methanol Fuel Cell Using Reduced Catalyst Loading MEAs

    Directory of Open Access Journals (Sweden)

    D.S. Falcão

    2017-10-01

    Full Text Available The micro direct methanol fuel cell (MicroDMFC is an emergent technology due to its special interest for portable applications. This work presents the results of a set of experiments conducted at room temperature using an active metallic MicroDMFC with an active area of 2.25 cm2. The MicroDMFC uses available commercial materials with low platinum content in order to reduce the overall fuel cell cost. The main goal of this work is to provide useful information to easily design an active MicroDMFC with a good performance recurring to cheaper commercial Membrane Electrode Assemblies MEAs. A performance/cost analysis for each MEA tested is provided. The maximum power output obtained was 18.1 mW/cm2 for a hot-pressed MEA with materials purchased from Quintech with very low catalyst loading (3 mg/cm2 Pt–Ru at anode side and 0.5 mg/cm2 PtB at the cathode side costing around 15 euros. Similar power values are reported in literature for the same type of micro fuel cells working at higher operating temperatures and substantially higher cathode catalyst loadings. Experimental studies using metallic active micro direct methanol fuel cells operating at room temperature are very scarce. The results presented in this work are, therefore, very useful for the scientific community.

  9. Row of fuel assemblies analysis under seismic loading: Modelling and experimental validation

    International Nuclear Information System (INIS)

    Ricciardi, Guillaume; Bellizzi, Sergio; Collard, Bruno; Cochelin, Bruno

    2009-01-01

    The aim of this study was to develop a numerical model for predicting the impact behaviour at fuel assembly level of a whole reactor core under seismic loading conditions. This model was based on a porous medium approach accounting for the dynamics of both the fluid and structure, which interact. The fluid is studied in the whole reactor core domain and each fuel assembly is modelled in the form of a deformable porous medium with a nonlinear constitutive law. The contact between fuel assemblies is modelled in the form of elastic stops, so that the impact forces can be assessed. Simulations were performed to predict the dynamics of a six fuel assemblies row immersed in stagnant water and the whole apparatus was placed on a shaking table mimicking seismic loading conditions. The maximum values of the impact forces predicted by the model were in good agreement with the experimental data. A Proper Orthogonal Decomposition analysis was performed on the numerical data to analyse the mechanical behaviour of the fluid and structure more closely.

  10. Resin-based preparation of HTGR fuels: operation of an engineering-scale uranium loading system

    International Nuclear Information System (INIS)

    Haas, P.A.

    1977-10-01

    The fuel particles for recycle of 233 U to High-Temperature Gas-Cooled Reactors are prepared from uranium-loaded carboxylic acid ion exchange resins which are subsequently carbonized, converted, and refabricated. The development and operation of individual items of equipment and of an integrated system are described for the resin-loading part of the process. This engineering-scale system was full scale with respect to a hot demonstration facility, but was operated with natural uranium. The feed uranium, which consisted of uranyl nitrate solution containing excess nitric acid, was loaded by exchange with resin in the hydrogen form. In order to obtain high loadings, the uranyl nitrate must be acid deficient; therefore, nitric acid was extracted by a liquid organic amine which was regenerated to discharge a NaNO 3 or NH 4 NO 3 solution waste. Water was removed from the uranyl nitrate solution by an evaporator that yielded condensate containing less than 0.5 ppM of uranium. The uranium-loaded resin was washed with condensate and dried to a controlled water content via microwave heating. The loading process was controlled via in-line measurements of the pH and density of the uranyl nitrate. The demonstrated capacity was 1 kg of uranium per hour for either batch loading contractors or a continuous column as the resin loading contractor. Fifty-four batch loading runs were made without a single failure of the process outlined in the chemical flowsheet or any evidence of inability to control the conditions dictated by the flowsheet

  11. EVALUATION OF VIBRATION LOAD ON COMMON RAIL FUEL SYSTEM COMPONENTS FOR DIESEL ENGINE

    Directory of Open Access Journals (Sweden)

    G. M. Kuharonak

    2014-01-01

    Full Text Available The objective of the paper is to develop a program, a methodology and execute vibration load tests of Common Rail fuel system components for a diesel engine. The paper contains an analysis of parameters that characterize vibration activity of research object and determine its applicability as a part of the specific mechanical system. A tests program has been developed that includes measurements of general peak values of vibration acceleration in the fuel system components, transformation of the obtained data while taking into account the fact that peak vibration acceleration values depend on crank-shaft rotation frequency and spectrum of vibration frequency, comparison of these dependences with the threshold limit values obtained in the process of component tests with the help of vibration shaker. The investigations have been carried out in one of the most stressed elements of the Common Rail fuel system that is a RDS 4.2-pressure sensor in a fuel accumulator manufactured by Robert Bosch GmbH and mounted on the MMZ D245.7E4-engines.According to the test methodology measurements have been performed on an engine test bench at all fullload engine curves. Vibration measurements have resulted in time history of the peak vibration acceleration values in three directions from every accelerometer and crank-shaft rotation frequency.It has been proposed to increase a diameter of mounting spacers of the fuel accumulator and install a damping clamp on high pressure tubes from a high pressure fuel pump to the fuel accumulator that permits to reduce a maximum peak vibration acceleration value on the pressure sensor in the fuel accumulator by 400 m/s2 and ensure its application in the given engine.

  12. Fuel assembly loads during a hypothetical blowdown event in a PWR

    International Nuclear Information System (INIS)

    Stabel, J.; Bosanyi, B.; Kim, J.D.

    1991-01-01

    As a consequence of a hypothetical sudden break of the main coolant pipe of a PWR, RPV-internals and fuel assemblies (FA's) are undergoing horizontal and vertical motions. FA's may impact against each other, against core shroud or against lower core support. The corresponding impact loads must be absorbed by the FA spacer grids and guide thimbles. In this paper FA-loads are calculated with and without consideration of Fluid-Structure-Interaction (FSI) effects for assumed different break sizes of the main coolant pipe. The analysis has been performed for a hypothetical cold leg break of a typical SIEMENS-4 loop plant. For this purpose the codes DAPSY/DAISY (GRS, Germany) were coupled with the structural code KWUSTOSS (SIEMENS). It is shown that the FA loads obtained in calculations with consideration of FSI effects are by a factor of 2-4 lower than those obtained in the corresponding calculations without consideration of FSI. (author)

  13. Study on dynamic measurement of fuel pellet length during loading into cladding tube

    International Nuclear Information System (INIS)

    Zhang Kai

    1993-09-01

    Various methods are presented for measuring the pellet length in the cladding tube (zirconium tube) during the loading process of the preparation of single rod of nuclear fuel assembly. These methods are used in former Soviet Union, west European countries and China in the manufacturing of nuclear power plant element. Different methods of dynamic measurement by using mechanics, optics and electricity and their special features are analysed and discussed. The structure and measuring principle of a developed measuring device,and its measuring precision and system deviation are also introduced. Finally, the length of loaded pellets is checked with analog pellets. The results are as expected and show that the method and principle used in the measuring device are feasible. It is an ideal and advanced method for the pellet loading of single cladding tube. The principle mentioned above can also be used in other industries

  14. Feasibility of fully ceramic microencapsulated (FCM) replacement fuel assembly for OPR-1000 core fully loaded with FCM fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Lee, W.J.; Lee, K.H.; Kwon, H.; Chun, J.H.; Kim, Y.M. [Korea Atomic Energy Research Inst., Daejeon (Korea, Republic of); Venneri, F. [Ultra Safe Nuclear Corp., Los Alamos, NM (United States)

    2014-07-01

    The feasibility of replacing conventional UO{sub 2} fuel assemblies (FAs) of light water reactors with accident-tolerant fully ceramic microencapsulated (FCM) FAs has been explored referencing OPR-1000, 1000MW{sub e} PWR. An optimum FCM FA design, 16x16 FCM FA with Silicon Carbide-coated Zircaloy cladding, was selected based on core-level scoping analysis for five FCM FA design candidates screened from FA-level study. For the selected FCM FA design, detailed core following analysis from initial to equilibrium cores, initially fully loaded with the FCM FAs, was carried out to quantify core physics parameters. Using these parameters, the core thermal-hydraulics and coated fuel particle performance of the FCM core was assessed, and the safety margin and accident-tolerance of the FCM core was evaluated for limiting design- and beyond design-basis-accidents. From the study, it has been demonstrated that the FCM fuel is a viable option in replacing the OPR-1000 core with enhanced safety and accident tolerance while maintaining the core neutronics, thermal-hydraulics and mechanical compatibility. (author)

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

    International Nuclear Information System (INIS)

    Jing Xingqing; Yang Yongwei; Gu Yuxiang; Shan Wenzhi

    2001-01-01

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

  16. The evaluation of minimum cooling period for loading of PWR spent nuclear fuel of a dual purpose metal cask

    Energy Technology Data Exchange (ETDEWEB)

    Dho, Ho Seog; Kim, Tae Man; Cho, Chun Hyung [Korea Radioactive Waste Agency, Daejeon (Korea, Republic of)

    2016-12-15

    Recently, because the wet pool storage facilities of NPPs in Korea has become saturated, there has been much active R and D on an interim dry storage system using a transportation and storage cask. Generally, the shielding evaluation for the design of a spent fuel transportation and storage cask is performed by the design basis fuel, which selects the most conservative fuel among the fuels to be loaded into the cask. However, the loading of actual spent fuel into the transportation metal cask is not limited to the design basis fuel used in the shielding evaluation; the loading feasibility of actual spent fuel is determined by the shielding evaluation that considers the characteristics of the initial enrichment, the maximum burnup and the minimum cooling period. This study describes a shielding analysis method for determining the minimum cooling period of spent fuel that meets the domestic transportation standard of the dual purpose metal cask. In particular, the spent fuel of 3.0-4.5wt% initial enrichment, which has a large amount of release, was evaluated by segmented shielding calculations for efficient improvement of the results. The shielding evaluation revealed that about 81% of generated spent fuel from the domestic nuclear power plants until 2008 could be transported by the dual purpose metal cask. The results of this study will be helpful in establishing a technical basis for developing operating procedures for transportation of the dual purpose metal cask.

  17. The evaluation of minimum cooling period for loading of PWR spent nuclear fuel of a dual purpose metal cask

    International Nuclear Information System (INIS)

    Dho, Ho Seog; Kim, Tae Man; Cho, Chun Hyung

    2016-01-01

    Recently, because the wet pool storage facilities of NPPs in Korea has become saturated, there has been much active R and D on an interim dry storage system using a transportation and storage cask. Generally, the shielding evaluation for the design of a spent fuel transportation and storage cask is performed by the design basis fuel, which selects the most conservative fuel among the fuels to be loaded into the cask. However, the loading of actual spent fuel into the transportation metal cask is not limited to the design basis fuel used in the shielding evaluation; the loading feasibility of actual spent fuel is determined by the shielding evaluation that considers the characteristics of the initial enrichment, the maximum burnup and the minimum cooling period. This study describes a shielding analysis method for determining the minimum cooling period of spent fuel that meets the domestic transportation standard of the dual purpose metal cask. In particular, the spent fuel of 3.0-4.5wt% initial enrichment, which has a large amount of release, was evaluated by segmented shielding calculations for efficient improvement of the results. The shielding evaluation revealed that about 81% of generated spent fuel from the domestic nuclear power plants until 2008 could be transported by the dual purpose metal cask. The results of this study will be helpful in establishing a technical basis for developing operating procedures for transportation of the dual purpose metal cask

  18. REFLOS, Fuel Loading and Cost from Burnup and Heavy Atomic Mass Flow Calculation in HWR

    International Nuclear Information System (INIS)

    Boettcher, W.; Schmidt, E.

    1969-01-01

    1 - Nature of physical problem solved: REFLOS is a programme for the evaluation of fuel-loading schemes in heavy water moderated reactors. The problems involved in this study are: a) Burn-up calculation for the reactor cell. b) Determination of reactivity behaviour, power distribution, attainable burn-up for both the running-in period and the equilibrium of a 3-dimensional heterogeneous reactor model; investigation of radial fuel movement schemes. c) Evaluation of mass flows of heavy atoms through the reactor and fuel cycle costs for the running-in, the equilibrium, and the shut down of a power reactor. If the subroutine for treating the reactor cell were replaced by a suitable routine, other reactors with weakly absorbing moderators could be analyzed. 2 - Method of solution: Nuclear constants and isotopic compositions of the different fuels in the reactor are calculated by the cell-burn-up programme and tabulated as functions of the burn-up rate (MWD/T). Starting from a known state of the reactor, the 3-dimensional heterogeneous reactor programme (applying an extension of the technique of Feinberg and Galanin) calculates reactivity and neutron flux distribution using one thermal and one or two fast neutron groups. After a given irradiation time, the new state of the reactor is determined, and new nuclear constants are assigned to the various defined locations in the reactor. Reloading of fuel may occur if the prescribed life of the reactor is reached or if the effective multiplication factor or the power form factor falls below a specified level. The scheme of reloading to be carried out is specified by a load vector, giving the number of channels to be discharged, the kind of movement from one to another channel and the type of fresh fuel to be charged for each single reloading event. After having determined the core states characterizing the equilibrium period, and having decided the fuel reloading scheme for the running-in period of the reactor life, the fuel

  19. Nectar loads as fuel for collecting nectar and pollen in honeybees: adjustment by sugar concentration.

    Science.gov (United States)

    Harano, Ken-Ichi; Nakamura, Jun

    2016-06-01

    When honeybee foragers leave the nest, they receive nectar from nest mates for use as fuel for flight or as binding material to build pollen loads. We examined whether the concentration of nectar carried from the nest changes with the need for sugar. We found that pollen foragers had more-concentrated nectar (61.8 %) than nectar foragers (43.8 %). Further analysis revealed that the sugar concentration of the crop load increased significantly with waggle duration, an indicator of food-source distance, in both groups of foragers. Crop volume also increased with waggle duration. The results support our argument that foragers use concentrated nectar when the need for sugar is high and suggest that they precisely adjust the amount of sugar in the crop by altering both volume and nectar concentrations. We also investigated the impact of the area where foragers receive nectar on the crop load concentration at departure. Although nectar and pollen foragers tend to load nectar at different areas in the nest, area did not have a significant effect on crop load concentration. Departing foragers showed an average of 2.2 momentary (nectar with inappropriate concentrations during these contacts.

  20. Optimization strategies for cask design and container loading in long term spent fuel storage

    International Nuclear Information System (INIS)

    2006-12-01

    As delays are incurred in implementing reprocessing and in planning for geologic repositories, storage of increasing quantities of spent fuel for extended durations is becoming a growing reality. Accordingly, effective management of spent fuel continues to be a priority topic. In response, the IAEA has organized a series of meetings to identify cask loading optimisation issues in preparation for a technical publication on Optimization Strategies for Cask/Container Loading in Long Term Spent Fuel Storage. This publication outlines the optimisation process for cask design, licensing and utilization, describing three principal groups of optimization activities in terms of relevant technical considerations such as criticality, shielding, structural design, operations, maintenance and retrievability. The optimization process for cask design, licensing, and utilization is outlined. The general objectives for the design of storage casks, including storage casks that are intended to be transportable, are summarized. The nature of optimization within the design process is described. The typical regulatory and licensing process is outlined, focusing on the roles of safety regulations, the regulator, and the designer/applicant in the optimization process. Based on the foregoing, a description of the three principal groups of optimization activities is provided. The subsequent chapters of this document then describe the specific optimization activities within these three activity groups, in each of the several design disciplines

  1. A distributed fluid level sensor suitable for monitoring fuel load on board a moving fuel tank

    Science.gov (United States)

    Arkwright, John W.; Parkinson, Luke A.; Papageorgiou, Anthony W.

    2018-02-01

    A temperature insensitive fiber Bragg grating sensing array has been developed for monitoring fluid levels in a moving tank. The sensors are formed from two optical fibers twisted together to form a double helix with pairs of fiber Bragg gratings located above one another at the points where the fibers are vertically disposed. The sensing mechanism is based on a downwards deflection of the section of the double helix containing the FBGs which causes the tension in the upper FBG to decrease and the tension in the lower FBG to increase with concomitant changes in Bragg wavelength in each FBG. Changes in ambient temperature cause a common mode increase in Bragg wavelength, thus monitoring the differential change in wavelength provides a temperature independent measure of the applied pressure. Ambient temperature can be monitored simultaneously by taking the average wavelength of the upper and lower FBGs. The sensors are able to detect variations in pressure with resolutions better than 1 mmH2O and when placed on the bottom of a tank can be used to monitor fluid level based on the recorded pressure. Using an array of these sensors located along the bottom of a moving tank it was possible to monitor the fluid level at multiple points and hence dynamically track the total fluid volume in the tank. The outer surface of the sensing array is formed from a thin continuous Teflon sleeve, making it suitable for monitoring the level of volatile fluids such as aviation fuel and gasoline.

  2. Welding issues associated with design, fabrication and loading of spent fuel storage casks

    International Nuclear Information System (INIS)

    Battige, C.K. Jr.; Howe, A.G.; Sturz, F.C.

    1999-01-01

    The U.S. Nuclear Regulatory Commission (NRC) has observed a number of welding issues associated with design, fabrication, and loading of spent fuel storage casks. These emerging welding-related issues involving a certain dry cask storage system have challenged the safety basis for which NRC approved the casks for storage of spent nuclear fuel. During closure welding, problems have been encountered with cracking. Although the cracks have been attributed to several causes including material suitability, joint restraint and residual stresses, NRC believes hydrogen-induced cracking is the most likely explanation. In light of these cracking events and the potential for flaws in any welding process, NRC sought verification of the corrective actions and the integrity of the lid closure welds before allowing additional casks to be loaded. As a result, the affected utility companies modified the closure welding procedures and developed an acceptable ultrasonic inspection (UT) method. In addition, the casks already loaded at three power reactor sites will require additional non-destructive examinations (NDE) to determine their suitability for continued use. NRC plans to evaluate the generic implications of this issue for current designs and for those in the licensing process. (author)

  3. Effect of fuel assembly when changing from AFA 2G to AFA 3G on seismic loads of reactor internal

    International Nuclear Information System (INIS)

    Liu Wenjin; Zeng Zhongxiu; Ye Xianhui; Wu Wanjun

    2013-01-01

    Nonlinear seismic model for reactor with fuel assemblies of AFA 2G and AFA 3G is established. Using ANSYS software, seismic nonlinear time -history analysis is completed and the effects on seismic loads of reactor system are obtained. The result shows that when the fuel assembly changing from AFA 2G to AFA 3G, it is necessary to reevaluate the fuel assembly itself, but not the reactor internal. (authors)

  4. Optimal control of a fuel cell/wind/PV/grid hybrid system with thermal heat pump load

    CSIR Research Space (South Africa)

    Sichilalu, S

    2016-10-01

    Full Text Available This paper presents an optimal energy management strategy for a grid-tied photovoltaic–wind-fuel cell hybrid power supply system. The hybrid system meets the load demand consisting of an electrical load and a heat pump water heater supplying thermal...

  5. 40 CFR 86.129-94 - Road load power, test weight, inertia weight class determination, and fuel temperature profile.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 18 2010-07-01 2010-07-01 false Road load power, test weight, inertia... Procedures § 86.129-94 Road load power, test weight, inertia weight class determination, and fuel temperature... duty trucks 1,2,3 Test weightbasis 4,5 Test equivalent test weight(pounds) Inertia weight class(pounds...

  6. A risk-informed evaluation of MOX fuel loading in PWRS

    International Nuclear Information System (INIS)

    Lyman, E.S.

    2001-01-01

    The full text follows: The U.S. Department of Energy (DOE) has signed a contract with Duke Cogema Stone and Webster (DCS) for fabrication of mixed-oxide (MOX) fuel and irradiation of the MOX fuel at the Catawba and McGuire pressurized-water reactors (PWRs), operated by Duke Power. The first load of MOX fuel is scheduled for 2007. In order to use MOX in these plants, Duke Power will have to apply to the Nuclear Regulatory Commission (NRC) for amendments to their operating licenses. Until recently, there have been no numerical guidelines for determining the acceptability of license amendment requests. However, such guidelines are now at hand with the adoption in 1998 of NRC Regulatory Guide 1.174, which defines a maximum value for the permissible increase in risk to the public resulting from a proposed change to a nuclear plant's licensing basis (LB). The substitution of MOX fuel for low-enriched uranium (LEU) fuel in LWRs will have an impact on risk to the public that will require regulatory evaluation. One of the major differences is that use of MOX will increase the inventories of plutonium and minor actinides in the reactor core, thereby increasing the source term for certain severe accidents, such as a core melt with early containment failure or a spent fuel pool drain-down. The goal of this paper is to quantitatively evaluate the increase in risk associated with the greater actinide source term in MOX-fueled reactors, and to compare this increase with RG 1.174 guidelines. Standard computer programs (SCALE and MACCS2) are used to estimate the increase in severe accident risk to the public associated with the DCS plan to use 40% cores of weapons-grade MOX fuel. These values are then compared to the RG 1.174 acceptance criteria, using publicly available risk information. Since RG 1.174 guidelines are based on the assumption that severe accident source terms are not affected by LB changes, the RG 1.174 formalism must be modified for this case. A similar

  7. Reactors as a Source of Antineutrinos: Effects of Fuel Loading and Burnup for Mixed-Oxide Fuels

    Science.gov (United States)

    Bernstein, Adam; Bowden, Nathaniel S.; Erickson, Anna S.

    2018-01-01

    In a conventional light-water reactor loaded with a range of uranium and plutonium-based fuel mixtures, the variation in antineutrino production over the cycle reflects both the initial core fissile inventory and its evolution. Under an assumption of constant thermal power, we calculate the rate at which antineutrinos are emitted from variously fueled cores, and the evolution of that rate as measured by a representative ton-scale antineutrino detector. We find that antineutrino flux decreases with burnup for low-enriched uranium cores, increases for full mixed-oxide (MOX) cores, and does not appreciably change for cores with a MOX fraction of approximately 75%. Accounting for uncertainties in the fission yields in the emitted antineutrino spectra and the detector response function, we show that the difference in corewide MOX fractions at least as small as 8% can be distinguished using a hypothesis test. The test compares the evolution of the antineutrino rate relative to an initial value over part or all of the cycle. The use of relative rates reduces the sensitivity of the test to an independent thermal power measurement, making the result more robust against possible countermeasures. This rate-only approach also offers the potential advantage of reducing the cost and complexity of the antineutrino detectors used to verify the diversion, compared to methods that depend on the use of the antineutrino spectrum. A possible application is the verification of the disposition of surplus plutonium in nuclear reactors.

  8. AN INVESTIGATION TO RESOLVE THE INTERACTION BETWEEN FUEL CELL, POWER CONDITIONING SYSTEM AND APPLICATION LOADS

    Energy Technology Data Exchange (ETDEWEB)

    Sudip K. Mazumder; Chuck McKintyre; Dan Herbison; Doug Nelson; Comas Haynes; Michael von Spakovsky; Joseph Hartvigsen; S. Elangovan

    2003-11-03

    Solid-Oxide Fuel Cell (SOFC) stacks respond quickly to changes in load and exhibit high part- and full-load efficiencies due to its rapid electrochemistry. However, this is not true for the thermal, mechanical, and chemical balance-of-plant subsystem (BOPS), where load-following time constants are, typically, several orders of magnitude higher. This dichotomy diminishes the reliability and performance of the electrode with increasing demand of load. Because these unwanted phenomena are not well understood, the manufacturers of SOFC use conservative schemes (such as, delayed load-following to compensate for slow BOPS response or expensive inductor filtering) to control stack responses to load variations. This limits the applicability of SOFC systems for load-varying stationary and transportation applications from a cost standpoint. Thus, a need exists for the synthesis of component- and system-level models of SOFC power-conditioning systems and the development of methodologies for investigating the system-interaction issues (which reduce the lifetime and efficiency of a SOFC) and optimizing the responses of each subsystem, leading to optimal designs of power-conditioning electronics and optimal control strategies, which mitigate the electrical-feedback effects. Equally important are ''multiresolution'' finite-element modeling and simulation studies, which can predict the impact of changes in system-level variables (e.g., current ripple and load-transients) on the local current densities, voltages, and temperature (these parameters are very difficult or cumbersome, if not impossible to obtain) within a SOFC cell. Towards that end, for phase I of this project, sponsored by the U.S. DOE (NETL), we investigate the interactions among fuel cell, power-conditioning system, and application loads and their effects on SOFC reliability (durability) and performance. A number of methodologies have been used in Phase I to develop the steady-state and transient

  9. Spatial correction factors for YALINA Booster facility loaded with medium and low enriched fuels

    International Nuclear Information System (INIS)

    Talamo, A.; Gohar, Y.; Bournos, V.; Fokov, Y.; Kiyavitskaya, H.; Routkovskaya, C.

    2012-01-01

    The Bell and Glasstone spatial correction factor is used in analyses of subcritical assemblies to correct the experimental reactivity as function of the detector position. Besides the detector position, several other parameters affect the correction factor: the energy weighting function of the detector, the detector size, the energy-angle distribution of source neutrons, and the reactivity of the subcritical assembly. This work focuses on the dependency of the correction factor on the detector material and it investigates the YALINA Booster subcritical assembly loaded with medium (36%) and low (10%) enriched fuels. (authors)

  10. Spatial correction factors for YALINA Booster facility loaded with medium and low enriched fuels

    Energy Technology Data Exchange (ETDEWEB)

    Talamo, A.; Gohar, Y. [Argonne National Laboratory, 9700 S. Cass Ave, Argonne, IL 60439 (United States); Bournos, V.; Fokov, Y.; Kiyavitskaya, H.; Routkovskaya, C. [Joint Inst. for Power and Nuclear Research-Sosny, 99 Academician A.K.Krasin Str, Minsk 220109 (Belarus)

    2012-07-01

    The Bell and Glasstone spatial correction factor is used in analyses of subcritical assemblies to correct the experimental reactivity as function of the detector position. Besides the detector position, several other parameters affect the correction factor: the energy weighting function of the detector, the detector size, the energy-angle distribution of source neutrons, and the reactivity of the subcritical assembly. This work focuses on the dependency of the correction factor on the detector material and it investigates the YALINA Booster subcritical assembly loaded with medium (36%) and low (10%) enriched fuels. (authors)

  11. Preliminary report on the experiment performed in MARIUS reactor loaded with teledial fuel

    Energy Technology Data Exchange (ETDEWEB)

    Estiot, J C; Morier, F

    1972-06-15

    The experimental work described in this paper is part of a collaborative programme agreed between CEA and the Dragon Project. The aim of the programme is the measurement of the relative conversion ratio in a reactor loaded with Teledial fuel elements. The results will allow us to check our calculational methods and assumptions upon which the calculations are based, in the case of a teledial core, which represents a very complicated geometry, specially, due to the presence of the U238 with its resonance. The programme of experiments described in the paper have been completed. Some preliminary results are presented in the second part of this report (Part 2).

  12. Consequences of mis-loading and the power distribution in bowed fuel assemblies

    International Nuclear Information System (INIS)

    Andersson, Magnus

    2002-04-01

    The thesis is divided in two parts. The first part will investigate consequences of a mis-loaded fuel assembly in Ringhals 3, which is a pressurised water reactor (PWR). The aim of this work is to show that there are no or very small benefits from making an additional flux map at 30 % power in order to detect anomalies. Out of the 17 simulations, there exists only one type of mis-loading, which leads to problems. The case, which leads to problems, is when a Gd fitted assembly changes place with a non Gd. This leads to a too high power peaking factor and increased quadrant power tilt. The gain of a flux map at 30% power is small

  13. Development of the model for the stress calculation of fuel assembly under accident load

    International Nuclear Information System (INIS)

    Kim, Il Kon

    1993-01-01

    The finite element model for the stress calculation in guide thimbles of a fuel assembly (FA) under seismic and loss-of-coolant-accident (LOCA) load is developed. For the stress calculation of FA under accident load, at first the program MAIN is developed to select the worst bending mode shaped FA from core model. And then the model for the stress calculation of FA is developed by means of the finite element code. The calculated results of program MAIN are used as the kinematic constraints of the finite element model of a FA. Compared the calculated results of the stiffness of the finite element model of FA with the test results they have good agreements. (Author)

  14. Consequences of mis-loading and the power distribution in bowed fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Andersson, Magnus

    2002-04-01

    The thesis is divided in two parts. The first part will investigate consequences of a mis-loaded fuel assembly in Ringhals 3, which is a pressurised water reactor (PWR). The aim of this work is to show that there are no or very small benefits from making an additional flux map at 30 % power in order to detect anomalies. Out of the 17 simulations, there exists only one type of mis-loading, which leads to problems. The case, which leads to problems, is when a Gd fitted assembly changes place with a non Gd. This leads to a too high power peaking factor and increased quadrant power tilt. The gain of a flux map at 30% power is small.

  15. Computerized optimum distribution of loads among the turbogenerators of fossil-fuel electric power plants

    Energy Technology Data Exchange (ETDEWEB)

    Foshko, L S; Zusmanovich, L B; Flos, S L; Pal' chik, V A; Konevskii, B I

    1979-04-01

    The problem of determining the optimum distribution of loads among turbogenerators in a fossil-fuel power plant is considered based on satisfying the following requirements: distribution of electrical and thermal loads to minimize the heat expended on the turbine unit; calculation based on turbogenerator characteristics that most completely describe operating conditions; no constraints on the configuration of turbogenerator performance characteristics; calculation of load distribution based on net characteristics including the internal needs of the turbogenerators; consideration of all operational limitations in turbogenerator working conditions; results should be applicable to any predetermined differential of the load change. A flowchart is given showing the organization of the Optim-76 program complex for solution of this problem. An example is given showing application of the Optim-76 program implemented by a Minsk-32 computer in the case of a heat and electric power station with three turbogenerators. The results show that a dynamic programming method has considerable advantages for this applicaton on third-generation computers.

  16. Optimization of fuel core loading pattern design in a VVER nuclear power reactors using Particle Swarm Optimization (PSO)

    International Nuclear Information System (INIS)

    Babazadeh, Davood; Boroushaki, Mehrdad; Lucas, Caro

    2009-01-01

    The two main goals in core fuel loading pattern design optimization are maximizing the core effective multiplication factor (K eff ) in order to extract the maximum energy, and keeping the local power peaking factor (P q ) lower than a predetermined value to maintain fuel integrity. In this research, a new strategy based on Particle Swarm Optimization (PSO) algorithm has been developed to optimize the fuel core loading pattern in a typical VVER. The PSO algorithm presents a simple social model by inspiration from bird collective behavior in finding food. A modified version of PSO algorithm for discrete variables has been developed and implemented successfully for the multi-objective optimization of fuel loading pattern design with constraints of keeping P q lower than a predetermined value and maximizing K eff . This strategy has been accomplished using WIMSD and CITATION calculation codes. Simulation results show that this algorithm can help in the acquisition of a new pattern without contravention of the constraints.

  17. Performance of Transuranic-Loaded Fully Ceramic Micro-Encapsulated Fuel in LWRs Final Report, Including Void Reactivity Evaluation

    International Nuclear Information System (INIS)

    Pope, Michael A.; Sen, R. Sonat; Boer, Brian; Ougouag, Abderrafi M.; Youinou, Gilles

    2011-01-01

    The current focus of the Deep Burn Project is on once-through burning of transuranics (TRU) in light-water reactors (LWRs). The fuel form is called Fully-Ceramic Micro-encapsulated (FCM) fuel, a concept that borrows the tri-isotropic (TRISO) fuel particle design from high-temperature reactor technology. In the Deep Burn LWR (DB-LWR) concept, these fuel particles are pressed into compacts using SiC matrix material and loaded into fuel pins for use in conventional LWRs. The TRU loading comes from the spent fuel of a conventional LWR after 5 years of cooling. Unit cell and assembly calculations have been performed using the DRAGON-4 code to assess the physics attributes of TRU-only FCM fuel in an LWR lattice. Depletion calculations assuming an infinite lattice condition were performed with calculations of various reactivity coefficients performed at each step. Unit cells and assemblies containing typical UO2 and mixed oxide (MOX) fuel were analyzed in the same way to provide a baseline against which to compare the TRU-only FCM fuel. Then, assembly calculations were performed evaluating the performance of heterogeneous arrangements of TRU-only FCM fuel pins along with UO2 pins.

  18. High platinum utilization in ultra-low Pt loaded PEM fuel cell cathodes prepared by electrospraying

    Energy Technology Data Exchange (ETDEWEB)

    Martin, S.; Garcia-Ybarra, P.L.; Castillo, J.L. [Dept. Fisica Matematica y de Fluidos, Facultad de Ciencias, UNED, Senda del Rey 9, 28040 Madrid (Spain)

    2010-10-15

    Cathode electrodes for proton exchange membrane fuel cells (PEMFCs) with ultra-low platinum loadings as low as 0.012 mg{sub Pt}cm{sup -2} have been prepared by the electrospray method. The electrosprayed layers have nanostructured fractal morphologies with dendrites formed by clusters (about 100 nm diameter) of a few single catalyst particles rendering a large exposure surface of the catalyst. Optimization of the control parameters affecting this morphology has allowed us to overcome the state of the art for efficient electrodes prepared by electrospraying. Thus, using these cathodes in membrane electrode assemblies (MEAs), a high platinum utilization in the range 8-10 kW g{sup -1} was obtained for the fuel cell operating at 40 C and atmospheric pressure. Moreover, a platinum utilization of 20 kW g{sup -1} was attained under more suitable operating conditions (70 C and 3.4 bar over-pressure). These results substantially improve the performances achieved previously with other low platinum loading electrodes prepared by electrospraying. (author)

  19. Full-sized plates irradiation with high UMo fuel loading. Final results of IRIS 1 experiment

    International Nuclear Information System (INIS)

    Huet, F.; Marelle, V.; Noirot, J.; Sacristan, P.; Lemoine, P.

    2003-01-01

    As a part of the French UMo Group qualification program, IRIS 1 experiment contained full-sized plates with high uranium loading in the meat of 8 g.cm -3 . The fuel particles consisted of 7 and 9 wt% Mo-uranium alloys ground powders. The plate were irradiated at OSIRIS reactor in IRIS device up to 67.5% peak burnup within the range of 136 W.cm - '2 for the heat flux and 72 deg. C for the cladding temperature. After each reactor cycle the plates thickness were measured. The results show no swelling behaviour differences versus burnup between UMo7 and UMo9 plates. The maximum plate swelling for peak burnup location remains lower than 6%. The wide set of PIE has shown that, within the studied irradiation conditions, the interaction product have a global formulation of '(U-Mo)Al -7 ' and that there is no aluminium dissolution in UMo particles. IRIS1 experiment, as the first step of the UMo fuel qualification for research reactor, has established the good behaviour of UMo7 and UMo9 high uranium loading full-sized plate within the tested conditions. (author)

  20. An Investigation to Resolve the Interaction Between Fuel Cell, Power Conditioning System and Application Loads

    Energy Technology Data Exchange (ETDEWEB)

    Sudip K. Mazumder

    2005-12-31

    Development of high-performance and durable solidoxide fuel cells (SOFCs) and a SOFC power-generating system requires knowledge of the feedback effects from the power-conditioning electronics and from application-electrical-power circuits that may pass through or excite the power-electronics subsystem (PES). Therefore, it is important to develop analytical models and methodologies, which can be used to investigate and mitigate the effects of the electrical feedbacks from the PES and the application loads (ALs) on the reliability and performance of SOFC systems for stationary and non-stationary applications. However, any such attempt to resolve the electrical impacts of the PES on the SOFC would be incomplete unless one utilizes a comprehensive analysis, which takes into account the interactions of SOFC, PES, balance-of-plant system (BOPS), and ALs as a whole. SOFCs respond quickly to changes in load and exhibit high part- and full-load efficiencies due to its rapid electrochemistry, which is not true for the thermal and mechanical time constants of the BOPS, where load-following time constants are, typically, several orders of magnitude higher. This dichotomy can affect the lifetime and durability of the SOFCSs and limit the applicability of SOFC systems for load-varying stationary and transportation applications. Furthermore, without validated analytical models and investigative design and optimization methodologies, realizations of cost-effective, reliable, and optimal PESs (and power-management controls), in particular, and SOFC systems, in general, are difficult. On the whole, the research effort can lead to (a) cost-constrained optimal PES design for high-performance SOFCS and high energy efficiency and power density, (b) effective SOFC power-system design, analyses, and optimization, and (c) controllers and modulation schemes for mitigation of electrical impacts and wider-stability margin and enhanced system efficiency.

  1. Uncertainty Quantification of Fork Detector Measurements from Spent Fuel Loading Campaigns

    International Nuclear Information System (INIS)

    Vaccaro, S.; De Baere, P.; Schwalbach, P.; Gauld, I.; Hu, J.

    2015-01-01

    With increasing activities at the end of the fuel cycle, the requirements for the verification of spent nuclear fuel for safeguards purposes are continuously growing. In the European Union we are experiencing a dramatic increase in the number of cask loadings for interim dry storage. This is caused by the progressive shut-down of reactors, related to facility ageing but also due to politically motivated phase-out of nuclear power. On the other hand there are advanced plans for the construction of encapsulation plants and geological repositories. The cask loading or the encapsulation process will provide the last occasion to verify the spent fuel assemblies. In this context, Euratom and the US DOE have carried out a critical review of the widely used Fork measurements method of irradiated assemblies. The Nuclear Safeguards directorates of the European Commission's Directorate General for Energy and Oak Ridge National Laboratory have collaborated to improve the Fork data evaluation process and simplify its use for inspection applications. Within the Commission's standard data evaluation package CRISP, we included a SCALE/ORIGEN-based irradiation and depletion simulation of the measured assembly and modelled the fork transfer function to calculate expected count rates based on operator's declarations. The complete acquisition and evaluation process has been automated to compare expected (calculated) with measured count rates. This approach allows a physics-based improvement of the data review and evaluation process. At the same time the new method provides the means for better measurement uncertainty quantification. The present paper will address the implications of the combined approach involving measured and simulated data to the quantification of measurement uncertainty and the consequences of these uncertainties in the possible use of the Fork detector as a partial defect detection method. (author)

  2. Pinus contorta invasions increase wildfire fuel loads and may create a positive feedback with fire.

    Science.gov (United States)

    Taylor, Kimberley T; Maxwell, Bruce D; McWethy, David B; Pauchard, Aníbal; Nuñez, Martín A; Whitlock, Cathy

    2017-03-01

    Invasive plant species that have the potential to alter fire regimes have significant impacts on native ecosystems. Concern that pine invasions in the Southern Hemisphere will increase fire activity and severity and subsequently promote further pine invasion prompted us to examine the potential for feedbacks between Pinus contorta invasions and fire in Patagonia and New Zealand. We determined how fuel loads and fire effects were altered by P. contorta invasion. We also examined post-fire plant communities across invasion gradients at a subset of sites to assess how invasion alters the post-fire vegetation trajectory. We found that fuel loads and soil heating during simulated fire increase with increasing P. contorta invasion age or density at all sites. However, P. contorta density did not always increase post-fire. In the largest fire, P. contorta density only increased significantly post-fire where the pre-fire P. contorta density was above an invasion threshold. Below this threshold, P. contorta did not dominate after fire and plant communities responded to fire in a similar manner as uninvaded communities. The positive feedback observed at high densities is caused by the accumulation of fuel that in turn results in greater soil heating during fires and high P. contorta density post-fire. Therefore, a positive feedback may form between P. contorta invasions and fire, but only above an invasion density threshold. These results suggest that management of pine invasions before they reach the invasion density threshold is important for reducing fire risk and preventing a transition to an alternate ecosystem state dominated by pines and novel understory plant communities. © 2016 by the Ecological Society of America.

  3. Mixed PWR core loadings with inert matrix Pu-fuel assemblies

    International Nuclear Information System (INIS)

    Stanculescu, A.; Kasemeyer, U.; Paratte, J.-M.; Chawla, R.

    1999-01-01

    The most efficient way to enhance plutonium consumption in light water reactors is to eliminate the production of plutonium all together. This requirement leads to fuel concepts in which the uranium is replaced by an inert matrix. At PSI, studies have focused on employing ZrO 2 as inert matrix. Adding a burnable poison to such a fuel proves to be necessary. As a result of scoping studies, Er 2 O 3 was identified as the most suitable burnable poison material. The results of whole-core three-dimensional neutronics analyses indicated, for a present-day 1000 MW e pressurised water reactor, the feasibility of an asymptotic equilibrium four-batch cycle fuelled solely with the proposed PuO 2 -Er 2 O 3 -ZrO 2 inert matrix fuel (IMF). The present paper presents the results of more recent investigations related to 'real-life' situations, which call for transition configurations in which mixed IMF and UO 2 assembly loadings must be considered. To determine the influence of the introduction of IMF assemblies on the characteristics of a UO 2 -fuelled core, three-dimensional full-core calculations have been performed for a present-day 1000 MW e PWR containing up to 12 optimised IMF assemblies. (author)

  4. Control and maintenance of the Superphenix knowledge and its specific sodium skills through an innovative partnership between EDF and AREVA

    International Nuclear Information System (INIS)

    Calais, Thomas; Rauber, Jean-Claude

    2016-01-01

    Superphenix is a 1200 MWe sodium cooled Fast Breeder Reactor (FBR) located in Creys-Malville (France). Its grid coupling occurred in 1986 and its final shutdown pronounced through a decree, 12 years later, in 1998. This Superphenix final shutdown decision marked a new stage in the life of the nuclear plant. Decommissioning activities were highly challenging due to the following: - Non recurrent and first-of-a-kind (FOAK) characteristics; - Environment constraints: radiation level, high temperatures, presence of argon, sodium, NaK, soda, hydrogen, etc.; - Complexity of the primary vessel internal structures; - Numerous interfaces to manage; - Numerous technical uncertainties due to the difficulty in anticipating the effective state of components (sodium and aerosols retentions, tritium concentration, NaK alteration, etc.). At the end of 1998, exchanges took place between EDF as 'Superphenix nuclear operator' and AREVA as 'Superphenix Nuclear Steam System Supply (NSSS) designer' in order to find the best way to meet the new challenge of decommissioning Superphenix. A key ingredient to achieving success was to ensure that existing local and specific sodium skills were controlled and maintained. AREVA was selected by EDF as its industrial partner for the sodium activities on this project being entrusted with the following missions: - Maintaining and adapting a strong EDF / AREVA partnership within the project duration; - Supplying support as the 'NSSS Designer'; - Rolling-out multidisciplinary skills from the design to the on-site operations; - Relying on its best technical experts to solve each technical challenge; - Developing and adapting durable specific skills of its technical team (sodium, mechanical, process, I and C, statutory, etc.) following each stage of the decommissioning. This EDF/AREVA partnership on the sodium activities has taken different forms according to the different stages of the project. From 1998 to 2005, AREVA was

  5. Comparison of thermal and radical effects of EGR gases on combustion process in dual fuel engines at part loads

    International Nuclear Information System (INIS)

    Pirouzpanah, V.; Khoshbakhti Saray, R.; Sohrabi, A.; Niaei, A.

    2007-01-01

    Dual fuel engines at part load inevitably suffer from lower thermal efficiency and higher emission of carbon monoxide and unburned fuel. This work is conducted to investigate the combustion characteristics of a dual fuel (Diesel-gas) engine at part loads using a single zone combustion model with detailed chemical kinetics for combustion of natural gas fuel. In this home made software, the presence of the pilot fuel is considered as a heat source that is deriving form two superposed Wiebe's combustion functions to account for its contribution to ignition of the gaseous fuel and the rest of the total released energy. The chemical kinetics mechanism consists of 112 reactions with 34 species. This combustion model is able to establish the development of the combustion process with time and the associated important operating parameters, such as pressure, temperature, heat release rate (HRR) and species concentration. Therefore, this work is an attempt to investigate the combustion phenomenon at part load and using exhaust gas recirculation (EGR) to improve the above mentioned problems. Also, the results of this work show that each of the different cases of EGR (thermal, chemical and radical cases) has an important role on the combustion process in dual fuel engines at part loads. It is found that all the different cases of EGR have positive effects on the performance and emission parameters of dual fuel engines at part loads despite the negative effect of some diluent gases in the chemical case, which moderates too much the positive effects of the thermal and radical cases of EGR. Predicted values show good agreement with corresponding experimental values over the whole range of engine operating conditions. Implications will be discussed in detail

  6. Advanced and flexible genetic algorithms for BWR fuel loading pattern optimization

    International Nuclear Information System (INIS)

    Martin-del-Campo, Cecilia; Palomera-Perez, Miguel-Angel; Francois, Juan-Luis

    2009-01-01

    This work proposes advances in the implementation of a flexible genetic algorithm (GA) for fuel loading pattern optimization for Boiling Water Reactors (BWRs). In order to avoid specific implementations of genetic operators and to obtain a more flexible treatment, a binary representation of the solution was implemented; this representation had to take into account that a little change in the genotype must correspond to a little change in the phenotype. An identifier number is assigned to each assembly by means of a Gray Code of 7 bits and the solution (the loading pattern) is represented by a binary chain of 777 bits of length. Another important contribution is the use of a Fitness Function which includes a Heuristic Function and an Objective Function. The Heuristic Function which is defined to give flexibility on the application of a set of positioning rules based on knowledge, and the Objective Function that contains all the parameters which qualify the neutronic and thermal hydraulic performances of each loading pattern. Experimental results illustrating the effectiveness and flexibility of this optimization algorithm are presented and discussed.

  7. Thermal Characteristic Of AIMg2 Cladding And Fuel Plates Of U3Si2-Al With Various Uranium Loading

    International Nuclear Information System (INIS)

    Aslina, Br. G.; Suparjo; Aggraini, D.; Hasbullah, N.

    1998-01-01

    Thermal characteristic analyzed in this paper included linear expansion value, coefficient expansion, and enthalpy of cladding material fuel core and fuel plate of U 3 Si 2 -AI. Before analyzing, the fresh cladding of AIMg2 (without treatment) and the rolled AIMg2 were annealed at temperature of 425 o C for 1 hour, and the fuel plates of U 3 Si 2 -AI was prepared for various uranium loading of 0.9 - 3.6 - 4.2 - 4.8 and 5.2 g/cm 3 . Linear expansion nominal value and expansion coefficient were analyzed by using Dilatometer whereas enthalpy determination used Differential Thermal Analysis (DTA). The linear expansion and expansion coefficient analysis was performed to study the dimension cladding and of fuel plates during their stay in the reactor core, whereas determination of enthalpy was carried out to estimate the energy absorbed and released by fuel meat of U 3 Si 2 -AI to the cooling water through AlMg2 as a cladding. The result showed that the linear expansion and expansion coefficient of fresh AIMg2 cladding, rolled AIMg2 and fuel plates of U 3 Si 2 -AI are increased with the increase of temperature as well as the increase of uranium loading. The enthalpy measure showed that the enthalpy of fresh AIMg2 is smaller than that of rolled AIMg2 but melting temperature of fresh AIMg2 is greater than that of rolled AIMg2. The enthalpy of fuel plates and meat of U 3 Si 2 -AI is less than that of plates of U 3 Si 2 -AI. The enthalpy of fuel platers and meat of U 3 Si 2 -AI decrease with the increase of uranium loading. It is concluded that the fuel meat more reactive than fuel plates of U 3 Si 2 -AI

  8. Performance improvement of a battery/PV/fuel cell/grid hybrid energy system considering load uncertainty modeling using IGDT

    International Nuclear Information System (INIS)

    Nojavan, Sayyad; Majidi, Majid; Zare, Kazem

    2017-01-01

    Highlights: • Optimum performance of PV/battery/fuel cell/grid hybrid system under load uncertainty. • Employing information gap decision theory (IGDT) to model the load uncertainty. • Robustness and opportunity functions of IGDT are modeled for risk-averse and risk-taker. • Robust strategy of hybrid system's operation obtained from robustness function. • Opportunistic strategy of hybrid system's operation obtained from opportunity function. - Abstract: Nowadays with the speed that electrical loads are growing, system operators are challenged to manage the sources they use to supply loads which means that that besides upstream grid as the main sources of electric power, they can utilize renewable and non-renewable energy sources to meet the energy demand. In the proposed paper, a photovoltaic (PV)/fuel cell/battery hybrid system along with upstream grid has been utilized to supply two different types of loads: electrical load and thermal load. Operators should have to consider load uncertainty to manage the strategies they employ to supply load. In other words, operators have to evaluate how load variation would affect their energy procurement strategies. Therefore, information gap decision theory (IGDT) technique has been proposed to model the uncertainty of electrical load. Utilizing IGDT approach, robustness and opportunity functions are achieved which can be used by system operator to take the appropriate strategy. The uncertainty modeling of load enables operator to make appropriate decisions to optimize the system’s operation against possible changes in load. A case study has been simulated to validate the effects of proposed technique.

  9. [Spatial pattern of land surface dead combustible fuel load in Huzhong forest area in Great Xing'an Mountains].

    Science.gov (United States)

    Liu, Zhi-Hua; Chang, Yu; Chen, Hong-Wei; Zhou, Rui; Jing, Guo-Zhi; Zhang, Hong-Xin; Zhang, Chang-Meng

    2008-03-01

    By using geo-statistics and based on time-lag classification standard, a comparative study was made on the land surface dead combustible fuels in Huzhong forest area in Great Xing'an Mountains. The results indicated that the first level land surface dead combustible fuel, i. e., 1 h time-lag dead fuel, presented stronger spatial auto-correlation, with an average of 762.35 g x m(-2) and contributing to 55.54% of the total load. Its determining factors were species composition and stand age. The second and third levels land surface dead combustible fuel, i. e., 10 h and 100 h time-lag dead fuels, had a sum of 610.26 g x m(-2), and presented weaker spatial auto-correlation than 1 h time-lag dead fuel. Their determining factor was the disturbance history of forest stand. The complexity and heterogeneity of the factors determining the quality and quantity of forest land surface dead combustible fuels were the main reasons for the relatively inaccurate interpolation. However, the utilization of field survey data coupled with geo-statistics could easily and accurately interpolate the spatial pattern of forest land surface dead combustible fuel loads, and indirectly provide a practical basis for forest management.

  10. Recent developments in ultrasonic probes working up to 180 deg C for the inspection of the Superphenix fast breeder reactor

    International Nuclear Information System (INIS)

    Gondard, C.

    1987-01-01

    The main and safety vessels of SUPERPHENIX were designed to allow In-Service-Inspections. The remote controlled device MIR was developed for this purpose. The ultrasonic examination has required the development of all new focused transducers fitted with severe operating conditions prevailing in the intervessels interval: nitrogen gas at 180 0 C. We give a list of problems to be resolved and technological solutions which were found. Measurements of acoustical properties on actual probes are compared with theoretical values. We produce some examples obtained in actual conditions which show the detection of reference reflectors located in welds at various depth, with various disalignements against focus beam. Inspite of the severe environment and the perturbations caused by the austenitic welds, the I.S.I of SPX1 using high temperatures transducers is possible with a good spatial resolution and signal to noise ratio

  11. Reracking of fuel pools, experience with improved codes and design for reactor sites with high seismic loads

    International Nuclear Information System (INIS)

    Banck, J.; Wirtz, K.

    1998-01-01

    Reracking of existing pools to the maximum extent is desirable from the economical point of view. Although the load onto the storage rack structure and the fuel pool bottom will be increased, new improved codes, optimized structural qualification procedures and advanced design enable to demonstrate the structural integrity for all normal and accident conditions so that the design provides a safe compact storage of spent fuel under any condition.(author)

  12. Effect of fuel assembly mechanical design changes on dynamic response of reactor pressure vessel system under extreme loadings

    International Nuclear Information System (INIS)

    Bhandari, D.R.; Hankinson, M.F.

    1993-01-01

    This paper presents the results of a study to assess the effect of fuel assembly mechanical design changes on the dynamic response of a pressurized water reactor vessel and reactor internals under Loss-Of-Coolant Accident (LOCA) conditions. The results of this study show that the dynamic response of the reactor vessel internals and the core under extreme loadings, such as LOCA, is very sensitive to fuel assembly mechanical design changes. (author)

  13. Lowering the platinum loading of high temperature polymer electrolyte membrane fuel cells with acid doped polybenzimidazole membranes

    DEFF Research Database (Denmark)

    Fernandez, Santiago Martin; Li, Qingfeng; Jensen, Jens Oluf

    2015-01-01

    Membrane electrode assemblies (MEAs) with ultra-low Pt loading electrodes were prepared for high temperature polymer electrolyte membrane fuel cells (HT-PEMFCs) based on acid doped polybenzimidazole. With no electrode binders or ionomers, the triple phase boundary of the catalyst layer was establ......Membrane electrode assemblies (MEAs) with ultra-low Pt loading electrodes were prepared for high temperature polymer electrolyte membrane fuel cells (HT-PEMFCs) based on acid doped polybenzimidazole. With no electrode binders or ionomers, the triple phase boundary of the catalyst layer...

  14. New fuel air control strategy for reducing NOx emissions from corner-fired utility boilers at medium-low loads

    DEFF Research Database (Denmark)

    Zhao, Sinan; Fang, Qingyan; Yin, Chungen

    2017-01-01

    Due to the rapidly growing renewable power, the fossil fuel power plants have to be increasingly operated under large and rapid load change conditions, which can induce various challenges. This work aims to reduce NOx emissions of large-scale corner-fired boilers operated at medium–low loads....... The combustion characteristics and NOx emissions from a 1000 MWe corner-fired tower boiler under different loads are investigated experimentally and numerically. A new control strategy for the annular fuel air is proposed and implemented in the boiler, in which the secondary air admitted to the furnace through...... the air annulus around each coal nozzle tip is controlled by the boiler load, instead of being controlled by the output of the connected mill as commonly used in this kind of power plant. Both the experimental and simulation results show that the new control strategy reduces NOx emissions at the entrance...

  15. Fast Response, Load-Matching Hybrid Fuel Cell: Final Technical Progress Report

    Energy Technology Data Exchange (ETDEWEB)

    Key, T. S.; Sitzlar, H. E.; Geist, T. D.

    2003-06-01

    Hybrid DER technologies interconnected with the grid can provide improved performance capabilities compared to a single power source, and, add value, when matched to appropriate applications. For example, in a typical residence, the interconnected hybrid system could provide power during a utility outage, and also could compensate for voltage sags in the utility service. Such a hybrid system would then function as a premium power provider and eliminate the potential need for an uninterruptible power supply. In this research project, a proton exchange membrane (PEM) fuel cell is combined with an asymmetrical ultracapacitor to provide robust power response to changes in system loading. This project also considers the potential of hybrid DER technologies to improve overall power system compatibility and performance. This report includes base year accomplishments of a proposed 3-year-option project.

  16. Modeling Low-Platinum-Loading Effects in Fuel-Cell Catalyst Layers

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Wonseok; Weber, Adam Z.

    2011-01-01

    The cathode catalyst layer within a proton-exchange-membrane fuel cell is the most complex and critical, yet least understood, layer within the cell. The exact method and equations for modeling this layer are still being revised and will be discussed in this paper, including a 0.8 reaction order, existence of Pt oxides, possible non-isopotential agglomerates, and the impact of a film resistance towards oxygen transport. While the former assumptions are relatively straightforward to understand and implement, the latter film resistance is shown to be critically important in explaining increased mass-transport limitations with low Pt-loading catalyst layers. Model results demonstrate agreement with experimental data that the increased oxygen flux and/or diffusion pathway through the film can substantially decrease performance. Also, some scale-up concepts from the agglomerate scale to the more macroscopic porous-electrode scale are discussed and the resulting optimization scenarios investigated.

  17. Fuel load and flight ranges of blackcaps Sylvia atricapilla in northern Iberia during autumn and spring migrations

    Directory of Open Access Journals (Sweden)

    JUAN ARIZAGA, EMILIO BARBA

    2009-12-01

    Full Text Available Fuel accumulation, mainly as fatty acids, is one of the main characteristics of migratory birds. Studying to what extent each population or species manages fuel load and how it varies along routes of migration or between seasons (autumn and spring migrations is crucial to our understanding of bird migration strategies. Our aim here was to analyse whether migratory blackcaps Sylvia atricapilla passing through northern Iberia differ in their mean fuel loads, rate of fuel accumulation and 'potential' flight ranges between migration seasons. Blackcaps were mist netted for 4 h-periods beginning at dawn from 16 September to 15 November 2003–2005, and from 1 March to 30 April 2004–2006 in a European Atlantic hedgerow at Loza, northern Iberia. Both fuel load and fuel deposition rate (this latter assessed with difference in body mass of within-season recaptured individuals were higher in autumn than in spring. Possible hypotheses explaining these results could be seasonal-associated variations in food availability (likely lower during spring than during autumn, the fact that a fraction of the migrants captured in spring could breed close to the study area and different selective pressures for breeding and wintering [Current Zoology 55 (6: 401–410, 2009].

  18. Enzymatic hydrolysis at high-solids loadings for the conversion of agave bagasse to fuel ethanol

    International Nuclear Information System (INIS)

    Caspeta, Luis; Caro-Bermúdez, Mario A.; Ponce-Noyola, Teresa; Martinez, Alfredo

    2014-01-01

    Highlights: • Conversion of agave bagasse to fuel ethanol. • Ethanosolv-pretreatment variables were statistically adjusted. • 91% of total sugars found in agave bagasse were recovered. • 225 g/L glucose from 30%-consistency hydrolysis using mini-reactors with peg-mixers. • 0.25 g of ethanol per g of dry agave bagasse was obtained. - Abstract: Agave bagasse is the lignocellulosic residue accumulated during the production of alcoholic beverages in Mexico and is a potential feedstock for the production of biofuels. A factorial design was used to investigate the effect of temperature, residence time and concentrations of acid and ethanol on ethanosolv pretreatment and enzymatic hydrolysis of agave bagasse. This method and the use of a stirred in-house-made mini-reactor increased the digestibility of agave bagasse from 30% observed with the dilute-acid method to 98%; also allowed reducing the quantity of enzymes used to hydrolyze samples with solid loadings of 30% w/w and glucose concentrations up to 225 g/L were obtained in the enzymatic hydrolysates. Overall this process allows the recovery of 91% of the total fermentable sugars contained in the agave bagasse (0.51 g/g) and 69% of total lignin as co-product (0.11 g/g). The maximum ethanol yield under optimal conditions using an industrial yeast strain for the fermentation was 0.25 g/g of dry agave bagasse, which is 86% of the maximum theoretical (0.29 g/g). The effect of the glucose concentration and solid loading on the conversion of cellulose to glucose is discussed, in addition to prospective production of about 50 million liters of fuel ethanol using agave bagasse residues from the tequila industry as a potential solution to the disposal problems

  19. Electrocatalytic performance of fuel cell reactions at low catalyst loading and high mass transport.

    Science.gov (United States)

    Zalitis, Christopher M; Kramer, Denis; Kucernak, Anthony R

    2013-03-28

    An alternative approach to the rotating disk electrode (RDE) for characterising fuel cell electrocatalysts is presented. The approach combines high mass transport with a flat, uniform, and homogeneous catalyst deposition process, well suited for studying intrinsic catalyst properties at realistic operating conditions of a polymer electrolyte fuel cell (PEFC). Uniform catalyst layers were produced with loadings as low as 0.16 μgPt cm(-2) and thicknesses as low as 200 nm. Such ultra thin catalyst layers are considered advantageous to minimize internal resistances and mass transport limitations. Geometric current densities as high as 5.7 A cm(-2)Geo were experimentally achieved at a loading of 10.15 μgPt cm(-2) for the hydrogen oxidation reaction (HOR) at room temperature, which is three orders of magnitude higher than current densities achievable with the RDE. Modelling of the associated diffusion field suggests that such high performance is enabled by fast lateral diffusion within the electrode. The electrodes operate over a wide potential range with insignificant mass transport losses, allowing the study of the ORR at high overpotentials. Electrodes produced a specific current density of 31 ± 9 mA cm(-2)Spec at a potential of 0.65 V vs. RHE for the oxygen reduction reaction (ORR) and 600 ± 60 mA cm(-2)Spec for the peak potential of the HOR. The mass activity of a commercial 60 wt% Pt/C catalyst towards the ORR was found to exceed a range of literature PEFC mass activities across the entire potential range. The HOR also revealed fine structure in the limiting current range and an asymptotic current decay for potentials above 0.36 V. These characteristics are not visible with techniques limited by mass transport in aqueous media such as the RDE.

  20. HTGR fuel development: loading of uranium on carboxylic acid cation-exchange resins using solvent extraction of nitrate

    International Nuclear Information System (INIS)

    Haas, P.A.

    1975-09-01

    The reference fuel kernel for recycle of 233 U to HTGR's (High-Temperature Gas-Cooled Reactors) is prepared by loading carboxylic acid cation-exchange resins with uranium and carbonizing at controlled conditions. The purified 233 UO 2 (NO 3 ) 2 solution from a fuel reprocessing plant contains excess HNO 3 (NO 3 - /U ratio of approximately 2.2). The reference flowsheet for a 233 U recycle fuel facility at Oak Ridge uses solvent extraction of nitrate by a 0.3 M secondary amine in a hydrocarbon diluent to prepare acid-deficient uranyl nitrate. This nitrate extraction, along with resin loading and amine regeneration steps, was demonstrated in 14 runs. No significant operating difficulties were encountered. The process is controlled via in-line pH measurements for the acid-deficient uranyl nitrate solutions. Information was developed on pH values for uranyl nitrate solution vs NO 3 - /U mole ratios, resin loading kinetics, resin drying requirements, and other resin loading process parameters. Calculations made to estimate the capacities of equipment that is geometrically safe with respect to control of nuclear criticality indicate 100 kg/day or more of uranium for single nitrate extraction lines with one continuous resin loading contactor or four batch loading contactors. (auth)

  1. Highly Zeolite-Loaded Polyvinyl Alcohol Composite Membranes for Alkaline Fuel-Cell Electrolytes

    Directory of Open Access Journals (Sweden)

    Po-Ya Hsu

    2018-01-01

    Full Text Available Having a secure and stable energy supply is a top priority for the global community. Fuel-cell technology is recognized as a promising electrical energy generation system for the twenty-first century. Polyvinyl alcohol/zeolitic imidazolate framework-8 (PVA/ZIF-8 composite membranes were successfully prepared in this work from direct ZIF-8 suspension solution (0–45.4 wt % and PVA mixing to prevent filler aggregation for direct methanol alkaline fuel cells (DMAFCs. The ZIF-8 fillers were chosen for the appropriate cavity size as a screening aid to allow water and suppress methanol transport. Increased ionic conductivities and suppressed methanol permeabilities were achieved for the PVA/40.5% ZIF-8 composites, compared to other samples. A high power density of 173.2 mW cm−2 was achieved using a KOH-doped PVA/40.5% ZIF-8 membrane in a DMAFC at 60 °C with 1–2 mg cm−2 catalyst loads. As the filler content was raised beyond 45.4 wt %, adverse effects resulted and the DMAFC performance (144.9 mW cm−2 was not improved further. Therefore, the optimal ZIF-8 content was approximately 40.5 wt % in the polymeric matrix. The specific power output was higher (58 mW mg−1 than most membranes reported in the literature (3–18 mW mg−1.

  2. Influence of Fuel Load Dynamics on Carbon Emission by Wildfires in the Clay Belt Boreal Landscape

    Directory of Open Access Journals (Sweden)

    Aurélie Terrier

    2016-12-01

    Full Text Available Old-growth forests play a decisive role in preserving biodiversity and ecological functions. In an environment frequently disturbed by fire, the importance of old-growth forests as both a carbon stock as well as a source of emissions when burnt is not fully understood. Here, we report on carbon accumulation with time since the last fire (TSF in the dominant forest types of the Clay Belt region in eastern North America. To do so, we performed a fuel inventory (tree biomass, herbs and shrubs, dead woody debris, and duff loads along four chronosequences. Carbon emissions by fire through successional stages were simulated using the Canadian Fire Effects Model. Our results show that fuel accumulates with TSF, especially in coniferous forests. Potential carbon emissions were on average 11.9 t·ha−1 and 29.5 t·ha−1 for old-growth and young forests, respectively. In conclusion, maintaining old-growth forests in the Clay Belt landscape not only ensures a sustainable management of the boreal forest, but it also optimizes the carbon storage.

  3. CSER 94-014: Storage of metal-fuel loaded EBR-II casks in concrete vault on PFP grounds

    International Nuclear Information System (INIS)

    Hess, A.L.

    1994-01-01

    A criticality safety evaluation is presented to permit EBR-2 spent fuel casks loaded with metallic fuel rods to be stored in an 8-ft diameter, cylindrical concrete vault inside the PFP security perimeter. The specific transfer of three casks with Pu alloy fuel from the Los Alamos Molten Plutonium Reactor Experiment from the burial grounds to the vault is thus covered. Up to seven casks may be emplaced in the casing with 30 inches center to center spacing. Criticality safety is assured by definitive packaging rules which keep the fissile medium dry and at a low effective volumetric density

  4. Fuel management for off-load annual refuelling of the D-HHT 600 MW(e) reference core

    Energy Technology Data Exchange (ETDEWEB)

    Hansen, U

    1973-03-16

    The reference design for the Dragon-HHT reactor has been optimised for on-load continuous refuelling. The possiblity to operate the reactor on a discontinuous annual reloading schedule might prove of interest and/or necessity. In this paper the influence of an annual 4-batch fuel management scheme on the core physics and fuel cycle economics is investigated. The results of the present investigation give a good indication of the relative merits of the two fuel management schemes. Although a broader parameter survey and a more detailed scrutinising of special cases would be desirable, we feel that the main conclusions are correct and that the principle differences have been elicited.

  5. Diffusion and Gas Conversion Analysis of Solid Oxide Fuel Cells at Loads via AC Impedance

    Directory of Open Access Journals (Sweden)

    Robert U. Payne

    2011-01-01

    Full Text Available Impedance measurements were conducted under practical load conditions in solid oxide fuel cells of differing sizes. For a 2 cm2 button cell, impedance spectra data were separately measured for the anode, cathode, and total cell. Improved equivalent circuit models are proposed and applied to simulate each of measured impedance data. Circuit elements related to the chemical and physical processes have been added to the total-cell model to account for an extra relaxation process in the spectra not measured at either electrode. The processes to which elements are attributed have been deduced by varying cell temperature, load current, and hydrogen concentration. Spectra data were also obtained for a planar stack of five 61 cm2 cells and the individual cells therein, which were fitted to a simplified equivalent circuit model of the total button cell. Similar to the button cell, the planar cells and stack exhibit a pronounced low-frequency relaxation process, which has been attributed to concentration losses, that is, the combined effects of diffusion and gas conversion. The simplified total-cell model approximates well the dynamic behavior of the SOFC cells and the whole stack.

  6. Impact on vehicle fuel economy of the soot loading on diesel particulate filters made of different substrate materials

    International Nuclear Information System (INIS)

    Millo, Federico; Andreata, Maurizio; Rafigh, Mahsa; Mercuri, Davide; Pozzi, Chiara

    2015-01-01

    Wall flow DPFs (Diesel Particulate Filters) are nowadays universally adopted for all European passenger cars. Since the properties of the filter substrate material play a fundamental role in determining the optimal soot loading level to be reached before DPF regeneration, three different filter material substrates (Silicon Carbide, Aluminum Titanate and Cordierite) were investigated in this work, considering different driving conditions, after treatment layouts and regeneration strategies. In the first step of the research, an experimental investigation on the three different substrates over the NEDC (New European Driving Cycle) was performed. The data obtained from experiments were then used for the calibration and the validation of a one dimensional fluid-dynamic engine and after treatment simulation model. Afterward, the model was used to predict the vehicle fuel consumption increments as a function of the exhaust back pressure due to the soot loading for different driving cycles. The results showed that appreciable fuel consumption increments could be noticed only in particular driving conditions, and, as a consequence, in most of the cases the optimal filter regeneration strategy corresponds to reach the highest soot loading that still ensures the component safety even in case of uncontrolled regeneration events. - Highlights: • Three different substrate materials for a Diesel Particulate Filter were investigated. • Fuel consumption increases due to DPF soot loading were generally not appreciable. • Optimal soot loading before regeneration was the highest safeguarding DPF integrity. • SiC substrate showed highest soot load limit and lowest fuel consumption penalties. • AT and Cd substrate properties lead to lower soot load limits than SiC

  7. 40 CFR 86.1229-85 - Dynamometer load determination and fuel temperature profile.

    Science.gov (United States)

    2010-07-01

    ... VEHICLES AND ENGINES (CONTINUED) Evaporative Emission Test Procedures for New Gasoline-Fueled, Natural Gas-Fueled, Liquefied Petroleum Gas-Fueled and Methanol-Fueled Heavy-Duty Vehicles § 86.1229-85 Dynamometer... has more than one fuel tank, a profile shall be established for each tank. Manufacturers may also...

  8. High-Uranium-Loaded U3O8-Al fuel element development program. Part 2

    International Nuclear Information System (INIS)

    Knight, R.

    1993-01-01

    Texas Instruments is a product intensive company that manufactures very high volumes of different products, and because of this, their technique in manufacturing is what we call hard tooling. So all of the tools we use at this site whether it is for HFIR, ORR or HFBR are hard tooling. A fuel plate never sees a lathe, milling machine, or any other tool of that nature. I have just a few viewgraphs here that will illustrate some of the types of tooling we use to keep away from machining and get high production at as low as possible cost. Figure I shows weighing aluminum powder. It's done in a glove box more to keep air flow away from the balance than any other reason. The weighing of the U 3 O 8 is similar and the glove box is for personnel protection. Figure 2 shows our blender, and I won't try to explain why it works. This is the only one we have ever found that really blends our powder and does a good job. Figure 3 shows our powder die on the press, and you can see the rectangular compact being extracted. Here is the way we make our frames in a blanking die Figure 4. You will notice there are two holes in the frame. We start off with two cores in a frame. Our lot size is 24, but twelve billets go into the furnace for preheating, at the seventh pass, we cut the two cores apart and at that point they become individual fuel plates. Figure 5 shows the loading of the compacts into the frame. We use a loose fit. We can just drop the cores into the frame with, I think, about 2 mils side clearance and it works very satisfactorily. Figure 6 shows a forming die. Once you make the investment for the fuel plate blanking die shown in Figure 7, you can blank out a fuel plate on the order of about one per minute, to size and to the tolerances required. Figure 8 shows a unique tool developed at Oak Ridge. It's a Homogeneity Scanner. It works on the principal of x-ray attenuation going through an electronic analysis

  9. The influence of prescribed fire and burn interval on fuel loads in four North Carolina forest ecosystems

    Science.gov (United States)

    M.J. Gavazzi; S.G. McNulty

    2014-01-01

    Prescribed fire is an important management tool in southern US forests, with more acres burned in the South than any other region of the US. Research from prescribed fire studies shows high temporal and spatial variability in available fuel loads due to physiographic, edaphic, meteorological and biological factors. In an effort to account for parts of this variation...

  10. Sodium test of the Super-Phenix full size primary pump shaft on the CPV-1 test rig at ENEA-Brasimone

    International Nuclear Information System (INIS)

    Contardi, T.; Rapezzi, L.; Partiti, C.; Zola, M.; Denimal, P.

    1984-01-01

    Tests on FBR Superphenix primary pump shaft were performed within the sodium-cooled FBR common research and development programs provided for by the cooperation agreement between ENEA and CEA. These tests were performed in CPV-1 plant ENEA - Brasimone Energy Research Center. The CPV-1 rig was built by FIAT-TTG and reproduces the reactor operating conditions (sodium-temperature and level, shaft inclination, etc..). Furthermore, CPV-1 rig's most interesting feature is its possibility to apply seismic stresses to test section by means of an oleodynamic actuator. Pivoterie-1 test section was made by JEUMONT-SCHNEIDER which built Superphenix pumps too; it was given to ENEA by FIAT-TTG. Seismic tests were performed with the cooperation of ISMES and FIAT-TTG. (author)

  11. Active load current sharing in fuel cell and battery fed DC motor drive for electric vehicle application

    International Nuclear Information System (INIS)

    Pany, Premananda; Singh, R.K.; Tripathi, R.K.

    2016-01-01

    Highlights: • Load current sharing in FC and battery fed dc drive. • Active current sharing control using LabVIEW. • Detail hardware implementation. • Controller performance is verified through MATLAB simulation and experimental results. - Abstract: In order to reduce the stress on fuel cell based hybrid source fed electric drive system the controller design is made through active current sharing (ACS) technique. The effectiveness of the proposed ACS technique is tested on a dc drive system fed from fuel cell and battery energy sources which enables both load current sharing and source power management. High efficiency and reliability of the hybrid system can be achieved by proper energy conversion and management of power to meet the load demand in terms of required voltage and current. To overcome the slow dynamics feature of FC, a battery bank of adequate power capacity has to be incorporated as FC voltage drops heavily during fast load demand. The controller allows fuel cell to operate in normal load region and draw the excess power from battery. In order to demonstrate the performance of the drive using ACS control strategy different modes of operation of the hybrid source with the static and dynamic behavior of the control system is verified through simulation and experimental results. This control scheme is implemented digitally in LabVIEW with PCI 6251 DAQ I/O interface card. The efficacy of the controller performance is demonstrated in system changing condition supplemented by experimental validation.

  12. Catalase measurement: A new field procedure for rapidly estimating microbial loads in fuels and water-bottoms

    Energy Technology Data Exchange (ETDEWEB)

    Passman, F.J. [Biodeterioration Control Associates, Inc., Chicago, IL (United States); Daniels, D.A. [Basic Fuel Services, Dover, NJ (United States); Chesneau, H.F.

    1995-05-01

    Low-grade microbial infections of fuel and fuel systems generally go undetected until they cause major operational problems. Three interdependent factors contribute to this: mis-diagnosis, incorrect or inadequate sampling procedures and perceived complexity of microbiological testing procedures. After discussing the first two issues, this paper describes a rapid field test for estimating microbial loads in fuels and associated water. The test, adapted from a procedure initially developed to measure microbial loads in metalworking fluids, takes advantage of the nearly universal presence of the enzyme catalase in the microbes that contaminated fuel systems. Samples are reacted with a peroxide-based reagent; liberating oxygen gas. The gas generates a pressure-head in a reaction tube. At fifteen minutes, a patented, electronic pressure-sensing device is used to measure that head-space pressure. The authors present both laboratory and field data from fuels and water-bottoms, demonstrating the excellent correlation between traditional viable test data (acquired after 48-72 hours incubation) and catalase test data (acquired after 15 min.-4 hours). We conclude by recommending procedures for developing a failure analysis data-base to enhance our industry`s understanding of the relationship between uncontrolled microbial contamination and fuel performance problems.

  13. Potential Fuel Loadings, Fire Ignitions, and Smoke Emissions from Nuclear Bursts in Megacities

    Science.gov (United States)

    Turco, R. P.; Toon, O. B.; Robock, A.; Bardeen, C.; Oman, L.; Stenchikov, G. L.

    2006-12-01

    We consider the effects of "small" nuclear detonations in modern "megacities," focusing on the possible extent of fire ignitions, and the properties of corresponding smoke emissions. Explosive devices in the multi-kiloton yield range are being produced by a growing number of nuclear states (Toon et al., 2006), and such weapons may eventually fall into the hands of terrorists. The numbers of nuclear weapons that might be used in a regional conflict, and their potential impacts on population and infrastructure, are discussed elsewhere. Here, we estimate the smoke emissions that could lead to widespread environmental effects, including large-scale climate anomalies. We find that low-yield weapons, which emerging nuclear states have been stockpiling, and which are likely to be targeted against cities in a regional war, can generate up to 100 times as much smoke per kiloton of yield as the high-yield weapons once associated with a superpower nuclear exchange. The fuel loadings in modern cities are estimated using a variety of data, including extrapolations from earlier detailed studies. The probability of ignition and combustion of fuels, smoke emission factors and radiative properties, and prompt scavenging and dispersion of the smoke are summarized. We conclude that a small regional nuclear war might generate up to 5 teragrams of highly absorbing particles in urban firestorms, and that this smoke could initially be injected into the middle and upper troposphere. These results are used to develop smoke emission scenarios for a climate impact analysis reported by Oman et al. (2006). Uncertainties in the present smoke estimates are outlined. Oman, L., A. Robock, G. L. Stenchikov, O. B. Toon, C. Bardeen and R. P. Turco, "Climatic consequences of regional nuclear conflicts," AGU, Fall 2006. Toon, O. B., R. P. Turco, A. Robock, C. Bardeen, L. Oman and G. L. Stenchikov, "Consequences of regional scale nuclear conflicts and acts of individual nuclear terrorism," AGU, Fall

  14. Landscape variation in tree regeneration and snag fall drive fuel loads in 24-year old post-fire lodgepole pine forests.

    Science.gov (United States)

    Nelson, Kellen N; Turner, Monica G; Romme, William H; Tinker, Daniel B

    2016-12-01

    Escalating wildfire in subalpine forests with stand-replacing fire regimes is increasing the extent of early-seral forests throughout the western USA. Post-fire succession generates the fuel for future fires, but little is known about fuel loads and their variability in young post-fire stands. We sampled fuel profiles in 24-year-old post-fire lodgepole pine (Pinus contorta var. latifolia) stands (n = 82) that regenerated from the 1988 Yellowstone Fires to answer three questions. (1) How do canopy and surface fuel loads vary within and among young lodgepole pine stands? (2) How do canopy and surface fuels vary with pre- and post-fire lodgepole pine stand structure and environmental conditions? (3) How have surface fuels changed between eight and 24 years post-fire? Fuel complexes varied tremendously across the landscape despite having regenerated from the same fires. Available canopy fuel loads and canopy bulk density averaged 8.5 Mg/ha (range 0.0-46.6) and 0.24 kg/m 3 (range: 0.0-2.3), respectively, meeting or exceeding levels in mature lodgepole pine forests. Total surface-fuel loads averaged 123 Mg/ha (range: 43-207), and 88% was in the 1,000-h fuel class. Litter, 1-h, and 10-h surface fuel loads were lower than reported for mature lodgepole pine forests, and 1,000-h fuel loads were similar or greater. Among-plot variation was greater in canopy fuels than surface fuels, and within-plot variation was greater than among-plot variation for nearly all fuels. Post-fire lodgepole pine density was the strongest positive predictor of canopy and fine surface fuel loads. Pre-fire successional stage was the best predictor of 100-h and 1,000-h fuel loads in the post-fire stands and strongly influenced the size and proportion of sound logs (greater when late successional stands had burned) and rotten logs (greater when early successional stands had burned). Our data suggest that 76% of the young post-fire lodgepole pine forests have 1,000-h fuel loads that exceed levels

  15. Dynamic characteristics of an automotive fuel cell system for transitory load changes

    DEFF Research Database (Denmark)

    Rabbani, Raja Abid; Rokni, Masoud

    2013-01-01

    A dynamic model of Polymer Electrolyte Membrane Fuel Cell (PEMFC) system is developed to investigate the behavior and transient response of a fuel cell system for automotive applications. Fuel cell dynamics are subjected to reactant flows, heat management and water transportation inside the fuel...

  16. Thermal-hydraulic analyses of the TN-24P cask loaded with consolidated and unconsolidated spent nuclear fuel

    International Nuclear Information System (INIS)

    Michener, T.E.; McKinnon, M.A.; Rector, D.R.; Creer, J.M.

    1989-06-01

    This paper presents the results of comparisons of COBRA-SFS (spent fuel storage) temperature predictions with experimental data from the TN-24P (Transnuclear) spent fuel storage cask loaded with unconsolidated and consolidated spent PWR fuel. Peak cladding temperature predictions using the COBRA-SFS code are compared with test data and predicted axial and radial temperature distributions are compared with measured temperature profiles. The pre-test accuracy of the COBRA-SFS code in predicting temperature distributions is discussed, along with the effect of post-test model improvements on temperature predictions. This paper also briefly describes the COBRA-SFS code, which is designed to accurately predict flow and temperature distributions in spent nuclear fuel storage and transportation systems. 6 refs., 14 figs

  17. Fabrication of high-uranium-loaded U/sub 3/O/sub 8/-Al developmental fuel plates

    Energy Technology Data Exchange (ETDEWEB)

    Copeland, G.L.; Martin, M.M.

    1980-12-01

    A common plate-type fuel for research and test reactors is U/sub 3/O/sub 8/ dispersed in aluminum and clad with an aluminum alloy. There is an impetus to reduce the /sup 235/U enrichment from above 90% to below 20% for these fuels to lessen the risk of diversion of the uranium for nonpeaceful uses. Thus, the uranium content of the fuel plates has to be increased to maintain the performance of the reactors. This paper describes work at ORNL to determine the maximal uranium loading for these fuels that can be fabricated with commercially proven materials and techniques and that can be expected to perform satisfactorily in service.

  18. Fuel rod analysis to respond to high burnup and demanding loading requirements. Probabilistic methodology recovers design margins narrowed by degrading fuel thermal conductivity and progressing FGR

    Energy Technology Data Exchange (ETDEWEB)

    Eberle, R; Heins, L; Sontheimer, F [Siemens AG Unternehmensbereich KWU, Erlangen (Germany)

    1997-08-01

    The proof that fuel rods will safely withstand all loads arising from inpile service conditions is generally achieved through the assessment of a number of design criteria by using a conservative analysis methodology in conjunction with design limits ``on the safe side``. The classical approach is the application of a fuel rod code to the Worst Case which is defined by the combination of most unfavorable conditions and assumptions with respect to the criterion under consideration. As it is evident that the deterministic construction of such Worst Cases imply an (unknown but) intuitively very high degree of conservatism, it is not surprising that this will develop to cause problems the more demanding fuel insertion conditions have to be anticipated (increased burnup, high efficiency loading schemes, etc.). A certain relief can be gained form cautious revisions of single design limits based on grown performance experience. But this increase of knowledge allows as well to change the established deterministic ``go/no-go`` conception into a better differentiating assessment methodology by which the quantification of the implied conservatism and the remaining design margins is possible: the Probabilistic Design Methodology (PDM). Principles and elements of the PDM are described. An essential prerequisite is a best-estimate fuel rod code which incorporates the latest state of knowledge about potential performance limiting phenomena (e.g. burnup degradation of fuel oxide thermal conductivity) as Siemens/KWU`s CARO-E does. An example is given how input distributions for rod data and model parameters transfer into a frequency distribution of maximum rod internal pressure, and indications are given how this is to be interpreted in view of a probabilistically re-formulated design criterion. The PDM provides a realistic conservative assessment of design criteria and will thus recover design margins for increasingly aggravated loading conditions. (author). 9 refs, 9 figs, 2 tabs.

  19. Fuel rod analysis to respond to high burnup and demanding loading requirements. Probabilistic methodology recovers design margins narrowed by degrading fuel thermal conductivity and progressing FGR

    International Nuclear Information System (INIS)

    Eberle, R.; Heins, L.; Sontheimer, F.

    1997-01-01

    The proof that fuel rods will safely withstand all loads arising from inpile service conditions is generally achieved through the assessment of a number of design criteria by using a conservative analysis methodology in conjunction with design limits ''on the safe side''. The classical approach is the application of a fuel rod code to the Worst Case which is defined by the combination of most unfavorable conditions and assumptions with respect to the criterion under consideration. As it is evident that the deterministic construction of such Worst Cases imply an (unknown but) intuitively very high degree of conservatism, it is not surprising that this will develop to cause problems the more demanding fuel insertion conditions have to be anticipated (increased burnup, high efficiency loading schemes, etc.). A certain relief can be gained form cautious revisions of single design limits based on grown performance experience. But this increase of knowledge allows as well to change the established deterministic ''go/no-go'' conception into a better differentiating assessment methodology by which the quantification of the implied conservatism and the remaining design margins is possible: the Probabilistic Design Methodology (PDM). Principles and elements of the PDM are described. An essential prerequisite is a best-estimate fuel rod code which incorporates the latest state of knowledge about potential performance limiting phenomena (e.g. burnup degradation of fuel oxide thermal conductivity) as Siemens/KWU's CARO-E does. An example is given how input distributions for rod data and model parameters transfer into a frequency distribution of maximum rod internal pressure, and indications are given how this is to be interpreted in view of a probabilistically re-formulated design criterion. The PDM provides a realistic conservative assessment of design criteria and will thus recover design margins for increasingly aggravated loading conditions. (author). 9 refs, 9 figs, 2 tabs

  20. Pellet-press-to-sintering-boat nuclear fuel pellet loading system

    International Nuclear Information System (INIS)

    Bucher, G.D.

    1988-01-01

    This patent describes a system for loading nuclear fuel pellets into a sintering boat from a pellet press which ejects newly made the pellets from a pellet press die table surface. The system consists of: (a) a bowl having an inner surface, a longitudinal axis, an open and generally circular top of larger diameter, and an open and generally circular bottom of smaller diameter; (b) means for supporting the bowl in a generally upright position such that the bowl is rotatable about its longitudinal axis; (c) means for receiving the ejected pellets proximate the die table surface of the pellet press and for discharging the received pellets into the bowl at a location proximate the inner surface towards the top of the bowl with a pellet velocity having a horizontal component which is generally tangent to the inner surface of the bowl proximate the location; (d) means for rotating the bowl about the longitudinal axis such that the bowl proximate the location has a velocity generally equal, in magnitude and direction, to the horizontal component of the pellet velocity at the location; and (e) means for moving the sintering boat generally horizontally beneath and proximate the bottom of the bowl

  1. Experimental and theoretical analysis of the combustion process at low loads of a diesel natural gas dual-fuel engine

    International Nuclear Information System (INIS)

    Li, Weifeng; Liu, Zhongchang; Wang, Zhongshu

    2016-01-01

    To construct an effective method to analyze the combustion process of dual fuel engines at low loads, effects of combustion boundaries on the combustion process of an electronically controlled diesel natural gas dual-fuel engine at low loads were investigated. Three typical combustion modes, including h, m and n, appeared under different combustion boundaries. In addition, the time-sequenced characteristic and the heat release rate-imbalanced characteristic were found in the dual fuel engine combustion process. To quantify these characteristics, two quantitative indicators, including the TSC (time-sequenced coefficient) and the HBC (HRR-balanced coefficient) were defined. The results show that increasing TSC and HBC can decrease HC (hydrocarbon) emissions and improve the BTE (brake thermal efficiency) significantly. The engine with the n combustion mode can obtain the highest BTE and the lowest HC emissions, followed by m, and then h. However, the combustion process of the engine will deteriorate sharply if boundary conditions are not strictly controlled in the n combustion mode. Based on the n combustion mode, advancing the start of diesel injection significantly, using large EGR (exhaust gas recirculation) rate and appropriately intake throttling can effectively reduce HC emissions and improve the BTE of dual fuel engines at low loads with relatively high natural gas PES (percentage energy substitution). - Highlights: • We reported three typical combustion modes of a dual-fuel engine at low loads. • Time-sequenced characteristic was put forward and qualified. • HRR-imbalanced characteristic was put forward and qualified. • Three combustion modes appeared as equivalence ratio/diesel injection timing varied. • The engine performance varied significantly with different combustion mode.

  2. Puget Sound Area Electric Reliability Plan. Appendix D, Conservation, Load Management and Fuel Switching Analysis : Draft Environmental Impact Statement.

    Energy Technology Data Exchange (ETDEWEB)

    United States. Bonneville Power Administration.

    1991-09-01

    Various conservation, load management, and fuel switching programs were considered as ways to reduce or shift system peak load. These programs operate at the end-use level, such as residential water heat. Figure D-1a shows what electricity consumption for water heat looks like on normal and extreme peak days. Load management programs, such as water heat control, are designed to reduce electricity consumption at the time of system peak. On the coldest day in average winter, system load peaks near 8:00 a.m. In a winter with extremely cold weather, electricity consumption increases fr all hours, and the system peak shifts to later in the morning. System load shapes in the Puget Sound area are shown in Figure D-1b for a normal winter peak day (February 2, 1988) and extreme peak day (February 3, 1989). Peak savings from any program are calculated to be the reduction in loads on the entire system at the hour of system peak. Peak savings for all programs are measured at 8:00 a.m. on a normal peak day and 9:00 a.m. on an extreme peak day. On extremely cold day, some water heat load shifts to much later in the morning, with less load available for shedding at the time of system peak. Models of hourly end-use consumption were constructed to simulate the impact of conservation, land management, and fuel switching programs on electricity consumption. Javelin, a time-series simulating package for personal computers, was chosen for the hourly analysis. Both a base case and a program case were simulated. 15 figs., 7 tabs.

  3. A tri-generation system based on polymer electrolyte fuel cell and desiccant wheel – Part A: Fuel cell system modelling and partial load analysis

    International Nuclear Information System (INIS)

    Najafi, Behzad; De Antonellis, Stefano; Intini, Manuel; Zago, Matteo; Rinaldi, Fabio; Casalegno, Andrea

    2015-01-01

    Highlights: • A mathematical model for a PEMFC based cogeneration system is developed. • Developed model is validated using the available experimental data. • Performance of the plant at full load conditions is investigated. • Performance indices while applying two different modifications are determined. • System’s performance with and without modifications at partial loads is investigated. - Abstract: Polymer Electrolyte Membrane Fuel Cell (PEMFC) based systems have recently received increasing attention as a viable alternative for meeting the residential electrical and thermal demands. However, as the intermittent demand profiles of a building can only be addressed by a tri-generative unit which can operate at partial loads, the variation of performance of the system at partial loads might affect its corresponding potential benefits significantly. Nonetheless, no previous study has been carried out on assessing the performance of this type of tri-generative systems in such conditions. The present paper is the first of a two part study dedicated to the investigation of the performance of a tri-generative system in which a PEMFC based system is coupled with a desiccant wheel unit. This study is focused on evaluating the performance of the PEMFC subsystem while operating at partial loads. Accordingly, a detailed mathematical model of the fuel cell subsystem is first developed and validated using the experimental data obtained from the plant’s and the fuel cell stack’s manufacturer. Next, in order to increase the performance of the plant, two modifications have been proposed and the resulting performance at partial load have been determined. The obtained results demonstrate that applying both modifications results in increasing the electrical efficiency of the plant by 5.5%. It is also shown that, while operating at partial loads, the electrical efficiency of the plant does not significantly change; the fact which corresponds to the trade-off between

  4. High-uranium-loaded U3O8-Al fuel element development program [contributed by N.M. Martin, ORNL

    International Nuclear Information System (INIS)

    Martin, M.M.

    1993-01-01

    The High-Uranium-Loaded U 3 O 8 -Al Fuel Element Development Program supports Argonne National Laboratory efforts to develop high-uranium-density research and test reactor fuel to accommodate use of low-uranium enrichment. The goal is to fuel most research and test reactors with uranium of less than 20% enrichment for the purpose of lowering the potential for diversion of highly-enriched material for nonpeaceful usages. The specific objective of the program is to develop the technological and engineering data base for U 3 O 8 -Al plate-type fuel elements of maximal uranium content to the point of vendor qualification for full scale fabrication on a production basis. A program and management plan that details the organization, supporting objectives, schedule, and budget is in place and preparation for fuel and irradiation studies is under way. The current programming envisions a program of about four years duration for an estimated cost of about two million dollars. During the decades of the fifties and sixties, developments at Oak Ridge National Laboratory led to the use of U 3 O 8 -Al plate-type fuel elements in the High Flux Isotope Reactor, Oak Ridge Research Reactor, Puerto Rico Nuclear Center Reactor, and the High Flux Beam Reactor. Most of the developmental information however applies only up to a uranium concentration of about 55 wt % (about 35 vol % U 3 O 8 ). The technical issues that must be addressed to further increase the uranium loading beyond 55 wt % involve plate fabrication phenomena of voids and dogboning, fuel behavior under long irradiation, and potential for the thermite reaction between U 3 O 8 and aluminum. (author)

  5. Transportation of 33 irradiated MTR fuel assemblies from FRM/Garching to Savannah River Site, USA, using a GNS transport cask and using a new loading device

    International Nuclear Information System (INIS)

    Dreesen, K.; Goetze, H.G.; Holst, L.; Gerstenberg, H.; Schreckenbach, K.

    2000-01-01

    According to the Department of Energy program of the return spent fuel from the foreign research reactors operators, 33 irradiated MTR box shaped fuel assemblies from the Technical University Munich were shipped to SRS/USA. The fuel assemblies were irradiated for typically 800 full days and, after a sufficient cooling time, loaded into a GNS 16 cask. The GNS 16 cask is a new transport cask for box shaped MTR fuel assemblies and TRIGA fuel assemblies and was used for the first time at the FRM Garching. The capacity of the cask is 33 box shaped MTR fuel assemblies. During the loading of the fuel assemblies, a newly developed loading device was used. The main components of the loading device are the transfer flask, the shielded loading lock, adapter plate and a mobile water tank. The loading device works mechanically with manpower. For the handling of the transfer flask, a crane with a capacity of 5 metric tons is necessary. During installation of the lid the mobile water pool is filled with demineralized water and the shielded loading passage is taken away. After that the lid is put on the cask. After drainage, the mobile water pool is disassembled, and the cask is dewatered. Finally leak tests of all seals are made. The achieved leakage rate was -5 Pa x I/s. The work in FRM was done between 03.02.99 and 12.02.99 including a dry run and leak test. (author)

  6. Fission gas release behavior of MOX fuels under simulated daily-load-follow operation condition. IFA-554/555 test evaluation with FASTGRASS code

    International Nuclear Information System (INIS)

    Ikusawa, Yoshihisa; Ozawa, Takayuki

    2008-03-01

    IFA-554/555 load-follow tests were performed in HALDEN reactor (HBWR) to study the MOX fuel behavior under the daily-load-follow operation condition in the framework of ATR-MOX fuel development in JAEA. IFA-554/555 rig had the instruments of rod inner pressure, fuel center temperature, fuel stack elongation, and cladding elongation. Although the daily-load-follow operation in nuclear power plant is one of the available options for economical improvement, the power change in a short period in this operation causes the change of thermal and mechanical irradiation conditions. In this report, FP gas release behavior of MOX fuel rod was evaluated under the daily-load-follow operation condition with the examination data from IFA-554/555 by using the computation code 'FASTGRASS'. From the computation results of FASTGRASS code which could compute the FP gas release behavior under the transient condition, it could be concluded that FP gas was released due to the relaxation of fuel pellet inner stress and pellet temperature increase, which were caused by the cyclic power change during the daily-load-follow operation. In addition, since the amount of released FP gas decreased during the steady operation after the daily-load-follow, it could be mentioned that the total of FP gas release at the end of life with the daily-load-follow is not so much different from that without the daily-load-follow. (author)

  7. Fission product release assessment for end fitting failure in Candu reactor loaded with CANFLEX-NU fuel bundles

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Dirk Joo; Jeong, Chang Joon; Lee, Kang Moon; Suk, Ho Chun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    Fission product release (FPR) assessment for End Fitting Failure (EFF) in CANDU reactor loaded with CANFLEX-natural uranium (NU) fuel bundles has been performed. The predicted results are compared with those for the reactor loaded with standard 37-element bundles. The total channel I-131 release at the end of transient for EFF accident is calculated to be 380.8 TBq and 602.9 TBq for the CANFLEX bundle and standard bundle channel cases, respectively. They are 4.9% and 7.9% of total inventory, respectively. The lower total releases of the CANFLEX bundle O6 channel are attributed to the lower initial fuel temperatures caused by the lower linear element power of the CANFLEX bundle compared with the standard bundle. 4 refs., 1 fig., 4 tabs. (Author)

  8. Fission product release assessment for end fitting failure in Candu reactor loaded with CANFLEX-NU fuel bundles

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Dirk Joo; Jeong, Chang Joon; Lee, Kang Moon; Suk, Ho Chun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    Fission product release (FPR) assessment for End Fitting Failure (EFF) in CANDU reactor loaded with CANFLEX-natural uranium (NU) fuel bundles has been performed. The predicted results are compared with those for the reactor loaded with standard 37-element bundles. The total channel I-131 release at the end of transient for EFF accident is calculated to be 380.8 TBq and 602.9 TBq for the CANFLEX bundle and standard bundle channel cases, respectively. They are 4.9% and 7.9% of total inventory, respectively. The lower total releases of the CANFLEX bundle O6 channel are attributed to the lower initial fuel temperatures caused by the lower linear element power of the CANFLEX bundle compared with the standard bundle. 4 refs., 1 fig., 4 tabs. (Author)

  9. Studies of fuel loading pattern optimization for a typical pressurized water reactor (PWR) using improved pivot particle swarm method

    International Nuclear Information System (INIS)

    Liu, Shichang; Cai, Jiejin

    2012-01-01

    Highlights: ► The mathematical model of loading pattern problems for PWR has been established. ► IPPSO was integrated with ‘donjon’ and ‘dragon’ into fuel arrangement optimizing code. ► The novel method showed highly efficiency for the LP problems. ► The core effective multiplication factor increases by about 10% in simulation cases. ► The power peaking factor decreases by about 0.6% in simulation cases. -- Abstract: An in-core fuel reload design tool using the improved pivot particle swarm method was developed for the loading pattern optimization problems in a typical PWR, such as Daya Bay Nuclear Power Plant. The discrete, multi-objective improved pivot particle swarm optimization, was integrated with the in-core physics calculation code ‘donjon’ based on finite element method, and assemblies’ group constant calculation code ‘dragon’, composing the optimization code for fuel arrangement. The codes of both ‘donjon’ and ‘dragon’ were programmed by Institute of Nuclear Engineering of Polytechnique Montréal, Canada. This optimization code was aiming to maximize the core effective multiplication factor (Keff), while keeping the local power peaking factor (Ppf) lower than a predetermined value to maintain fuel integrity. At last, the code was applied to the first cycle loading of Daya Bay Nuclear Power Plant. The result showed that, compared with the reference loading pattern design, the core effective multiplication factor increased by 9.6%, while the power peaking factor decreased by 0.6%, meeting the safety requirement.

  10. Direct sorbitol proton exchange membrane fuel cell using moderate catalyst loadings

    International Nuclear Information System (INIS)

    Oyarce, Alejandro; Gonzalez, Carlos; Lima, Raquel Bohn; Lindström, Rakel Wreland; Lagergren, Carina; Lindbergh, Göran

    2014-01-01

    Highlights: •The performance of a direct sorbitol fuel cell was evaluated at different temperatures. •The performance was compared to the performance of a direct glucose fuel cell. •The mass specific peak power density of the direct sorbitol fuel cell was 3.6 mW mg −1 totalcatalystloading at 80 °C. •Both sorbitol and glucose fuel cell suffer from deactivation. -- Abstract: Recent progress in biomass hydrolysis has made it interesting to study the use of sorbitol for electricity generation. In this study, sorbitol and glucose are used as fuels in proton exchange membrane fuel cells having 0.9 mg cm −2 PtRu/C at the anode and 0.3 mg cm −2 Pt/C at the cathode. The sorbitol oxidation was found to have slower kinetics than glucose oxidation. However, at low temperatures the direct sorbitol fuel cell shows higher performance than the direct glucose fuel cell, attributed to a lower degree of catalyst poisoning. The performance of both fuel cells is considerably improved at higher temperatures. High temperatures lower the poisoning, allowing the direct glucose fuel cell to reach a higher performance than the direct sorbitol fuel cell. The mass specific peak power densities of the direct sorbitol and direct glucose fuel cells at 65 °C was 3.2 mW mg −1 catalyst and 3.5 mW mg −1 catalyst , respectively. Both of these values are one order of magnitude larger than mass specific peak power densities of earlier reported direct glucose fuel cells using proton exchange membranes. Furthermore, both the fuel cells showed a considerably decrease in performance with time, which is partially attributed to sorbitol and glucose crossover poisoning the Pt/C cathode

  11. Fuel management inside the reactor. Report of generation of the nuclear bank for the fuel of the initial load of the Laguna Verde U-1 reactor with the FMS codes

    International Nuclear Information System (INIS)

    Alonso V, G.; Torres A, C.

    1991-06-01

    In this work in a general way the form in that it was generated the database of the initial fuel load of the Laguna Verde Unit 1 reactor is described. The initial load is formed with fuel of the GE6 type. The obtained results during the formation of the database in as much as to the behavior of the different cell parameters regarding the one burnt of the fuel and the variation of vacuums in the coolant channel its are compared very favorably with those reported by the General Electric fuel supplier and reported in the design documents of the same one. (Author)

  12. Fabrication of high-uranium-loaded U{sub 3}O{sub 8}-Al developmental fuel plates

    Energy Technology Data Exchange (ETDEWEB)

    Copeland, G L; Martin, M M [Oak Ridge National Laboratory, TN (United States)

    1983-08-01

    A common plate-type fuel for Research and Test Reactors (RERTR) is U{sub 3}0{sub 8} dispersed in aluminum and clad with an aluminum alloy. There is an impetus to reduce the {sup 235}U enrichment from above 90% to below 20% for these fuels to lessen the risk of diversion of the uranium for non-peaceful uses. Thus, the uranium content of the fuel plates has to be increased to maintain the performance of the reactors. This paper describes work at ORNL to determine the maximal uranium loading for these fuels that can be fabricated with commercially proven materials and techniques and that can be expected to perform satisfactorily in service. We fabricated developmental fuel plates with cores containing from 60 to 100 wt U{sub 3}0{sub 8} in aluminum encapsulated in 6061 aluminum alloy and evaluated them for aspects of fabricability, nondestructive testing, and expected performance. We recommend 75 wt U{sub 3}0{sub 8}-Al 3.1 Mg U/m{sup 3}) as the highest loading in the initial irradiation test. This upper limit is based on a qualitative assessment of the mechanical integrity of the core made by using current fabrication techniques and materials. As the oxide loading is increased beyond this point, planar areas and extensive stringers of oxide and voids develop, which leave little strength in the thickness direction. Fuel plates may then blister over these areas as fission gases collect during irradiation. Current size plates are easily fabricable to the 75 wt % U{sub 3}0{sub 8}-Al core loading by current fabrication techniques. Dogboning is a potential problem at this loading for some applications; however, this can be easily solved by using tapered compact ends. Current nondestructive radiography and transmission x-ray scanning are applicable to the highly loaded plates. Ultrasonic testing for non-bonds is marginal because of the abrupt change in conductance at the cladding-core interface. Plate thickness can be increased if desired; we fabricated 75 wt % plates with

  13. The development of fuel pins and material specimens mixed loading irradiation test rig in the experimental fast reactor Joyo. The development of the fuel-material hybrid rig

    International Nuclear Information System (INIS)

    Oyamatsu, Yasuko; Someya, Hiroyuki

    2013-02-01

    In the experimental fast reactor Joyo, there were many tests using the irradiation rigs that it was possible to be set irradiation conditions for each compartment independently. In case of no alternative fuel element to irradiate after unloading the irradiated compartments, the irradiation test was restarted with the dummy compartment which the fuel elements was not mounted. If the material specimens are mounted in this space, it is possible to use the irradiation space effectively. For these reasons, the irradiation rig (hybrid rig) is developed that is consolidated with material specimens compartment and fuel elements compartment. Fuel elements and material specimens differ greatly with heat generation, so that the most important issue in developing of hybrid rig is being able to distribute appropriately the coolant flow which satisfies irradiation conditions. The following is described by this report. (1) It was confirmed that the flow distribution of loading the same irradiation rig with the compartment from which a flow demand differs could be satisfied. (2) It was confirmed that temperature setting range of hybrid rig could be equivalent to that of irradiation condition. (3) By standardizing the coolant entrance structure of the compartment lower part, the prospect which can perform easily recombination of the compartment from which a type differs between irradiation rigs was acquired. (author)

  14. Part-load performance and emissions of a spark ignition engine fueled with RON95 and RON97 gasoline: Technical viewpoint on Malaysia’s fuel price debate

    International Nuclear Information System (INIS)

    Mohamad, Taib Iskandar; How, Heoy Geok

    2014-01-01

    Highlights: • Recent Malaysia’s gasoline price hike affects mass perception and vehicle sales. • Effects of RON95 and RON97 on a representative engine was experimentally studied. • RON95 produced better torque, power, fuel efficiency and lower NO x . • RON97 gasoline resulted in lower BSFC and lower emissions of CO 2 , CO and HC. • Performance-emission-price cross-analysis indicated RON95 as the better option. - Abstract: Due to world crude oil price hike in the recent years, many countries have experienced increase in gasoline price. In Malaysia, where gasoline are sold in two grades; RON95 and RON97, and fuel price are regulated by the government, gasoline price have been gradually increased since 2009. Price rise for RON97 is more significant. By 2014, its per liter price is 38% more than that of RON95. This has resulted in escalated dissatisfaction among the mass. People argued they were denied from using a better fuel (RON97). In order to evaluate the claim, there is a need to investigate engine response to these two gasoline grades. The effect of gasoline RON95 and RON97 on performance and exhaust emissions in spark ignition engine was investigated on a representative engine: 1.6L, 4-cylinder Mitsubishi 4G92 engine with CR 11:1. The engine was run at constant speed between 1500 and 3500 rpm with 500 rpm increment at various part-load conditions. The original engine ECU, a hydraulic dynamometer and control, a combustion analyzer and an exhaust gas analyzer were used to determine engine performance, cylinder pressure and emissions. Results showed that RON95 produced higher engine performance for all part-load conditions within the speed range. RON95 produced on average 4.4% higher brake torque, brake power, brake mean effective pressure as compared to RON97. The difference in engine performance was more significant at higher engine speed and loads. Cylinder pressure and ROHR were evaluated and correlated with engine output. With RON95, the engine

  15. Reliability of the fuel identification procedure used by COGEMA during cask loading for shipment to LA HAGUE

    International Nuclear Information System (INIS)

    Pretesacque, P.; Eid, M.; Zachar, M.

    1993-01-01

    This study has been carried out to demonstrate the reliability of the system of the spent fuel identification used by COGEMA and NTL prior to shipment to the reprocessing plant of La Hague. This was a prerequisite for the French competent authority to accept the 'burnup credit' assumption in the criticality assessment of spent fuel packages. The probability to load a non-irradiated and non-specified fuel assembly was considered as acceptable if our identification and irradiation status measurement procedures were used. Furthermore, the task analysis enabled us to improve the working conditions at reactor sites, the quality of the working documentation, and consequently to improve the reliability of the system. The NTL experience of transporting to La Hague, as consignor, more than 10,000 fuel assemblies since the date of implementation of our system in 1984 without any non-conformance on fuel identification, validated the formalism of this study as well as our assumptions on basic events probabilities. (J.P.N.)

  16. Loading of fuel and reflector elements in the Fort St. Vrain initial core (results of start-up test A-1)

    International Nuclear Information System (INIS)

    Marshall, A.C.; Brown, J.R.

    1974-01-01

    R A description is given of the experimental equipment and techniques used in the fuel and reflector loading. The analysis methods are described and test data are compared with predicted results. (U.S.)

  17. Upper limits to americium concentration in large sized sodium-cooled fast reactors loaded with metallic fuel

    International Nuclear Information System (INIS)

    Zhang, Youpeng; Wallenius, Janne

    2014-01-01

    Highlights: • The americium transmutation capability of Integral Fast Reactor was investigated. • The impact from americium introduction was parameterized by applying SERPENT Monte Carlo calculations. • Higher americium content in metallic fuel leads to a power penalty, preserving consistent safety margins. - Abstract: Transient analysis of a large sized sodium-cooled reactor loaded with metallic fuel modified by different fractions of americium have been performed. Unprotected loss-of-offsite power, unprotected loss-of-flow and unprotected transient-over-power accidents were simulated with the SAS4A/SASSYS code based on the geometrical model of an IFR with power rating of 2500 MW th , using safety parameters obtained with the SERPENT Monte Carlo code. The Ti-modified austenitic D9 steel, having higher creep rupture strength, was considered as the cladding and structural material apart from the ferritic/martensitic HT9 steel. For the reference case of U–12Pu–1Am–10Zr fuel at EOEC, the margin to fuel melt during a design basis condition UTOP is about 50 K for a maximum linear rating of 30 kW/m. In order to maintain a margin of 50 K to fuel failure, the linear power rating has to be reduced by ∼3% and 6% for 2 wt.% and 3 wt.% Am introduction into the fuel respectively. Hence, an Am concentration of 2–3 wt.% in the fuel would lead to a power penalty of 3–6%, permitting a consumption rate of 3.0–5.1 kg Am/TW h th . This consumption rate is significantly higher than the one previously obtained for oxide fuelled SFRs

  18. First fuel re-load of Angra-1 reactor - Inspection and hearing plan

    International Nuclear Information System (INIS)

    Pollis, W.; Alvarenga, M.A.B.; Meldonian, N.L.; Paiva, R.L.C. de; Pollis, R.

    1985-01-01

    The plan of inspection and hearing of the first fuel reload of Angra-1 nuclear reactor is detailed. It consists in five steps: receiving and storage of the fuel; reload preparation; activities during; post-reload activities, and preliminary activities. (M.I.)

  19. Remotely sensed measurements of forest structure and fuel loads in the Pinelands of New Jersey

    Science.gov (United States)

    Nicholas Skowronski; Kenneth Clark; Ross Nelson; John Hom; Matt Patterson

    2007-01-01

    We used a single-beam, first return profiling LIDAR (Light Detection and Ranging) measurements of canopy height, intensive biometric measurements in plots, and Forest Inventory and Analysis (FIA) data to quantify forest structure and ladder fuels (defined as vertical fuel continuity between the understory and canopy) in the New Jersey Pinelands. The LIDAR data were...

  20. Experimental analysis of ethanol dual-fuel combustion in a heavy-duty diesel engine: An optimisation at low load

    International Nuclear Information System (INIS)

    Pedrozo, Vinícius B.; May, Ian; Dalla Nora, Macklini; Cairns, Alasdair; Zhao, Hua

    2016-01-01

    Highlights: • Dual-fuel combustion offers promising results on a stock heavy-duty diesel engine. • The use of split diesel injections extends the benefits of the dual-fuel mode. • Ethanol–diesel dual-fuel combustion results in high indicated efficiencies. • NOx and soot emissions are significantly reduced. • Combustion efficiency reaches 98% with an ethanol energy ratio of 53%. - Abstract: Conventional diesel combustion produces harmful exhaust emissions which adversely affect the air quality if not controlled by in-cylinder measures and exhaust aftertreatment systems. Dual-fuel combustion can potentially reduce the formation of nitrogen oxides (NOx) and soot which are characteristic of diesel diffusion flame. The in-cylinder blending of different fuels to control the charge reactivity allows for lower local equivalence ratios and temperatures. The use of ethanol, an oxygenated biofuel with high knock resistance and high latent heat of vaporisation, increases the reactivity gradient. In addition, renewable biofuels can provide a sustainable alternative to petroleum-based fuels as well as reduce greenhouse gas emissions. However, ethanol–diesel dual-fuel combustion suffers from poor engine efficiency at low load due to incomplete combustion. Therefore, experimental studies were carried out at 1200 rpm and 0.615 MPa indicated mean effective pressure on a heavy-duty diesel engine. Fuel delivery was in the form of port fuel injection of ethanol and common rail direct injection of diesel. The objective was to improve combustion efficiency, maximise ethanol substitution, and minimise NOx and soot emissions. Ethanol energy fractions up to 69% were explored in conjunction with the effect of different diesel injection strategies on combustion, emissions, and efficiency. Optimisation tests were performed for the optimum fuelling and diesel injection strategy. The resulting effects of exhaust gas recirculation, intake air pressure, and rail pressure were

  1. Analytical and experimental justification of safe operation of fuel loads of VVER reactors at Rovno NPP

    International Nuclear Information System (INIS)

    Andrianov, A.; Zagrebelny, L.

    2011-01-01

    The main task during the nuclear fuel operation us ensuring of the larger fuel burnup and as a result - reduction of the fuel constituent in the electricity production cost. The neutron-physic calculations are performed using the qualified codes BIPR-7A and PERMAK developed by Kurchatov Institute. The limits for calculated parameters are set by the Ukrainian regulations. Calculation results are documented in reports subject for independent expert review requested by the regulatory authority. In this report the following item have presented: 1) metrological check and calibration of measuring channels; 2) fuel cycles at Rivne NPP; 3) determination of experimental values of thermal-physic and neutron-physic parameters; 4) ICIS equipment check, execution of the program confirming correct connection of the temperature and neutron flux monitoring sensors; 5) monitoring of the core TPh and NPh parameters in all operating modes; 6) monitoring of the fuel condition in the core and 7) FE leak tightness monitoring at the operating reactor

  2. KfK analysis of the SUPER-PHENIX-1 control rod experiments. Pt. 1

    International Nuclear Information System (INIS)

    Giese, H.

    1991-03-01

    As proposed by the SPX-1 analysis task force, MSM (modified source multiplication) correction factors have been produced for a series of control rod configurations established in the first critical core C1D with minimum fissile loading and in the fully loaded core CMP. The report gives a complete description of the method used at KfK to produce these correction factors and summarises the evaluated experimental results obtained. The KfK method is characterized by a 'two-step-adjustment': A basic reactivity scale adjustment and a subsequent rod worth adjustment. The first adjustment was achieved by 'tuning' either the axialbuckling in the leakage term D B 2 or the average number of neutrons per fission in the production term so that the excess reactivity of the so-called 'Follower-core' with all control rods fully raised was properly reproduced. As this excess reactivity could not be directly determined by an experiment, it had to be assessed from the shut-down worth of the main control system in combination with measured fractions of the S-curve of this system. In the second adjustment, the absorber cross sections were tuned to reproduce experimental rod worths. While for the analysis of the C1D experiments, MSM correction factor calculations were performed in 2D centre-plane geometry only, the analysis of the CMP measurements employed both 2D and 3D calculations. (orig./HP)

  3. Contingency strategy for insufficient full core off load capability in spent fuel pool for Chinshan nuclear power station

    International Nuclear Information System (INIS)

    Huang, Pinghue

    2012-01-01

    The spent fuel pool (SFP) at Taiwan Power Company's (TUC's) Chinshan plant lost the full core off load (FCO) capability in 2010, even with the second SFP repacking project to expand the capacity as reported in 12PBNC. The TEPC had originally planned to move some spent fuel assemblies from SFP to dry storage facility, however, the dry storage project had seriously fell behind. Thus, it is required to address insufficient FCO capability, and the following contingency measures have been employed: The first step was to explore whether there was a specific regulatory requirement for FCO capability, and none were identified. Also, the industrial experiences were explored. The refueling strategy is changed from FCO to in-core shuffling. A feasibility evaluation performed indicates the Technical Specifications require: alternate method of decay heat removal, and verification of shutdown margin for each in vessel fuel movement. Specific methods have been successfully established. A safety evaluation for operation without FCO capability was performed, and no safety concerns were identified. The risk for operation without FCO capability was assessed. The previous operational experiences were identified. Moreover, such works are not expected in subsequent cycles. The new fuel vault is used to store new fuel assemblies. The criticality analysis has been performed and some new approaches are proposed to enhance the storage flexibility as reported in 17PBNC. An inter-unit transfer cask has been designed to transfer spent fuel from the SFP of one unit to the other. The FCO capability can be effectively extended for three more years with this consideration. The TPC discussed the contingency strategy with the ROCAEC in May 2006, and the ROCAEC's concurrence was attained. With the proposed strategy, Chinshan units have been operating smoothly

  4. Contingency strategy for insufficient full core off load capability in spent fuel pool for Chinshan nuclear power station

    Energy Technology Data Exchange (ETDEWEB)

    Huang, Pinghue [Taiwan Power Company, Taipei (China)

    2012-03-15

    The spent fuel pool (SFP) at Taiwan Power Company's (TUC's) Chinshan plant lost the full core off load (FCO) capability in 2010, even with the second SFP repacking project to expand the capacity as reported in 12PBNC. The TEPC had originally planned to move some spent fuel assemblies from SFP to dry storage facility, however, the dry storage project had seriously fell behind. Thus, it is required to address insufficient FCO capability, and the following contingency measures have been employed: The first step was to explore whether there was a specific regulatory requirement for FCO capability, and none were identified. Also, the industrial experiences were explored. The refueling strategy is changed from FCO to in-core shuffling. A feasibility evaluation performed indicates the Technical Specifications require: alternate method of decay heat removal, and verification of shutdown margin for each in vessel fuel movement. Specific methods have been successfully established. A safety evaluation for operation without FCO capability was performed, and no safety concerns were identified. The risk for operation without FCO capability was assessed. The previous operational experiences were identified. Moreover, such works are not expected in subsequent cycles. The new fuel vault is used to store new fuel assemblies. The criticality analysis has been performed and some new approaches are proposed to enhance the storage flexibility as reported in 17PBNC. An inter-unit transfer cask has been designed to transfer spent fuel from the SFP of one unit to the other. The FCO capability can be effectively extended for three more years with this consideration. The TPC discussed the contingency strategy with the ROCAEC in May 2006, and the ROCAEC's concurrence was attained. With the proposed strategy, Chinshan units have been operating smoothly.

  5. Numerical analysis of a downsized spark-ignition engine fueled by butanol/gasoline blends at part-load operation

    International Nuclear Information System (INIS)

    Scala, F.; Galloni, E.; Fontana, G.

    2016-01-01

    Highlights: • Bio-fuels will reduce the overall CO_2 emission. • The properties of butanol/gasoline–air mixtures have been determined. • A 1-D model of a SI engine has been calibrated and validated. • The butanol content reduces the combustion duration. • The optimal ignition timing slightly changes. - Abstract: In this paper, the performance of a turbocharged SI engine, firing with butanol/gasoline blends, has been investigated by means of numerical simulations of the engine behavior. When engine fueling is switched from gasoline to alcohol/gasoline mixture, engine control parameters must be adapted. The main necessary modifications in the Electronic Control Unit have been highlighted in the paper. Numerical analyses have been carried out at partial load operation and at two different engine speeds (3000 and 4000 rpm). Several n-butanol/gasoline mixtures, differing for the alcohol contents, have been analyzed. Such engine performances as torque and indicated efficiency have been evaluated. Both these characteristics decrease with the alcohol contents within the mixtures. On the contrary, when the engine is fueled by neat n-butanol, torque and efficiency reach values about 2% higher than those obtained with neat gasoline. Furthermore, the optimal spark timing, for alcohol/gasoline mixture operation, must be retarded (up to 13%) in comparison with the correspondent values of the gasoline operation. In general, engine performance and operation undergo little variations when fuel supplying is switched from gasoline to alcohol/gasoline blends.

  6. Full and part load exergetic analysis of a hybrid micro gas turbine fuel cell system based on existing components

    International Nuclear Information System (INIS)

    Bakalis, Diamantis P.; Stamatis, Anastassios G.

    2012-01-01

    Highlights: ► Hybrid SOFC/GT system based on existing components. ► Exergy analysis using AspenPlus™ software. ► Greenhouse gases emission is significantly affected by SOFC stack temperature. ► Comparison with a conventional GT of similar power. ► SOFC/GT is almost twice efficient in terms of second low efficiency and CO 2 emission. - Abstract: The paper deals with the examination of a hybrid system consisting of a pre-commercially available high temperature solid oxide fuel cell and an existing recuperated microturbine. The irreversibilities and thermodynamic inefficiencies of the system are evaluated after examining the full and partial load exergetic performance and estimating the amount of exergy destruction and the efficiency of each hybrid system component. At full load operation the system achieves an exergetic efficiency of 59.8%, which increases during the partial load operation, as a variable speed control method is utilized. Furthermore, the effects of the various performance parameters such as fuel cell stack temperature and fuel utilization factor are assessed. The results showed that the components in which chemical reactions occur have the higher exergy destruction rates. The exergetic performance of the system is affected significantly by the stack temperature. Based on the exergetic analysis, suggestions are given for reducing the overall system irreversibility. Finally, the environmental impact of the operation of the hybrid system is evaluated and compared with a similarly rated conventional gas turbine plant. From the comparison it is apparent that the hybrid system obtains nearly double exergetic efficiency and about half the amount of greenhouse gas emissions compared with the conventional plant.

  7. Manganese-Loaded Activated Carbon for the Removal of Organosulfur Compounds from High-Sulfur Diesel Fuels

    OpenAIRE

    Al-Ghouti, M.A.; Al-Degs, Y.S.

    2014-01-01

    The adsorptive capacity of activated carbon (AC) is significantly enhanced toward weakly interacting organosulfur compounds (OSC) from sulfur-rich diesel fuel. Sulfur compounds are selectively removed from diesel after surface modification by manganese dioxide (MnO2). A selective surface for OSC removal was created by loading MnO2 on the surface; π-complexation between the partially filled d-orbitals of Mn4+ and the S atom is the controlling mechanism for OSC removal. Principal component anal...

  8. Load follow operation in nuclear power plants and its influence on PWR fuel behaviour

    International Nuclear Information System (INIS)

    Nagino, Y.; Miyazaki, Y.

    1980-01-01

    The contribution of nuclear power generation to our company's grid system is becoming greater each year, which makes it necessary to operate nuclear power plants with load follow mode in the near future. (author)

  9. Deterministic and probabilistic analysis of damping device resistance under impact loads from nuclear fuel container drop

    Science.gov (United States)

    Kala, J.; Bajer, M.; Barnat, J.; Smutný, J.

    2010-12-01

    Pedestrian-induced vibrations are a criterion for serviceability. This loading is significant for light-weight footbridge structures, but was established as a basic loading for the ceilings of various ordinary buildings. Wide variations of this action exist. To verify the different conclusions of various authors, vertical pressure measurements invoked during walking were performed. In the article the approaches of different design codes are also shown.

  10. Experimental investigation of dynamic performance and transient responses of a kW-class PEM fuel cell stack under various load changes

    International Nuclear Information System (INIS)

    Tang Yong; Yuan Wei; Pan Minqiang; Li Zongtao; Chen Guoqing; Li Yong

    2010-01-01

    The dynamic performance is a very important evaluation index of proton exchange membrane (PEM) fuel cells used for real application, which is mostly related with water, heat and gas management. A commercial PEM fuel cell system of Nexa module is employed to experimentally investigate the dynamic behavior and transient response of a PEM fuel cell stack and reveal involved influential factors. Five groups of dynamic tests are conducted and divided into different stage such as start-up, shut-down, step-up load, regular load variation and irregular load variation. It is observed that the external load changes the current output proportionally and reverses stack voltage accordingly. The purge operation benefits performance recovery and enhancement during a constant load and its time strongly depends on the operational current level. Overshoot and undershoot behaviors are observed during transience. But the current undershoot does not appear due to charge double-layer effect. Additionally, magnitudes of the peaks of the voltage overshoot and undershoot vary at different current levels. The operating temperature responds fast to current load but changes slowly showing an arc-like profile without any overshoot and undershoot events. The air flow rate changes directly following the dynamic load demand. But the increased amount of air flow rate during different step-change is not identical, which depends on the requirement of internal reaction and flooding intensity. The results can be utilized for validation of dynamic fuel cell models, and regarded as reference for effective control and management strategies.

  11. Modification of fuel properties under thermal load; Veraenderung von Kraftstoffeigenschaften unter thermischer Belastung

    Energy Technology Data Exchange (ETDEWEB)

    Lau, Kornelia; Richter, Beate; Schuemann, Ulrike; Crusius, Svetlana; Streibel, Thorsten W.; Harndorf, Horst [Rostock Univ. (Germany). Abt. Analytische und Technische Chemie

    2013-10-01

    Diesel fuel without FAME and additives as well as rapeseed methyl ester were thermally stressed (150 C) at laboratory conditions. In the course of the performed study, the chemical composition of the fuels and possibly generated residues were analyzed with regards to the influence of oxygen availability and test duration. Therefore, chromatographic methods as GC-MS (Gas Chromatography-Mass Spectrometry) and HP-SEC (High Performance-Size Exclusion Chromatography) have successfully been applied. In the formed solid B 0-residues mainly oxygen-containing aromatic compounds were identified, while the biodiesel samples show mostly decomposition products of the biofuel like carbonic acids and aldehydes. In both fuels an increasing amount of compounds with high molecular weights was observed versus test duration. It is assumed that these components consist of oligo- and polymers with high oxygen content. However, this needs to be proved by further analytical methods. On the basis of these results prospective tailor-made additive packages for the stabilisation of fuels and the prevention of deposits in fuel injection systems, can be developed and applied. (orig.)

  12. Boiling water reactors with uranium-plutonium mixed oxide fuel. Report 5: Analysis of the reactivity coefficients and the stability of a BWR loaded with MOx fuel

    Energy Technology Data Exchange (ETDEWEB)

    Demaziere, C. [CEA Centre d' Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Direction des Reacteurs Nucleaires

    2000-01-01

    This report is a part of the project titled 'Boiling Water Reactors With Uranium-Plutonium Mixed Oxide (MOx) Fuel'. The aim of this study is to model the impact of a core loading pattern containing MOx bundles upon the main characteristics of a BWR (reactivity coefficients, stability, etc.). For this purpose, the Core Management System (CMS) codes of Studsvik Scandpower are used. This package is constituted by CASMO-4/TABLES-3/SIMULATE-3. It has been shown in previous reports that these codes are able to accurately represent and model MOx bundles. This report is thus devoted to the study of BWR cores loaded (partially or totally) with MOx bundles. The plutonium quality used is the Pu type 2016 (mostly Pu-239, 56 %, and Pu-240, 26 %), but a variation of the plutonium isotopic vector was also investigated, in case of a partial MOx loading. One notices that the reactivity coefficients do not present significant changes in comparison with a full UOx loading. Nevertheless, two main problems arise: the shutdown margin at BOC is lower than 1 % and the stability to in-phase oscillations is slightly decreased. (The SIMULATE-3 version used for this study does not contain the latest MOx enhancements described in literature, since these code developments have not been provided to the department. Nevertheless, as the nominal average enrichment of the MOx bundles is 5.41 % (total amount of plutonium), which can still be considered as a relatively low enrichment, the accuracy of the CMS codes is acceptable without the use of the MOx improvements for this level of Pu enrichment.

  13. Study on MAs transmutation of accelerator-driven system sodium-cooled fast reactor loaded with metallic fuel

    International Nuclear Information System (INIS)

    Han Song; Yang Yongwei

    2007-01-01

    Through the analysis of the effect of heavy metal actinides on the effective multiplication constant (k eff ) of the core in accelerator-driven system (ADS) sodium-cooled fast reactor loaded with metallic fuel, we gave the method for determining fuel components. the characteristics of minor actinides (MAs) transmutation was analyzed in detail. 3D burn-up code COUPLE, which couples MCNP4c3 and ORIGEN2, was applied to the neutron simulation and burn up calculation. The results of optimized scheme shows that adjusting the proportion of 239 Pu and maintaining the value during the burn-up cycle is an efficient method of designing k eff and keeping stable during the burn-up cycle. Spallation neutrons lead to the neutron spectrum harder at inner core than that at outer core. It is in favor of improving MA's fission cross sections and the capture-to-fission ratio. The total MAs transmutation support ratio 8.3 achieves excellent transmutation effect. For higher flux at inner core leads to obvious differences on transmutation efficiency,only disposing MAs at inner core is in favor of decreasing the loading mass and improving MAs transmutation effect. (authors)

  14. Water confinement effects in response of fuel assembly to faulted condition loads

    International Nuclear Information System (INIS)

    Shah, S.J.; Brenneman, B.; Williams, G.T.; Strumpel, J.H.

    2004-01-01

    It has been established by other authors that the accelerations of the water confined by the reactor core baffle plates has a significant effect on the responses of all the fuel assemblies during LOCA (loss of coolant accident) or seismic transients. This particular effect is a consequence of the water being essentially incompressible, and thus experiencing the same horizontal accelerations as the imposed baffle plate motions. These horizontal accelerations of the fluid induce lateral pressure gradients that cause horizontal buoyancy forces on any submerged structures. These forces are in the same direction as the baffle accelerations and, for certain frequencies at least, tend to reduce the relative displacements between the fuel and baffle plates. But there is another confinement effect: the imposed baffle plate velocities must also be transmitted to the water. If the fuel assembly grid strips are treated as simple hydro-foils, these horizontal velocity components change the fluid angle of attack on each strip, and thus may induce large horizontal lift forces on each grid in the same direction as the baffle plate velocity. There is a similar horizontal lift due to inclined flow over the rods when axial flow is present. These combined forces appear to reduce the relative displacements between the fuel and baffle plates for any significant axial flow velocity. Modeling this effect is very simple. It was shown in previous papers that the mechanism for the large fuel assembly damping due to axial flow may be the hydrodynamic forces on the grid strips, and that this is very well represented by discrete viscous dampers at each grid elevation. To include the imposed horizontal water velocity effects, on both the grids and rods, these dampers are simply attached to the baffle plate rather than 'ground'. The large flow-induced damping really acts in a relative reference frame rather than an inertial reference frame, and thus it becomes a flow-induced coupling between the

  15. Characterization of Exoelectrogenic Bacteria Enterobacter Strains Isolated from a Microbial Fuel Cell Exposed to Copper Shock Load

    Science.gov (United States)

    Feng, Cuijie; Li, Jiangwei; Qin, Dan; Chen, Lixiang; Zhao, Feng; Chen, Shaohua; Hu, Hongbo; Yu, Chang-Ping

    2014-01-01

    Microorganisms capable of generating electricity in microbial fuel cells (MFCs) have gained increasing interest. Here fourteen exoelectrogenic bacterial strains were isolated from the anodic biofilm in an MFC before and after copper (Cu) shock load by Hungate roll-tube technique with solid ferric (III) oxide as an electron acceptor and acetate as an electron donor. Phylogenetic analysis of the 16S rRNA gene sequences revealed that they were all closely related to Enterobacter ludwigii DSM 16688T within the Enterobacteriaceae family, although these isolated bacteria showed slightly different morphology before and after Cu shock load. Two representative strains R2B1 (before Cu shock load) and B4B2 (after Cu shock load) were chosen for further analysis. B4B2 is resistant to 200 mg L−1 of Cu(II) while R2B1 is not, which indicated the potential selection of the Cu shock load. Raman analysis revealed that both R2B1 and B4B2 contained c-type cytochromes. Cyclic voltammetry measurements revealed that strain R2B1 had the capacity to transfer electrons to electrodes. The experimental results demonstrated that strain R2B1 was capable of utilizing a wide range of substrates, including Luria-Bertani (LB) broth, cellulose, acetate, citrate, glucose, sucrose, glycerol and lactose to generate electricity, with the highest current density of 440 mA·m−2 generated from LB-fed MFC. Further experiments indicated that the bacterial cell density had potential correlation with the current density. PMID:25412475

  16. Characterization of exoelectrogenic bacteria enterobacter strains isolated from a microbial fuel cell exposed to copper shock load.

    Directory of Open Access Journals (Sweden)

    Cuijie Feng

    Full Text Available Microorganisms capable of generating electricity in microbial fuel cells (MFCs have gained increasing interest. Here fourteen exoelectrogenic bacterial strains were isolated from the anodic biofilm in an MFC before and after copper (Cu shock load by Hungate roll-tube technique with solid ferric (III oxide as an electron acceptor and acetate as an electron donor. Phylogenetic analysis of the 16S rRNA gene sequences revealed that they were all closely related to Enterobacter ludwigii DSM 16688T within the Enterobacteriaceae family, although these isolated bacteria showed slightly different morphology before and after Cu shock load. Two representative strains R2B1 (before Cu shock load and B4B2 (after Cu shock load were chosen for further analysis. B4B2 is resistant to 200 mg L-1 of Cu(II while R2B1 is not, which indicated the potential selection of the Cu shock load. Raman analysis revealed that both R2B1 and B4B2 contained c-type cytochromes. Cyclic voltammetry measurements revealed that strain R2B1 had the capacity to transfer electrons to electrodes. The experimental results demonstrated that strain R2B1 was capable of utilizing a wide range of substrates, including Luria-Bertani (LB broth, cellulose, acetate, citrate, glucose, sucrose, glycerol and lactose to generate electricity, with the highest current density of 440 mA·m-2 generated from LB-fed MFC. Further experiments indicated that the bacterial cell density had potential correlation with the current density.

  17. Innovation of genetic algorithm code GenA for WWER fuel loading optimization

    International Nuclear Information System (INIS)

    Sustek, J.

    2005-01-01

    One of the stochastic search techniques - genetic algorithms - was recently used for optimization of arrangement of fuel assemblies (FA) in core of reactors WWER-440 and WWER-1000. Basic algorithm was modified by incorporation of SPEA scheme. Both were enhanced and some results are presented (Authors)

  18. The evaluation of meta-analysis techniques for quantifying prescribed fire effects on fuel loadings.

    Science.gov (United States)

    Karen E. Kopper; Donald McKenzie; David L. Peterson

    2009-01-01

    Models and effect-size metrics for meta-analysis were compared in four separate meta-analyses quantifying surface fuels after prescribed fires in ponderosa pine (Pinus ponderosa Dougl. ex Laws.) forests of the Western United States. An aggregated data set was compiled from eight published reports that contained data from 65 fire treatment units....

  19. Fuel pin behavior of a pressurizer water reactor with load following

    International Nuclear Information System (INIS)

    Perrotta, J.A.

    1980-10-01

    The performance of a PWR fuel pin was evaluated, during power cycles that occur in normal operations, excluding accident cases. A code to perform the mechanical analysis of the cladding was developed using the Finite Element Method to take into account local effects of pellet-cladding interaction (PCI). (E.G.) [pt

  20. Describing wildland surface fuel loading for fire management: A review of approaches, methods and systems

    Science.gov (United States)

    Robert E. Keane

    2013-01-01

    Wildland fuelbeds are exceptionally complex, consisting of diverse particles of many sizes, types and shapes with abundances and properties that are highly variable in time and space. This complexity makes it difficult to accurately describe, classify, sample and map fuels for wildland fire research and management. As a result, many fire behaviour and effects software...

  1. Plan for Structural Analysis of Fuel Assembly for Seismic and Loss of Coolant Accident Loading Considering End-Of-Life Condition for APR1400 NRC Design Certification

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dong-Hak [KHNP Central Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    The evaluation of fuel assembly structural response to externally applied forces by earthquakes and postulated pipe breaks in the reactor coolant system is described in standard review plan (SRP) 4.2, appendix A. SRP 4.2, appendix A, section III, states, 'While P(crit) [the crushing load] will increase with irradiation, ductility will be reduced. The extra margin in P(crit) for irradiated spacer grids is thus assumed to offset the unknown deformation behavior of irradiated spacer grids beyond P(crit).' The assumption in the SRP concerning irradiated grids may suggest that only the beginning-of-life (BOL) condition for spacer grid strength needs to be evaluated for fuel assembly integrity under externally applied forces. However, U.S. NRC issued the NRC. To consider the EOL conditions for the structural analysis of the fuel assembly under a seismic and LOCA loading, the simulated fuel assembly for EOL conditions should be considered by determining the gap between the spacer grid and fuel rod. Using the simulated fuel assembly, spacer grid test and fuel assembly mechanical test should be conducted to determine the simplified model of fuel assembly which is used for the structural analysis. The structural analysis will be conducted using the fuel assembly model for EOL condition. The flow damping value will be also used for the structural analysis to reduce the impact force.

  2. The effect of load factor on fission product decay heat from discharged reactor fuel

    International Nuclear Information System (INIS)

    Davies, B.S.J.

    1978-07-01

    A sum-of-exponentials expression representing the decay heat power following a burst thermal irradiation of 235 U has been used to investigate the effect of load factor during irradiation on subsequent decay heat production. A sequence of random numbers was used to indicate reactor 'on' and 'off' periods for irradiations which continued for a total of 1500 days at power and were followed by 100 days cooling. It was found that for these conditions decay heat is almost proportional to load factor. Estimates of decay heat uncertainty arising from the random irradiation pattern are also given. (author)

  3. Control and load management of a fuel cell based hybrid system; Steuerung und Lademanagement eines brennstoffzellen-basierten Hybridsystems

    Energy Technology Data Exchange (ETDEWEB)

    Klausmann, Andreas

    2011-07-01

    Objective of this work is the development of a control for a hybrid electric power train. Initial point is an electric drive powered by a rechargeable battery. This battery shall be recharged during operation by a methanol-driven fuel cell. At this point it is not intended to deploy a direct methanol fuel cell but a combination of a methanol reformer generating hydrogen-rich gas and a high-temperature fuel cell (HTPEM-FC). This work covers the general strategy of operation like load cycles, standby phases etc., the reformer control and the fuel cell operation with a newly developed charge concept. While the basic research is done on a rapid prototyping system this work aims on porting the control system to an embedded platform. Here emphasis is put on the hardware independency of the control. The development of the reformer control contains the strategy for heating up the system with a minimum of electrical energy consumption, since this energy has to be supplied from the battery during the system start-up, increasing the minimum charge level of the battery required for an autarkic recharge. Unlike in common systems the reformer will be modulated according to the electric load and not vice versa, though the fuel cell serves as load sensor. Beside start-up and shutdown strategies the fuel cell control covers particularly the charge control. The electric load is assumed to be unknown, non-influenceable and unsteady. The charge control handles the charging of the battery under optimal utilization of the available hydrogen while avoiding an overload of the fuel cell caused by sudden load changes like powering up the drive. Therefore the common step-down circuit will be advanced so that all huge and heavy electronic components can be minimized or substituted by internal effects of battery and fuel cell. The fuel utilization will be feed back to the reformer control. After coupling of reformer and fuel cell control the system will be ported to an embedded control system

  4. Alkali resistant Ni-loaded yolk-shell catalysts for direct internal reforming in molten carbonate fuel cells

    Science.gov (United States)

    Jang, Won-Jun; Hong, Young Jun; Kim, Hak-Min; Shim, Jae-Oh; Roh, Hyun-Seog; Kang, Yun Chan

    2017-06-01

    A facile and scalable spray pyrolysis process is applied to synthesize multi-shelled Ni-loaded yolk-shell catalysts on various supports (Al2O3, CeO2, ZrO2, and La(OH)3). The prepared catalysts are applied to direct internal reforming (DIR) in a molten carbonate fuel cell (MCFC). Even on exposure to alkali hydroxide vapors, the Ni-loaded yolk-shell catalysts remain highly active for DIR-MCFCs. The Ni@Al2O3 microspheres show the highest conversion (92%) of CH4 and the best stability among the prepared Ni-loaded yolk-shell catalysts. Although the initial CH4 conversion of the Ni@ZrO2 microspheres is higher than that of the Ni@CeO2 microspheres, the Ni@CeO2 microspheres are more stable. The catalytic performance is strongly dependent on the surface area and acidity and also partly dependent on the reducibility. The acidic nature of Al2O3 combined with its high surface area and yolk-shell structure enhances the adsorption of CH4 and resistance against alkali poisoning, resulting in efficient DIR-MCFC reactions.

  5. Influence of fuel assembly loading pattern and fuel burnups upon leakage neutron flux spectra from light water reactor core (Joint research)

    International Nuclear Information System (INIS)

    Kojima, Kensuke; Okumura, Keisuke; Kosako, Kazuaki; Torii, Kazutaka

    2016-01-01

    At the decommissioning of light water reactors (LWRs), it is important to evaluate an amount of radioactivity in the ex-core structures such as a reactor containment vessel, radiation shieldings, and so on. It is thought that the leakage neutron spectra in these radioactivation regions, which strongly affect the induced radioactivity, would be changed by different reactor core configurations such as fuel assembly loading pattern and fuel burnups. This study was intended to evaluate these effects. For this purpose, firstly, partial neutron currents on the core surfaces were calculated for some core configurations. Then, the leakage neutron flux spectra in major radioactivation regions were calculated based on the provided currents. Finally, influence of the core configurations upon the neutron flux spectra was evaluated. As a result, it has been found that the influence is small on the spectrum shapes of neutron fluxes. However, it is necessary to pay attention to the facts that intensities of the leakage neutron fluxes are changed by the configurations and that intensities and spectrum shapes of the leakage neutron fluxes are changed depending on the angular direction around the core. (author)

  6. Enhanced oxygen reduction activity and solid oxide fuel cell performance with a nanoparticles-loaded cathode.

    Science.gov (United States)

    Zhang, Xiaomin; Liu, Li; Zhao, Zhe; Tu, Baofeng; Ou, Dingrong; Cui, Daan; Wei, Xuming; Chen, Xiaobo; Cheng, Mojie

    2015-03-11

    Reluctant oxygen-reduction-reaction (ORR) activity has been a long-standing challenge limiting cell performance for solid oxide fuel cells (SOFCs) in both centralized and distributed power applications. We report here that this challenge has been tackled with coloading of (La,Sr)MnO3 (LSM) and Y2O3 stabilized zirconia (YSZ) nanoparticles within a porous YSZ framework. This design dramatically improves ORR activity, enhances fuel cell output (200-300% power improvement), and enables superior stability (no observed degradation within 500 h of operation) from 600 to 800 °C. The improved performance is attributed to the intimate contacts between nanoparticulate YSZ and LSM particles in the three-phase boundaries in the cathode.

  7. Fuel cell plates with skewed process channels for uniform distribution of stack compression load

    Science.gov (United States)

    Granata, Jr., Samuel J.; Woodle, Boyd M.

    1989-01-01

    An electrochemical fuel cell includes an anode electrode, a cathode electrode, an electrolyte matrix sandwiched between electrodes, and a pair of plates above and below the electrodes. The plate above the electrodes has a lower surface with a first group of process gas flow channels formed thereon and the plate below the electrodes has an upper surface with a second group of process gas flow channels formed thereon. The channels of each group extend generally parallel to one another. The improvement comprises the process gas flow channels on the lower surface of the plate above the anode electrode and the process gas flow channels on the upper surface of the plate below the cathode electrode being skewed in opposite directions such that contact areas of the surfaces of the plates through the electrodes are formed in crisscross arrangements. Also, the plates have at least one groove in areas of the surfaces thereof where the channels are absent for holding process gas and increasing electrochemical activity of the fuel cell. The groove in each plate surface intersects with the process channels therein. Also, the opposite surfaces of a bipolar plate for a fuel cell contain first and second arrangements of process gas flow channels in the respective surfaces which are skewed the same amount in opposite directions relative to the longitudinal centerline of the plate.

  8. Power ramping, cycling and load following behaviour of water reactor fuel

    International Nuclear Information System (INIS)

    1988-05-01

    The present meeting was scheduled by the International Atomic Energy Agency upon proposal of the Members of the International Working Group on Water Reactor Fuel Performance and Technology. Sixty-three participants representing 15 countries and one international organization attended the meeting. Twenty papers were presented during three technical sessions, followed by panel discussions which allowed to formulate the conclusions of the meeting and recommendations to the Agency. The objective of this Technical Committee Meeting is to review the ''State-of-the-Art'', make critical comments and recommendations with the aim of improving fuel reliability and assure integrity of the cladding and core materials when subjected to ramping and cycling sequences. The Meeting was organized in three sessions: Session 1. ''Mechanical Behaviour and Fission Gas Release'' (7 papers); Session 2. ''Power Ramping and Power Cycling Demonstration Programmes in Research Reactors'' (5 papers); Session 3. ''Fuel Behaviour in Power Reactors'' (9 papers). Between the sessions, the session chairmen, together with the speakers, prepared and presented reports with summary, conclusions and recommendations of the individual sessions. These reports are added to this summary report. A separate abstract was prepared for each of these 21 presentations. Refs, figs and tabs

  9. Optimal Load-Tracking Operation of Grid-Connected Solid Oxide Fuel Cells through Set Point Scheduling and Combined L1-MPC Control

    Directory of Open Access Journals (Sweden)

    Siwei Han

    2018-03-01

    Full Text Available An optimal load-tracking operation strategy for a grid-connected tubular solid oxide fuel cell (SOFC is studied based on the steady-state analysis of the system thermodynamics and electrochemistry. Control of the SOFC is achieved by a two-level hierarchical control system. In the upper level, optimal setpoints of output voltage and the current corresponding to unit load demand is obtained through a nonlinear optimization by minimizing the SOFC’s internal power waste. In the lower level, a combined L1-MPC control strategy is designed to achieve fast set point tracking under system nonlinearities, while maintaining a constant fuel utilization factor. To prevent fuel starvation during the transient state resulting from the output power surging, a fuel flow constraint is imposed on the MPC with direct electron balance calculation. The proposed control schemes are testified on the grid-connected SOFC model.

  10. Study of the influence of fuel load and slope on a fire spreading across a bed of pine needles by using oxygen consumption calorimetry

    Science.gov (United States)

    Tihay, V.; Morandini, F.; Santoni, P. A.; Perez-Ramirez, Y.; Barboni, T.

    2012-11-01

    A set of experiments using a Large Scale Heat Release Rate Calorimeter was conducted to test the effects of slope and fuel load on the fire dynamics. Different parameters such as the geometry of the flame front, the rate of spread, the mass loss rate and the heat release rate were investigated. Increasing the fuel load or the slope modifies the fire behaviour. As expected, the flame length and the rate of spread increase when fuel load or slope increases. The heat release rate does not reach a quasi-steady state when the propagation takes place with a slope of 20° and a high fuel load. This is due to an increase of the length of the fire front leading to an increase of fuel consumed. These considerations have shown that the heat release can be estimated with the mass loss rate by considering the effective heat of combustion. This approach can be a good alternative to estimate accurately the fireline intensity when the measure of oxygen consumption is not possible.

  11. Study of the influence of fuel load and slope on a fire spreading across a bed of pine needles by using oxygen consumption calorimetry

    International Nuclear Information System (INIS)

    Tihay, V; Morandini, F; Santoni, P A; Perez-Ramirez, Y; Barboni, T

    2012-01-01

    A set of experiments using a Large Scale Heat Release Rate Calorimeter was conducted to test the effects of slope and fuel load on the fire dynamics. Different parameters such as the geometry of the flame front, the rate of spread, the mass loss rate and the heat release rate were investigated. Increasing the fuel load or the slope modifies the fire behaviour. As expected, the flame length and the rate of spread increase when fuel load or slope increases. The heat release rate does not reach a quasi-steady state when the propagation takes place with a slope of 20° and a high fuel load. This is due to an increase of the length of the fire front leading to an increase of fuel consumed. These considerations have shown that the heat release can be estimated with the mass loss rate by considering the effective heat of combustion. This approach can be a good alternative to estimate accurately the fireline intensity when the measure of oxygen consumption is not possible.

  12. Validation Data and Model Development for Fuel Assembly Response to Seismic Loads

    International Nuclear Information System (INIS)

    Bardet, Philippe; Ricciardi, Guillaume

    2016-01-01

    Vibrations are inherently present in nuclear reactors, especially in cores and steam generators of pressurized water reactors (PWR). They can have significant effects on local heat transfer and wear and tear in the reactor and often set safety margins. The simulation of these multiphysics phenomena from first principles requires the coupling of several codes, which is one the most challenging task in modern computer simulation. Here an ambitious multiphysics multidisciplinary validation campaign is conducted. It relied on an integrated team of experimentalists and code developers to acquire benchmark and validation data for fluid-structure interaction codes. Data are focused on PWR fuel bundle behavior during seismic transients.

  13. Validation Data and Model Development for Fuel Assembly Response to Seismic Loads

    Energy Technology Data Exchange (ETDEWEB)

    Bardet, Philippe [George Washington Univ., Washington, DC (United States); Ricciardi, Guillaume [Atomic Energy Commission (CEA) (France)

    2016-01-31

    Vibrations are inherently present in nuclear reactors, especially in cores and steam generators of pressurized water reactors (PWR). They can have significant effects on local heat transfer and wear and tear in the reactor and often set safety margins. The simulation of these multiphysics phenomena from first principles requires the coupling of several codes, which is one the most challenging tasks in modern computer simulation. Here an ambitious multiphysics multidisciplinary validation campaign is conducted. It relied on an integrated team of experimentalists and code developers to acquire benchmark and validation data for fluid-structure interaction codes. Data are focused on PWR fuel bundle behavior during seismic transients.

  14. Hierarchical Load Tracking Control of a Grid-connected Solid Oxide Fuel Cell for Maximum Electrical Efficiency Operation

    DEFF Research Database (Denmark)

    Li, Yonghui; Wu, Qiuwei; Zhu, Haiyu

    2015-01-01

    efficiency operation obtained at different active power output levels, a hierarchical load tracking control scheme for the grid-connected SOFC was proposed to realize the maximum electrical efficiency operation with the stack temperature bounded. The hierarchical control scheme consists of a fast active...... power control and a slower stack temperature control. The active power control was developed by using a decentralized control method. The efficiency of the proposed hierarchical control scheme was demonstrated by case studies using the benchmark SOFC dynamic model......Based on the benchmark solid oxide fuel cell (SOFC) dynamic model for power system studies and the analysis of the SOFC operating conditions, the nonlinear programming (NLP) optimization method was used to determine the maximum electrical efficiency of the grid-connected SOFC subject...

  15. Fuel assembly

    International Nuclear Information System (INIS)

    Yamazaki, Hajime.

    1995-01-01

    In a fuel assembly having fuel rods of different length, fuel pellets of mixed oxides of uranium and plutonium are loaded to a short fuel rod. The volume ratio of a pellet-loaded portion to a plenum portion of the short fuel rod is made greater than the volume ratio of a fuel rod to which uranium fuel pellets are loaded. In addition, the volume of the plenum portion of the short fuel rod is set greater depending on the plutonium content in the loaded fuel pellets. MOX fuel pellets are loaded on the short fuel rods having a greater degree of freedom relevant to the setting for the volume of the plenum portion compared with that of a long rod fuel, and the volume of the plenum portion is ensured greater depending on the plutonium content. Even if a large amount of FP gas and He gas are discharged from the MOX fuels compared with that from the uranium fuels, the internal pressure of the MOX fuel rod during operation is maintained substantially identical with that of the uranium fuel rod, so that a risk of generating excess stresses applied to the fuel cladding tubes and rupture of fuels are greatly reduced. (N.H.)

  16. An experimental study of the dynamic behavior of a 2 kW proton exchange membrane fuel cell stack under various loading conditions

    International Nuclear Information System (INIS)

    Jian, Qifei; Zhao, Yang; Wang, Haoting

    2015-01-01

    The dynamic behavior of the PEM (proton exchange membrane) fuel cell stack has great effect on the safety and effective operation of its applications. In this paper, a self-designed bulb-array is used to simulate the various loading conditions and study the dynamic behavior of a 2 kW PEM fuel cell stack. An evaluation index, including oscillation rate, pressure variation and dynamic resistance factor, is used to analyze the transient response of the PEM fuel cell stack. It is observed that the stack current increases about 8.6%, and the Oscillation rate decreases more rapidly after activation. In the step-up load stage, the oscillation rate and the dynamic resistance decrease more rapidly as the external load increases. Due to the periodic anodic purge process, a periodic voltage fluctuation can be seen. In addition, when the stack works in the open-loop state (working without the external load), the transient response of the stack current is significantly affected by the hydrogen humidity and the charge double-layer. - Highlights: • The working time of open-loop state significantly affects the transient response. • Oscillation rate decreases faster as the external load increases. • Dynamic resistance factor decreases as the external load increases. • The periodic anodic purge process leads to a slight periodic oscillation of voltage

  17. Hierarchical Load Tracking Control of a Grid-Connected Solid Oxide Fuel Cell for Maximum Electrical Efficiency Operation

    Directory of Open Access Journals (Sweden)

    Yonghui Li

    2015-03-01

    Full Text Available Based on the benchmark solid oxide fuel cell (SOFC dynamic model for power system studies and the analysis of the SOFC operating conditions, the nonlinear programming (NLP optimization method was used to determine the maximum electrical efficiency of the grid-connected SOFC subject to the constraints of fuel utilization factor, stack temperature and output active power. The optimal operating conditions of the grid-connected SOFC were obtained by solving the NLP problem considering the power consumed by the air compressor. With the optimal operating conditions of the SOFC for the maximum efficiency operation obtained at different active power output levels, a hierarchical load tracking control scheme for the grid-connected SOFC was proposed to realize the maximum electrical efficiency operation with the stack temperature bounded. The hierarchical control scheme consists of a fast active power control and a slower stack temperature control. The active power control was developed by using a decentralized control method. The efficiency of the proposed hierarchical control scheme was demonstrated by case studies using the benchmark SOFC dynamic model.

  18. The future fuel cycle plants

    International Nuclear Information System (INIS)

    Paret, L.; Touron, E.

    2016-01-01

    The future fuel cycle plants will have to cope with both the fuel for PWR and the fuel for the new generation of fast reactors. Furthermore, the MOX fuel, that is not recycled in PWR reactors will have the possibility to be recycled in fast reactors of 4. generation. Recycling MOX fuels will imply to handle nuclear fuels with higher concentration of Pu than today. The design of the nuclear fuel for the future fast reactors will be similar to that of the Astrid prototype. In order to simplify the fabrication of UPuO_2 pellets, all the fabrication process will take place in a dedicated glove box. Enhanced reality and virtual reality technologies have been used to optimize the glove-box design in order to have a better recovery of radioactive dust and to ease routine operations and its future dismantling. As a fuel assembly will contain 120 kg of UPuO_2 fuel, it will no longer be possible to mount these assemblies by hand contrary to what was done for Superphenix reactor. A new shielded mounting line has to be designed. Another point is that additive manufacturing for the fabrication of very small parts with a complex design will be broadly used. (A.C.)

  19. Human Error Prediction and Countermeasures based on CREAM in Loading and Storage Phase of Spent Nuclear Fuel (SNF)

    International Nuclear Information System (INIS)

    Kim, Jae San; Kim, Min Su; Jo, Seong Youn

    2007-01-01

    With the steady demands for nuclear power energy in Korea, the amount of accumulated SNF has inevitably increased year by year. Thus far, SNF has been on-site transported from one unit to a nearby unit or an on-site dry storage facility. In the near future, as the amount of SNF generated approaches the capacity of these facilities, a percentage of it will be transported to another SNF storage facility. In the process of transporting SNF, human interactions involve inspecting and preparing the cask and spent fuel, loading the cask, transferring the cask and storage or monitoring the cask, etc. So, human actions play a significant role in SNF transportation. In analyzing incidents that have occurred during transport operations, several recent studies have indicated that 'human error' is a primary cause. Therefore, the objectives of this study are to predict and identify possible human errors during the loading and storage of SNF. Furthermore, after evaluating human error for each process, countermeasures to minimize human error are deduced

  20. Analysis of carbon fiber brush loading in anodes on startup and performance of microbial fuel cells

    KAUST Repository

    Hutchinson, Adam J.

    2011-11-01

    Flat carbon anodes placed near a cathode in a microbial fuel cell (MFC) are adversely affected by oxygen crossover, but graphite fiber brush anodes placed near the cathode produce high power densities. The impact of the brush size and electrode spacing was examined by varying the distance of the brush end from the cathode and solution conductivity in multiple MFCs. The startup time was increased from 8 ± 1 days with full brushes (all buffer concentrations) to 13 days (50 mM), 14 days (25 mM) and 21 days (8 mM) when 75% of the brush anode was removed. When MFCs were all first acclimated with a full brush, up to 65% of the brush material could be removed without appreciably altering maximum power. Electrochemical impedance spectroscopy (EIS) showed that the main source of internal resistance (IR) was diffusion resistance, which together with solution resistance reached 100 Ω. The IR using EIS compared well with that obtained using the polarization data slope method, indicating no major components of IR were missed. These results show that using full brush anodes avoids adverse effects of oxygen crossover during startup, although brushes are much larger than needed to sustain high power. © 2011 Elsevier B.V.

  1. Analysis of carbon fiber brush loading in anodes on startup and performance of microbial fuel cells

    KAUST Repository

    Hutchinson, Adam J.; Tokash, Justin C.; Logan, Bruce E.

    2011-01-01

    Flat carbon anodes placed near a cathode in a microbial fuel cell (MFC) are adversely affected by oxygen crossover, but graphite fiber brush anodes placed near the cathode produce high power densities. The impact of the brush size and electrode spacing was examined by varying the distance of the brush end from the cathode and solution conductivity in multiple MFCs. The startup time was increased from 8 ± 1 days with full brushes (all buffer concentrations) to 13 days (50 mM), 14 days (25 mM) and 21 days (8 mM) when 75% of the brush anode was removed. When MFCs were all first acclimated with a full brush, up to 65% of the brush material could be removed without appreciably altering maximum power. Electrochemical impedance spectroscopy (EIS) showed that the main source of internal resistance (IR) was diffusion resistance, which together with solution resistance reached 100 Ω. The IR using EIS compared well with that obtained using the polarization data slope method, indicating no major components of IR were missed. These results show that using full brush anodes avoids adverse effects of oxygen crossover during startup, although brushes are much larger than needed to sustain high power. © 2011 Elsevier B.V.

  2. Effect of the relationship between particle size, inter-particle distance, and metal loading of carbon supported fuel cell catalysts on their catalytic activity

    Science.gov (United States)

    Corradini, Patricia Gon; Pires, Felipe I.; Paganin, Valdecir A.; Perez, Joelma; Antolini, Ermete

    2012-09-01

    The effect of the relationship between particle size ( d), inter-particle distance ( x i ), and metal loading ( y) of carbon supported fuel cell Pt or PtRu catalysts on their catalytic activity, based on the optimum d (2.5-3 nm) and x i / d (>5) values, was evaluated. It was found that for y fuel cell electrode than that using catalysts with y ethanol oxidation on PtRu/C catalysts with same particle size and same degree of alloying but different metal loading. Tests in direct ethanol fuel cells showed that, compared to 20 wt% PtRu/C, the negative effect of the lower x i / d on the catalytic activity of 30 and 40 wt% PtRu/C catalysts was superior to the positive effect of the thinner catalyst layer.

  3. EFFECT OF TEFLON AND NAFION LOADING AT ANODE IN DIRECT FORMIC ACID FUEL CELL (DFAFC

    Directory of Open Access Journals (Sweden)

    M. S. MASDAR

    2016-08-01

    Full Text Available DFAFC has extensive hydrophilic nature and will cause problems in a limited mass transport in the anode side of electrode. Thus, the microporous layer (MPL of DFAFC needs a different in structure and morphology compared with that of PEMFC and DMFC because it will directly affect the performance. Therefore, in this study, the formulation of anode’s MPL has been investigated by varying the amount of Teflon and Nafion. Different loading of Teflon in MPL and Nafion in catalyst layer, i.e., 0 to 40% in weight, were used to fabricate the anode’s DFAFC. The characteristic of MPLs and anode (MPL with catalyst layer such as surface morphologies and resistivity, i.e., electrical impedance, have been analyzed using field emission scanning electron microscopy (FESEM and contact angle measurements as well as electrochemical impedance spectra (EIS. Meanwhile, the performance of fabricated anode was measured using cyclic voltammetry (CV technique with a half cell of DFAFC. From the result, it was obtained that the optimum content for both Teflon and Nafion on anode’s DFAFC was 20 wt% as shown in a highest electro-activity in electrode. The single cell DFAFC with optimum MEA formulation showed a good performance and hence, it is possible to apply the electricity power for electronic devices.

  4. Reactivity considerations for the on-line refuelling of a pebble bed modular reactor—Illustrating safety for the most reactive core fuel load

    International Nuclear Information System (INIS)

    Reitsma, Frederik

    2012-01-01

    In the multi-pass fuel management scheme employed for the pebble bed modular reactor the fuel pebbles are re-circulated until they reach the target burn-up. The rate at which fresh fuel is loaded and burned fuel is discharged is a result of the core neutronics cycle analysis but in practice (on the plant) this has to be controlled and managed by the fuel handling and storage system and use of the burnup measurement system. The excess reactivity is the additional reactivity available in the core during operating conditions that is the result of loading a fuel mixture in the core that is more reactive (less burned) than what is required to keep the reactor critical at full power operational conditions. The excess reactivity is balanced by the insertion of the control rods to keep the reactor critical. The excess reactivity allows flexibility in operations, for example to overcome the xenon build up when power is decreased as part of load follow. In order to limit reactivity excursions and to ensure safe shutdown the excess reactivity and thus the insertion depth of the control rods at normal operating conditions has to be managed. One way to do this is by operational procedures. The reactivity effect of long-term operation with the control rods inserted deeper than the design point is investigated and a control rod insertion limit is proposed that will not limit normal operations. The effects of other phenomena that can increase the power defect, such as higher-than-expected fuel temperatures, are also introduced. All of these cases are then evaluated by ensuring cold shutdown is still achievable and where appropriate by reactivity insertion accident analysis. These aspects are investigated on the PBMR 400 MW design.

  5. Experience with the loading and transport of fuel assembly transport casks, including CASTOR casks, and the radiation exposure of personnel

    International Nuclear Information System (INIS)

    Bentele, W.; Kinzelmann, T.

    1999-01-01

    In 1997 and 1998, six spent fuel assembly transports started from the nuclear power plant Gemeinschaftskernkraftwerk Neckar (GKN), using CASTOR-V19 casks. Professor Kuni of Marburg University challenged the statement made by the German Federal Office for Radiation Protection (Bundesamt fuer Strahlenschutz (BfS)) based on accepted scientific knowledge, according to which so-called CASTOR transports present no risk, either to the population or to the escorting police units. This paper shows that the collective dose during the loading of the CASTOR casks amounted to 4.5 mSv (gamma and neutrons) per cask at the most, and that the maximum individual dose amounted to 0.26 mSv. In addition to these doses, the collective dose during handling and transport must be considered: this amounted to 0.35 mSv (gamma and neutrons). The dose to the police escort was -2 (limit for surface contamination), presented degrees of contamination >4 Bq cm -2 upon reaching the Valognes/Cogema terminal. However, transport casks coming from French plants also revealed degrees of contamination >4 Bq cm -2 , as well as 'hot spots'. No such contamination was found on NTL 11 casks transported from the GKN to Sellafield. Neither was any increased contamination found upon the arrival of CASTOR-V19 casks transported from GKN to Gorleben or Ahaus. The partially sensationalist media reports were inversely proportional to the actual radiological relevance of the matter. The German Commission on Radiation Protection (SSK) confirmed that the radiological effect of such contaminated spent fuel transports is negligible. (author)

  6. Gaseous and Particulate Emissions from Diesel Engines at Idle and under Load: Comparison of Biodiesel Blend and Ultralow Sulfur Diesel Fuels.

    Science.gov (United States)

    Chin, Jo-Yu; Batterman, Stuart A; Northrop, William F; Bohac, Stanislav V; Assanis, Dennis N

    2012-11-15

    Diesel exhaust emissions have been reported for a number of engine operating strategies, after-treatment technologies, and fuels. However, information is limited regarding emissions of many pollutants during idling and when biodiesel fuels are used. This study investigates regulated and unregulated emissions from both light-duty passenger car (1.7 L) and medium-duty (6.4 L) diesel engines at idle and load and compares a biodiesel blend (B20) to conventional ultralow sulfur diesel (ULSD) fuel. Exhaust aftertreatment devices included a diesel oxidation catalyst (DOC) and a diesel particle filter (DPF). For the 1.7 L engine under load without a DOC, B20 reduced brake-specific emissions of particulate matter (PM), elemental carbon (EC), nonmethane hydrocarbons (NMHCs), and most volatile organic compounds (VOCs) compared to ULSD; however, formaldehyde brake-specific emissions increased. With a DOC and high load, B20 increased brake-specific emissions of NMHC, nitrogen oxides (NO x ), formaldehyde, naphthalene, and several other VOCs. For the 6.4 L engine under load, B20 reduced brake-specific emissions of PM 2.5 , EC, formaldehyde, and most VOCs; however, NO x brake-specific emissions increased. When idling, the effects of fuel type were different: B20 increased NMHC, PM 2.5 , EC, formaldehyde, benzene, and other VOC emission rates from both engines, and changes were sometimes large, e.g., PM 2.5 increased by 60% for the 6.4 L/2004 calibration engine, and benzene by 40% for the 1.7 L engine with the DOC, possibly reflecting incomplete combustion and unburned fuel. Diesel exhaust emissions depended on the fuel type and engine load (idle versus loaded). The higher emissions found when using B20 are especially important given the recent attention to exposures from idling vehicles and the health significance of PM 2.5 . The emission profiles demonstrate the effects of fuel type, engine calibration, and emission control system, and they can be used as source profiles for

  7. Gaseous and Particulate Emissions from Diesel Engines at Idle and under Load: Comparison of Biodiesel Blend and Ultralow Sulfur Diesel Fuels

    Science.gov (United States)

    Chin, Jo-Yu; Batterman, Stuart A.; Northrop, William F.; Bohac, Stanislav V.; Assanis, Dennis N.

    2015-01-01

    Diesel exhaust emissions have been reported for a number of engine operating strategies, after-treatment technologies, and fuels. However, information is limited regarding emissions of many pollutants during idling and when biodiesel fuels are used. This study investigates regulated and unregulated emissions from both light-duty passenger car (1.7 L) and medium-duty (6.4 L) diesel engines at idle and load and compares a biodiesel blend (B20) to conventional ultralow sulfur diesel (ULSD) fuel. Exhaust aftertreatment devices included a diesel oxidation catalyst (DOC) and a diesel particle filter (DPF). For the 1.7 L engine under load without a DOC, B20 reduced brake-specific emissions of particulate matter (PM), elemental carbon (EC), nonmethane hydrocarbons (NMHCs), and most volatile organic compounds (VOCs) compared to ULSD; however, formaldehyde brake-specific emissions increased. With a DOC and high load, B20 increased brake-specific emissions of NMHC, nitrogen oxides (NOx), formaldehyde, naphthalene, and several other VOCs. For the 6.4 L engine under load, B20 reduced brake-specific emissions of PM2.5, EC, formaldehyde, and most VOCs; however, NOx brake-specific emissions increased. When idling, the effects of fuel type were different: B20 increased NMHC, PM2.5, EC, formaldehyde, benzene, and other VOC emission rates from both engines, and changes were sometimes large, e.g., PM2.5 increased by 60% for the 6.4 L/2004 calibration engine, and benzene by 40% for the 1.7 L engine with the DOC, possibly reflecting incomplete combustion and unburned fuel. Diesel exhaust emissions depended on the fuel type and engine load (idle versus loaded). The higher emissions found when using B20 are especially important given the recent attention to exposures from idling vehicles and the health significance of PM2.5. The emission profiles demonstrate the effects of fuel type, engine calibration, and emission control system, and they can be used as source profiles for apportionment

  8. Comparison of thermal, radical and chemical effects of EGR gases using availability analysis in dual-fuel engines at part loads

    International Nuclear Information System (INIS)

    Hosseinzadeh, A.; Khoshbakhti Saray, R.; Seyed Mahmoudi, S.M.

    2010-01-01

    Dual-fuel engines at part load inevitably suffer from lower thermal efficiency and higher emission of carbon monoxide and unburned fuel. A quasi-two-zone combustion model has been developed for studying the second-law analysis of a dual-fuel (diesel-gas) engine operating under part-load conditions. The model is composed of two divisions: a single-zone combustion model with chemical kinetics for combustion of natural gas fuel and a subsidiary zone for combustion of pilot fuel. In the latter zone, the pilot fuel is considered as a heat source derived from two superposed Wiebe's combustion functions to account for contribution of pilot fuel in ignition of gaseous fuel and the rest of the total released energy. This quasi-two-zone combustion model is able to establish the development of combustion process with time and associated important operating parameters, such as pressure, temperature, heat release rate (HRR) and species concentration. The present work is an attempt to investigate the combustion phenomenon from second-law point of view at part load and using exhaust gas recirculation (EGR) to improve the aforementioned problems. Therefore, the availability analysis is applied to the engine from inlet valve closing (IVC) until exhaust valve opening (EVO). Various availability components are identified and calculated separately with crank position. In this paper, the various availability components are identified and calculated separately with crank position. Then the different cases of EGR (chemical, radical and thermal cases) are applied to the availability analysis in dual-fuel engines at part loads. It is found that the chemical case of EGR has negative effect and in this case the unburned chemical availability is increased and the work availability decreases in comparison with baseline engine (without EGR). While the thermal and radical cases have positive effects on the availability terms especially on the unburned chemical availability and work availability

  9. Fast breeder reactor fuel reprocessing R and D: technological development for a commercial plant

    International Nuclear Information System (INIS)

    Colas, J.; Saudray, D.; Coste, J.A.; Roux, J.P.; Jouan, A.

    1987-01-01

    The technological developments undertaken by the CEA are applied to a plant project of a 50 t/y capacity, having to reprocess in particular the SUPERPHENIX 1 reactor fuel. French experience on fast breeder reactor fuel reprocessing is presented, then the 50 t/y capacity plant project and the research and development installations. The R and D programs are described, concerning: head-end operations, solvent extractions, Pu02 conversion and storage, out-of-specification Pu02 redissolution, fission products solution vitrification, conditioning of stainless steel hulls by melting, development of remote operation equipments, study of corrosion and analytical problems

  10. Maximum power output and load matching of a phosphoric acid fuel cell-thermoelectric generator hybrid system

    Science.gov (United States)

    Chen, Xiaohang; Wang, Yuan; Cai, Ling; Zhou, Yinghui

    2015-10-01

    Based on the current models of phosphoric acid fuel cells (PAFCs) and thermoelectric generators (TGs), a new hybrid system is proposed, in which the effects of multi-irreversibilities resulting from the activation, concentration, and ohmic overpotentials in the PAFC, Joule heat and heat leak in the TG, finite-rate heat transfer between the TG and the heat reservoirs, and heat leak from the PAFC to the environment are taken into account. Expressions for the power output and efficiency of the PAFC, TG, and hybrid system are analytically derived and directly used to discuss the performance characteristics of the hybrid system. The optimal relationship between the electric currents in the PAFC and TG is obtained. The maximum power output is numerically calculated. It is found that the maximum power output density of the hybrid system will increase about 150 Wm-2, compared with that of a single PAFC. The problem how to optimally match the load resistances of two subsystems is discussed. Some significant results for practical hybrid systems are obtained.

  11. Automatic determination of BWR fuel loading patterns based on K.E. technique with core physics simulation

    International Nuclear Information System (INIS)

    Ikehara, T.; Tsuiki, M.; Takeshita, T.

    1990-01-01

    On the basis oof a computerized search method, a prototype for a fuel loading pattern expert system has been developed to support designers in core design for BWRs. The method was implemented by coupling rules and core physics simulators into an inference engine to establish an automated generate-and-test cycle. A search control mechanism, which prunes paths to be searched and selects appropriate rules through the interaction with the user, was also introduced to accomplish an effective search. The constraints in BWR core design are: (1) cycle length more than L, (2) core shutdown margin more than S, and (3) thermal margin more than T. Here L, S, and T are the specified minimum values. In this system, individual rules contain the manipulation to improve the core shutdown margin explicitly. Other items were taken into account only implicitly. Several applications to the test cases were carried out. It was found that the results were comparable with those obtained by human expert engineers. Broad applicability of the present method in the BWR core design domain was proved

  12. Burnup analysis of a peu a peu fuel-loading scheme in a pebble bed reactor using the Monte Carlo method

    International Nuclear Information System (INIS)

    Irwanto, Dwi; Obara, Toru

    2010-01-01

    The design of a pebble bed reactor can be simplified by removing the unloading device from the system. For this reactor design, a suitable fuel-loading scheme is the peu a peu (little by little) fueling scheme. In the peu a peu mode, there is no unloading device; as such, the fuels are never discharged and remain at the bottom of the core during reactor operation. This means that the burnup cycle and reactivity is controlled by the addition of fuel. In this study, the Monte Carlo method is used to perform calculations with high accuracy. However, the calculation procedures for the peu a peu mode using the Monte Carlo method require lot of steps. Therefore, a computer code to automate the process of the peu a peu fuel-loading scheme based on the Monte Carlo MVP/MVP-BURN code has been developed using Fortran. Using the method developed in this study, burnup characteristics for a reference design of a small 20-MW pebble bed reactor with the peu a peu concept were analyzed. (author)

  13. MCNP calculation of the critical H_3BO_3 concentrations for the first fuel loading into the reactor core of NPP MO-3-4 units

    International Nuclear Information System (INIS)

    Vrban, B.; Lueley, J.; Farkas, G.; Hascik, J.; Hinca, R.; Petriska, M.; Slugen, V.

    2012-01-01

    The purpose of the analysis was the determination of critical H_3BO_3 concentrations for the first fuel loading into the reactor core of MO34 units using 2"n"d generation fuel during the first start-up of new unit using calculation code MCNP 1.60. H_3BO_3 concentrations were computed for the given temperature of the primary circuit and position of the 6"t"h safety control rod group. Because of the very first start-up of these units, detailed analyses of active-core parameters are required by National Regulatory Authority and needed for safe operation of nuclear facility. (authors)

  14. Achieving clean and efficient engine operation up to full load by combining optimized RCCI and dual-fuel diesel-gasoline combustion strategies

    International Nuclear Information System (INIS)

    Benajes, Jesús; García, Antonio; Monsalve-Serrano, Javier; Boronat, Vicente

    2017-01-01

    Highlights: • Optimized dual-fuel strategy to cover the whole engine load-speed map. • EURO VI NOx levels up to 14 bar IMEP with fully and highly premixed RCCI strategies. • Dual-fuel provides up to 7% higher efficiency than CDC if urea consumption is considered. - Abstract: This experimental work investigates the capabilities of the reactivity controlled compression ignition combustion concept to be operated in the whole engine map and discusses its benefits when compared to conventional diesel combustion. The experiments were conducted using a single-cylinder medium-duty diesel engine fueled with regular gasoline and diesel fuels. The main modification on the stock engine architecture was the addition of a port fuel injector in the intake manifold. In addition, with the aim of extending the reactivity controlled compression ignition operating range towards higher loads, the piston bowl volume was increased to reduce the compression ratio of the engine from 17.5:1 (stock) down to 15.3:1. To allow the dual-fuel operation over the whole engine map without exceeding the mechanical limitations of the engine, an optimized dual-fuel combustion strategy is proposed in this research. The combustion strategy changes as the engine load increases, starting from a fully premixed reactivity controlled compression ignition combustion up to around 8 bar IMEP, then switching to a highly premixed reactivity controlled compression ignition combustion up to 15 bar IMEP, and finally moving to a mainly diffusive dual-fuel combustion to reach the full load operation. The engine mapping results obtained using this combustion strategy show that reactivity controlled compression ignition combustion allows fulfilling the EURO VI NOx limit up to 14 bar IMEP. Ultra-low soot emissions are also achieved when the fully premixed combustion is promoted, however, the soot levels rise notably as the combustion strategy moves to a less premixed pattern. Finally, the direct comparison of

  15. Stress analysis on the valve of the rotating shield, coupled with fuel element loading-unloading machine in a PWR pressure vessel

    International Nuclear Information System (INIS)

    Albuquerque, L.B. de; Jesus Miranda, C.A. de.

    1992-01-01

    A finite element static analysis was performed with the valve of the Rotating Shield (RS) which is coupled with the Fuel. Element Loading-Unloading Machine under OBE earthquake. The applied leads were obtained from a previous seismic analysis with the response spectrum method of the MTC under OBE load. A 3-D model with shell elements was developed for the valve body and for a part of the RS. The ANSYS program, version 4.4 A, was used. The two main scopes of this work were to verify the valve stresses and the functionality of its moving parts during the earthquake. (author)

  16. Effect of temperature on the expansion and microstructure Of U3 Si2-AI mini plate fuel of 3.6 g/cm3 uranium loading

    International Nuclear Information System (INIS)

    Ginting, A. Br.; Samosir, N.; Suparjo; Nasution, H.

    2000-01-01

    Expansion analysis has been conducted to 50 x 20-mm U 3 Si 2 -AI mini plate of 3.6 g/cm 3 uranium loading using dilatometer. The analysis was carried out at various temperatures of 170 o C, 350 o C and 550 o C in Argon medium with delay time 4 days. The result showed that the fuel plate was relatively stable with increasing of heating time but underwent significant expansion. Heating at 170 o C, 350 o C and 550 o C resulted in the expansion of the U 3 Si 2 -AI fuel plate of to 83-212 mum, 333-475 mum, and 433-724 mum with coefficient expansion of 24.2x10 -6 / o C - 24.3x10 -6 / o C, 25.5x10 -6 / o C - 26.2x10 -6 /'oC and 26.6 x 10 -6 / o C - 28.2 x 10 -6 / o C respectively. Microanalysis of the U 3 Si 2 -AI mini plate fuel with SEM-EDS upon heating at those temperature variation showed that microstructure change didn't occur at 170 o C, mean while interaction between AIMg2 cladding and the fuel meat appeared to take place at 350 o C and 550 o C. Data on the expansion and microstructure change of U 3 Si 2 -AI fuel plate upon heating are of great important for the manufacture/fabrication of research fuel plate to produce silicide fuel element for higher uranium loading. (author)

  17. Effect of engine load and biogas flow rate to the performance of a compression ignition engine run in dual-fuel (dieselbiogas) mode

    Science.gov (United States)

    Ambarita, H.

    2018-02-01

    The Government of Indonesia (GoI) has released a target on reduction Green Houses Gases emissions (GHG) by 26% from level business-as-usual by 2020, and the target can be up to 41% by international supports. In the energy sector, this target can be reached effectively by promoting fossil fuel replacement or blending with biofuel. One of the potential solutions is operating compression ignition (CI) engine in dual-fuel (diesel-biogas) mode. In this study effects of engine load and biogas flow rate on the performance and exhaust gas emissions of a compression ignition engine run in dual-fuel mode are investigated. In the present study, the used biogas is refined with methane content 70% of volume. The objectives are to explore the optimum operating condition of the CI engine run in dual-fuel mode. The experiments are performed on a four-strokes CI engine with rated output power of 4.41 kW. The engine is tested at constant speed 1500 rpm. The engine load varied from 600W to 1500W and biogas flow rate varied from 0 L/min to 6 L/min. The results show brake thermal efficiency of the engine run in dual-fuel mode is better than pure diesel mode if the biogas flow rates are 2 L/min and 4 L/min. It is recommended to operate the present engine in a dual-fuel mode with biogas flow rate of 4 L/min. The consumption of diesel fuel can be replaced up to 50%.

  18. EdF speaks about economic advantages of fuel reprocessing as compared with interim storage

    International Nuclear Information System (INIS)

    Anon.

    1997-01-01

    The French company Electricite de France (EdF) will prefer nuclear fuel reprocessing and plutonium recycling to spent fuel storage also in the years after 2000. This option is economically advantageous if the proportional cost of reprocessing does not exceed 1900 FRF/kg heavy metal. Economic analysis shows that this is feasible. EdF will soon have to reprocess annually about 1000 Mt spent fuel to supply enough plutonium for MOX fuel fabrication to feed as many as 28 PWR units and the Superphenix reactor. Spent fuel reprocessing is seen as promising as long as the efficiency of the MOX fuel approaches that of natural uranium based fuel. The French national industrial, political and legal context of EdF operations is also considered. (P.A.)

  19. Photo guide for estimating fuel loading and fire behavior in mixed-oak forests of the Mid-Atlantic Region

    Science.gov (United States)

    Patrick H. Brose

    2009-01-01

    A field guide of 45 pairs of photographs depicting ericaceous shrub, leaf litter, and logging slash fuel types of eastern oak forests and observed fire behavior of these fuel types during prescribed burning. The guide contains instructions on how to use the photo guide to choose appropriate fuel models for prescribed fire planning.

  20. Used Nuclear Fuel Loading and Structural Performance Under Normal Conditions of Transport - Modeling, Simulation and Experimental Integration RD&D Plan

    Energy Technology Data Exchange (ETDEWEB)

    Adkins, Harold E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2013-04-01

    Under current U.S. Nuclear Regulatory Commission regulation, it is not sufficient for used nuclear fuel (UNF) to simply maintain its integrity during the storage period, it must maintain its integrity in such a way that it can withstand the physical forces of handling and transportation associated with restaging the fuel and moving it to treatment or recycling facilities, or a geologic repository. Hence it is necessary to understand the performance characteristics of aged UNF cladding and ancillary components under loadings stemming from transport initiatives. Researchers would like to demonstrate that enough information, including experimental support and modeling and simulation capabilities, exists to establish a preliminary determination of UNF structural performance under normal conditions of transport (NCT). This research, development and demonstration (RD&D) plan describes a methodology, including development and use of analytical models, to evaluate loading and associated mechanical responses of UNF rods and key structural components. This methodology will be used to provide a preliminary assessment of the performance characteristics of UNF cladding and ancillary components under rail-related NCT loading. The methodology couples modeling and simulation and experimental efforts currently under way within the Used Fuel Disposition Campaign (UFDC). The methodology will involve limited uncertainty quantification in the form of sensitivity evaluations focused around available fuel and ancillary fuel structure properties exclusively. The work includes collecting information via literature review, soliciting input/guidance from subject matter experts, performing computational analyses, planning experimental measurement and possible execution (depending on timing), and preparing a variety of supporting documents that will feed into and provide the basis for future initiatives. The methodology demonstration will focus on structural performance evaluation of

  1. Steady- and transient-state analyses of fully ceramic microencapsulated fuel loaded reactor core via two-temperature homogenized thermal-conductivity model

    International Nuclear Information System (INIS)

    Lee, Yoonhee; Cho, Nam Zin

    2015-01-01

    Highlights: • Fully ceramic microencapsulated fuel-loaded core is analyzed via a two-temperature homogenized thermal-conductivity model. • The model is compared to harmonic- and volumetric-average thermal conductivity models. • The three thermal analysis models show ∼100 pcm differences in the k eff eigenvalue. • The three thermal analysis models show more than 70 K differences in the maximum temperature. • There occur more than 3 times differences in the maximum power for a control rod ejection accident. - Abstract: Fully ceramic microencapsulated (FCM) fuel, a type of accident-tolerant fuel (ATF), consists of TRISO particles randomly dispersed in a SiC matrix. In this study, for a thermal analysis of the FCM fuel with such a high heterogeneity, a two-temperature homogenized thermal-conductivity model was applied by the authors. This model provides separate temperatures for the fuel-kernels and the SiC matrix. It also provides more realistic temperature profiles than those of harmonic- and volumetric-average thermal conductivity models, which are used for thermal analysis of a fuel element in VHTRs having a composition similar to the FCM fuel, because such models are unable to provide the fuel-kernel and graphite matrix temperatures separately. In this study, coupled with a neutron diffusion model, a FCM fuel-loaded reactor core is analyzed via a two-temperature homogenized thermal-conductivity model at steady- and transient-states. The results are compared to those from harmonic- and volumetric-average thermal conductivity models, i.e., we compare k eff eigenvalues, power distributions, and temperature profiles in the hottest single-channel at steady-state. At transient-state, we compare total powers, reactivity, and maximum temperatures in the hottest single-channel obtained by the different thermal analysis models. The different thermal analysis models and the availability of fuel-kernel temperatures in the two-temperature homogenized thermal

  2. Enabling the Distributed Generation Market of High Temperature Fuel Cell and Absorption Chiller Systems to Support Critical and Commercial Loads

    Science.gov (United States)

    DiMola, Ashley M.

    Buildings account for over 18% of the world's anthropogenic Greenhouse Gas (GHG) emissions. As a result, a technology that can offset GHG emissions associated with buildings has the potential to save over 9 Giga-tons of GHG emissions per year. High temperature fuel cell and absorption chiller (HTFC/AC) technology offers a relatively low-carbon option for meeting cooling and electric loads for buildings while producing almost no criteria pollutants. GHG emissions in the state of California would decrease by 7.48 million metric tons per year if every commercial building in the State used HTFC/AC technology to meet its power and cooling requirements. In order to realize the benefits of HTFC/AC technology on a wide scale, the distributed generation market needs to be exposed to the technology and informed of its economic viability and real-world potential. This work characterizes the economics associated with HTFC/AC technology using select scenarios that are representative of realistic applications. The financial impacts of various input factors are evaluated and the HTFC/AC simulations are compared to the economics of traditional building utilities. It is shown that, in addition to the emissions reductions derived from the systems, HTFC/AC technology is financially preferable in all of the scenarios evaluated. This work also presents the design of a showcase environment, centered on a beta-test application, that presents (1) system operating data gathered using a custom data acquisition module, and (2) HTFC/AC technology in a clear and approachable manner in order to serve the target audience of market stakeholders.

  3. Biophysical Mechanistic Modelling Quantifies the Effects of Plant Traits on Fire Severity: Species, Not Surface Fuel Loads, Determine Flame Dimensions in Eucalypt Forests.

    Science.gov (United States)

    Zylstra, Philip; Bradstock, Ross A; Bedward, Michael; Penman, Trent D; Doherty, Michael D; Weber, Rodney O; Gill, A Malcolm; Cary, Geoffrey J

    2016-01-01

    The influence of plant traits on forest fire behaviour has evolutionary, ecological and management implications, but is poorly understood and frequently discounted. We use a process model to quantify that influence and provide validation in a diverse range of eucalypt forests burnt under varying conditions. Measured height of consumption was compared to heights predicted using a surface fuel fire behaviour model, then key aspects of our model were sequentially added to this with and without species-specific information. Our fully specified model had a mean absolute error 3.8 times smaller than the otherwise identical surface fuel model (p fire severity are the species of plants present rather than the surface fuel load, and demonstrate the accuracy and versatility of the model for quantifying this.

  4. Texas Disasters II: Utilizing NASA Earth Observations to Assist the Texas Forest Service in Mapping and Analyzing Fuel Loads and Phenology in Texas Grasslands

    Science.gov (United States)

    Brooke, Michael; Williams, Meredith; Fenn, Teresa

    2016-01-01

    The risk of severe wildfires in Texas has been related to weather phenomena such as climate change and recent urban expansion into wild land areas. During recent years, Texas wild land areas have experienced sequences of wet and dry years that have contributed to increased wildfire risk and frequency. To prevent and contain wildfires, the Texas Forest Service (TFS) is tasked with evaluating and reducing potential fire risk to better manage and distribute resources. This task is made more difficult due to the vast and varied landscape of Texas. The TFS assesses fire risk by understanding vegetative fuel types and fuel loads. To better assist the TFS, NASA Earth observations, including Landsat and Moderate Resolution Imaging Specrtoradiometer (MODIS) data, were analyzed to produce maps of vegetation type and specific vegetation phenology as it related to potential wildfire fuel loads. Fuel maps from 2010-2011 and 2014-2015 fire seasons, created by the Texas Disasters I project, were used and provided alternating, complementary map indicators of wildfire risk in Texas. The TFS will utilize the end products and capabilities to evaluate and better understand wildfire risk across Texas.

  5. A neutronic feasibility study of the AP1000 design loaded with fully ceramic micro-encapsulated fuel

    International Nuclear Information System (INIS)

    Liang, C.; Ji, W.

    2013-01-01

    A neutronic feasibility study is performed to evaluate the utilization of fully ceramic microencapsulated (FCM) fuel in the AP1000 reactor design. The widely used Monte Carlo code MCNP is employed to perform the full core analysis at the beginning of cycle (BOC). Both the original AP1000 design and the modified design with the replacement of uranium dioxide fuel pellets with FCM fuel compacts are modeled and simulated for comparison. To retain the original excess reactivity, ranges of fuel particle packing fraction and fuel enrichment in the FCM fuel design are first determined. Within the determined ranges, the reactor control mechanism employed by the original design is directly used in the modified design and the utilization feasibility is evaluated. The worth of control of each type of fuel burnable absorber (discrete/integral fuel burnable absorbers and soluble boron in primary coolant) is calculated for each design and significant differences between the two designs are observed. Those differences are interpreted by the fundamental difference of the fuel form used in each design. Due to the usage of silicon carbide as the matrix material and the fuel particles fuel form in FCM fuel design, neutron slowing down capability is increased in the new design, leading to a much higher thermal spectrum than the original design. This results in different reactivity and fission power density distributions in each design. We conclude that a direct replacement of fuel pellets by the FCM fuel in the AP1000 cannot retain the original optimum reactor core performance. Necessary modifications of the core design should be done and the original control mechanism needs to be re-designed. (authors)

  6. Fuel and Core Design Options to Overcome the Heavy Metal Loading Limit and Improve Performance and Safety of Liquid Salt Cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Petrovic, Bojan [Georgia Inst. of Technology, Atlanta, GA (United States); Maldonado, Ivan [Univ. of Tennessee, Knoxville, TN (United States)

    2016-04-14

    The research performed in this project addressed the issue of low heavy metal loading and the resulting reduced cycle length with increased refueling frequency, inherent to all FHR designs with solid, non-movable fuel based on TRISO particles. Studies performed here focused on AHTR type of reactor design with plate (“plank”) fuel. Proposal to FY12 NEUP entitled “Fuel and Core Design Options to Overcome the Heavy Metal Loading Limit and Improve Performance and Safety of Liquid Salt Cooled Reactors” was selected for award, and the 3-year project started in August 2012. A 4-month NCE was granted and the project completed on December 31, 2015. The project was performed by Georgia Tech (Prof. Bojan Petrovic, PI) and University of Tennessee (Prof. Ivan Maldonado, Co-PI), with a total funding of $758,000 over 3 years. In addition to two Co-PIs, the project directly engaged 6 graduate students (at doctoral or MS level) and 2 postdoctoral researchers. Additionally, through senior design projects and graduate advanced design projects, another 23 undergraduate and 12 graduate students were exposed to and trained in the salt reactor technology. We see this as one of the important indicators of the project’s success and effectiveness. In the process, 1 journal article was published (with 3 journal articles in preparation), together with 8 peer-reviewed full conference papers, 8 peer-reviewed extended abstracts, as well as 1 doctoral dissertation and 2 master theses. The work included both development of models and methodologies needed to adequately analyze this type of reactor, fuel, and its fuel cycle, as well as extensive analyses and optimization of the fuel and core design.

  7. Fuel and Core Design Options to Overcome the Heavy Metal Loading Limit and Improve Performance and Safety of Liquid Salt Cooled Reactors

    International Nuclear Information System (INIS)

    Petrovic, Bojan; Maldonado, Ivan

    2016-01-01

    The research performed in this project addressed the issue of low heavy metal loading and the resulting reduced cycle length with increased refueling frequency, inherent to all FHR designs with solid, non-movable fuel based on TRISO particles. Studies performed here focused on AHTR type of reactor design with plate ('plank') fuel. Proposal to FY12 NEUP entitled 'Fuel and Core Design Options to Overcome the Heavy Metal Loading Limit and Improve Performance and Safety of Liquid Salt Cooled Reactors' was selected for award, and the 3-year project started in August 2012. A 4-month NCE was granted and the project completed on December 31, 2015. The project was performed by Georgia Tech (Prof. Bojan Petrovic, PI) and University of Tennessee (Prof. Ivan Maldonado, Co-PI), with a total funding of $758,000 over 3 years. In addition to two Co-PIs, the project directly engaged 6 graduate students (at doctoral or MS level) and 2 postdoctoral researchers. Additionally, through senior design projects and graduate advanced design projects, another 23 undergraduate and 12 graduate students were exposed to and trained in the salt reactor technology. We see this as one of the important indicators of the project's success and effectiveness. In the process, 1 journal article was published (with 3 journal articles in preparation), together with 8 peer-reviewed full conference papers, 8 peer-reviewed extended abstracts, as well as 1 doctoral dissertation and 2 master theses. The work included both development of models and methodologies needed to adequately analyze this type of reactor, fuel, and its fuel cycle, as well as extensive analyses and optimization of the fuel and core design.

  8. Four years Re-Use of low burned fuel assemblies from units 1 and 2 in core loadings of units 3 and 4 WWER-440 at Kozloduy NPP

    International Nuclear Information System (INIS)

    Stoyanova, I.; Antov, A.; Spasova, V.

    2006-01-01

    At the end of 2002 units 1 and 2 of Kozloduy NPP were shutdown before their design life time which left a large number of assemblies yet with a significant energy resources. A decision to load these assemblies into the cores of Units 3 and 4 during the next 4 cycles has been taken. In 2003, 43 assemblies from Unit 1 cycle 23 rd and 55 assemblies from Unit 2 cycle 24 th are loaded in the cores of units 3 and 4 respectively. In 2004, new 49 assemblies from Unit 1 cycle 23rd and new 55 assemblies from Unit 2 cycle 24th are loaded in the cores of units 3 and 4 respectively. In 2005, the next new 25 assemblies from Unit 1 cycle 23 rd and 66 assemblies from Unit 2 cycle 24th are loaded in the cores of units 3 and 4 respectively. In 2006, the next new 54 assemblies from Unit 1 cycle 23 rd and 52 assemblies from Unit 2 cycle 24 th + 2 assemblies from Unit 3 cycle 19th are loaded in the cores of Units 3 and 4 respectively. The SPPSHB computer code system is used for development and safety assessment of the fuel loading patterns of Units 3 and 4 at Kozloduy NPP with low burned assemblies from units 1 and 2 (Authors)

  9. Power generation in microbial fuel cells using platinum group metal-free cathode catalyst: Effect of the catalyst loading on performance and costs.

    Science.gov (United States)

    Santoro, Carlo; Kodali, Mounika; Herrera, Sergio; Serov, Alexey; Ieropoulos, Ioannis; Atanassov, Plamen

    2018-02-28

    Platinum group metal-free (PGM-free) catalyst with different loadings was investigated in air breathing electrodes microbial fuel cells (MFCs). Firstly, the electrocatalytic activity towards oxygen reduction reaction (ORR) of the catalyst was investigated by rotating ring disk electrode (RRDE) setup with different catalyst loadings. The results showed that higher loading led to an increased in the half wave potential and the limiting current and to a further decrease in the peroxide production. The electrons transferred also slightly increased with the catalyst loading up to the value of ≈3.75. This variation probably indicates that the catalyst investigated follow a 2x2e - transfer mechanism. The catalyst was integrated within activated carbon pellet-like air-breathing cathode in eight different loadings varying between 0.1 mgcm -2 and 10 mgcm -2 . Performance were enhanced gradually with the increase in catalyst content. Power densities varied between 90 ± 9 μWcm -2 and 262 ± 4 μWcm -2 with catalyst loading of 0.1 mgcm -2 and 10 mgcm -2 respectively. Cost assessments related to the catalyst performance are presented. An increase in catalyst utilization led to an increase in power generated with a substantial increase in the whole costs. Also a decrease in performance due to cathode/catalyst deterioration over time led to a further increase in the costs.

  10. A study on the optimal fuel loading pattern design in pressurized water reactors using the artificial neural network and the fuzzy rule based system

    International Nuclear Information System (INIS)

    Kim, Han Gon

    1993-02-01

    In pressurized water reactors, the fuel reloading problem has significant meaning in terms of both safety and economic aspects. Therefore the general problem of incore fuel management for a PWR consists of determining the fuel reloading policy for each cycle that minimize unit energy cost under the constraints imposed on various core parameters, e.g., a local power peaking factor and an assembly burnup. This is equivalent that a cycle length is maximized for a given energy cost under the various constraints. Existing optimization methods do not ensure the global optimum solution because of the essential limitation of their searching algorithms. They only find near optimal solutions. To solve this limitation, a hybrid artificial neural network system is developed for the optimal fuel loading pattern design using a fuzzy rule based system and an artificial neural networks. This system finds the patterns that P max is lower than the predetermined value and K eff is larger than the reference value. The back-propagation networks are developed to predict PWR core parameters. Reference PWR is an 121-assembly typical PWR. The local power peaking factor and the effective multiplication factor at BOC condition are predicted. To obtain target values of these two parameters, the QCC code are used. Using this code, 1000 training patterns are obtained, randomly. Two networks are constructed, one for P max and another for K eff Both of two networks have 21 input layer neurons, 18 output layer neurons, and 120 and 393 hidden layer neurons, respectively. A new learning algorithm is proposed. This is called the advanced adaptive learning algorithm. The weight change step size of this algorithm is optimally varied inversely proportional to the average difference between an actual output value and an ideal target value. This algorithm greatly enhances the convergence speed of a BPN. In case of P max prediction, 98% of the untrained patterns are predicted within 6% error, and in case

  11. Tests of experimental fuel elements by the method of nuclear-thermal pulse loadings in 'HYDRA' reactor

    International Nuclear Information System (INIS)

    Nastoyashchaya, O.V.; Lebedev, Yu. M.; Chechurov, A.M.; Khvostionov, Ye

    1997-01-01

    The results of tests of experimental fuel elements with uranium dioxide fuel composition embedded in Al and Zr matrix with the enrichment from 90% to 36% in respect to U-235 performed at the pulse 'HYDRA' reactor are presented in this paper. Testing is performed in the frame-work of extensive research program studying the behavior of fuel elements (FE) of research and mini nuclear power systems in case of practically immediate energy release in the fuel taking place during the RIA-type accidents. Duration of the neutron pulse when testing in 'HYDRA' reactor is from 7 to 20 ms. The methods of diagnostics of the state of FE prior to and after testing in the reactor are developed and verified. Mathematical model describing temperature fields inside the FE in the process of testing. and accounting for non-uniformity of fuel composition has been developed in order to summarize experimental results. Experimental data on the limiting values of the energy density leading to deformation and degradation of FE depending on the type of fuel composition have been obtained and the mechanisms for the development of these processes have been determined. The nature of physical-chemical processes taking place in the fuel composition and fuel cladding depending on material composition under different levels of energy deposition is demonstrated. The data on hydrogen generation and radioactive product release out of fuel after failure of FE are presented. (author)

  12. Effects of fuels, engine load and exhaust after-treatment on diesel engine SVOC emissions and development of SVOC profiles for receptor modeling

    Science.gov (United States)

    Huang, Lei; Bohac, Stanislav V.; Chernyak, Sergei M.; Batterman, Stuart A.

    2015-01-01

    Diesel exhaust emissions contain numerous semivolatile organic compounds (SVOCs) for which emission information is limited, especially for idling conditions, new fuels and the new after-treatment systems. This study investigates exhaust emissions of particulate matter (PM), polycyclic aromatic hydrocarbons (PAHs), nitro-PAHs (NPAHs), and sterane and hopane petroleum biomarkers from a heavy-duty (6.4 L) diesel engine at various loads (idle, 600 and 900 kPa BMEP), with three types of fuel (ultra-low sulfur diesel or ULSD, Swedish low aromatic diesel, and neat soybean biodiesel), and with and without a diesel oxidation catalyst (DOC) and diesel particulate filter (DPF). Swedish diesel and biodiesel reduced emissions of PM2.5, Σ15PAHs, Σ11NPAHs, Σ5Hopanes and Σ6Steranes, and biodiesel resulted in the larger reductions. However, idling emissions increased for benzo[k]fluoranthene (Swedish diesel), 5-nitroacenaphthene (biodiesel) and PM2.5 (biodiesel), a significant result given the attention to exposures from idling vehicles and the toxicity of high-molecular-weight PAHs and NPAHs. The DOC + DPF combination reduced PM2.5 and SVOC emissions during DPF loading (>99% reduction) and DPF regeneration (83–99%). The toxicity of diesel exhaust, in terms of the estimated carcinogenic risk, was greatly reduced using Swedish diesel, biodiesel fuels and the DOC + DPF. PAH profiles showed high abundances of three and four ring compounds as well as naphthalene; NPAH profiles were dominated by nitro-naphthalenes, 1-nitropyrene and 9-nitroanthracene. Both the emission rate and the composition of diesel exhaust depended strongly on fuel type, engine load and after-treatment system. The emissions data and chemical profiles presented are relevant to the development of emission inventories and exposure and risk assessments. PMID:25709535

  13. Effects of fuels, engine load and exhaust after-treatment on diesel engine SVOC emissions and development of SVOC profiles for receptor modeling.

    Science.gov (United States)

    Huang, Lei; Bohac, Stanislav V; Chernyak, Sergei M; Batterman, Stuart A

    2015-02-01

    Diesel exhaust emissions contain numerous semivolatile organic compounds (SVOCs) for which emission information is limited, especially for idling conditions, new fuels and the new after-treatment systems. This study investigates exhaust emissions of particulate matter (PM), polycyclic aromatic hydrocarbons (PAHs), nitro-PAHs (NPAHs), and sterane and hopane petroleum biomarkers from a heavy-duty (6.4 L) diesel engine at various loads (idle, 600 and 900 kPa BMEP), with three types of fuel (ultra-low sulfur diesel or ULSD, Swedish low aromatic diesel, and neat soybean biodiesel), and with and without a diesel oxidation catalyst (DOC) and diesel particulate filter (DPF). Swedish diesel and biodiesel reduced emissions of PM 2.5 , Σ 15 PAHs, Σ 11 NPAHs, Σ 5 Hopanes and Σ 6 Steranes, and biodiesel resulted in the larger reductions. However, idling emissions increased for benzo[k]fluoranthene (Swedish diesel), 5-nitroacenaphthene (biodiesel) and PM 2.5 (biodiesel), a significant result given the attention to exposures from idling vehicles and the toxicity of high-molecular-weight PAHs and NPAHs. The DOC + DPF combination reduced PM 2.5 and SVOC emissions during DPF loading (>99% reduction) and DPF regeneration (83-99%). The toxicity of diesel exhaust, in terms of the estimated carcinogenic risk, was greatly reduced using Swedish diesel, biodiesel fuels and the DOC + DPF. PAH profiles showed high abundances of three and four ring compounds as well as naphthalene; NPAH profiles were dominated by nitro-naphthalenes, 1-nitropyrene and 9-nitroanthracene. Both the emission rate and the composition of diesel exhaust depended strongly on fuel type, engine load and after-treatment system. The emissions data and chemical profiles presented are relevant to the development of emission inventories and exposure and risk assessments.

  14. Optimization of enrichment distributions in nuclear fuel assemblies loaded with Uranium and Plutonium via a modified linear programming technique

    Energy Technology Data Exchange (ETDEWEB)

    Cuevas Vivas, Gabriel Francisco

    1999-12-01

    A methodology to optimize enrichment distributions in Light Water Reactor (LWR) fuel assemblies is developed and tested. The optimization technique employed is the linear programming revised simplex method, and the fuel assembly's performance is evaluated with a neutron transport code that is also utilized in the calculation of sensitivity coefficients. The enrichment distribution optimization procedure begins from a single-value (flat) enrichment distribution until a target, maximum local power peaking factor, is achieved. The optimum rod enrichment distribution, with 1.00 for the maximum local power peaking factor and with each rod having its own enrichment, is calculated at an intermediate stage of the analysis. Later, the best locations and values for a reduced number of rod enrichments is obtained as a function of a target maximum local power peaking factor by applying sensitivity to change techniques. Finally, a shuffling process that assigns individual rod enrichments among the enrichment groups is performed. The relative rod power distribution is then slightly modified and the rod grouping redefined until the optimum configuration is attained. To verify the accuracy of the relative rod power distribution, a full computation with the neutron transport code using the optimum enrichment distribution is carried out. The results are compared and tested for assembly designs loaded with fresh Low Enriched Uranium (LEU) and plutonium Mixed Oxide (MOX) isotopics for both reactor-grade and weapons-grade plutonium were utilized to demonstrate the wide range of applicability of the optimization technique. The feature of the assembly designs used for evaluation purposes included burnable absorbers and internal water regions, and were prepared to resemble the configurations of modern assemblies utilized in commercial Boiling Water Reactor (BWRs) and Pressurized Water Reactors (PWRs). In some cases, a net improvement in the relative rod power distribution or in the

  15. ALARA Principle Application for Loading Spent Nuclear Fuel Assemblies from Nuclear Research Reactor WR-S Mergal-Bucharest Romania into Transportation Casks

    International Nuclear Information System (INIS)

    Dragusin, M.

    2009-01-01

    Safety implementation of Spent Nuclear Fuels Assemblies (SNFA) handling procedures at the WR-S reactor site is ensured by technical perfection and reliability of equipment, monitoring of its condition, qualification and discipline of personnel as well as organization and execution of work complied with requirements of regulatory documents, process procedures, guidance and manuals. The personnel training for execution loading of SNF FAs is other important aspect for radiation protection and safely activities. Estimations carried out using Micro Shield software show that maximal dose rate upon working site when loading four FAs into basket of cask will not exceed 1.7 and 956;Sv/h, excluding natural radiation. Radiation Safety Analyses estimates for loading 70 SNFA in 18 transportation casks are: maximal individual dose: 4274.7 and 956;Sv, maximal expected collective dose persons: 17 031.2 man and 956;Sv. By application ALARA principle with technical and administrative measures the loading process developed in the following conditions: maximal individual dose: 68 and 956;Sv, the collective dose persons: 732 man and 956;Sv. The work will presented the technical measures and procedures applied in loading process.

  16. Effect of the relationship between particle size, inter-particle distance, and metal loading of carbon supported fuel cell catalysts on their catalytic activity

    Energy Technology Data Exchange (ETDEWEB)

    Gon Corradini, Patricia; Pires, Felipe I.; Paganin, Valdecir A.; Perez, Joelma, E-mail: jperez@iqsc.usp.br [Instituto de Quimica de Sao Carlos, USP (Brazil); Antolini, Ermete [Scuola di Scienza dei Materiali (Italy)

    2012-09-15

    The effect of the relationship between particle size (d), inter-particle distance (x{sub i}), and metal loading (y) of carbon supported fuel cell Pt or PtRu catalysts on their catalytic activity, based on the optimum d (2.5-3 nm) and x{sub i}/d (>5) values, was evaluated. It was found that for y < 30 wt%, the optimum values of both d and x{sub i}/d can be always obtained. For y {>=} 30 wt%, instead, the positive effect of a thinner catalyst layer of the fuel cell electrode than that using catalysts with y < 30 wt% is concomitant to a decrease of the effective catalyst surface area due to an increase of d and/or a decrease of x{sub i}/d compared to their optimum values, with in turns gives rise to a decrease in the catalytic activity. The effect of the x{sub i}/d ratio has been successfully verified by experimental results on ethanol oxidation on PtRu/C catalysts with same particle size and same degree of alloying but different metal loading. Tests in direct ethanol fuel cells showed that, compared to 20 wt% PtRu/C, the negative effect of the lower x{sub i}/d on the catalytic activity of 30 and 40 wt% PtRu/C catalysts was superior to the positive effect of the thinner catalyst layer.

  17. Dynamic Thermal Loads and Cooling Requirements Calculations for V ACs System in Nuclear Fuel Processing Facilities Using Computer Aided Energy Conservation Models

    International Nuclear Information System (INIS)

    EL Fawal, M.M.; Gadalla, A.A.; Taher, B.M.

    2010-01-01

    In terms of nuclear safety, the most important function of ventilation air conditioning (VAC) systems is to maintain safe ambient conditions for components and structures important to safety inside the nuclear facility and to maintain appropriate working conditions for the plant's operating and maintenance staff. As a part of a study aimed to evaluate the performance of VAC system of the nuclear fuel cycle facility (NFCF) a computer model was developed and verified to evaluate the thermal loads and cooling requirements for different zones of fuel processing facility. The program is based on transfer function method (TFM) and it is used to calculate the dynamic heat gain by various multilayer walls constructions and windows hour by hour at any orientation of the building. The developed model was verified by comparing the obtained calculated results of the solar heat gain by a given building with the corresponding calculated values using finite difference method (FDM) and total equivalent temperature different method (TETD). As an example the developed program is used to calculate the cooling loads of the different zones of a typical nuclear fuel facility the results showed that the cooling capacities of the different cooling units of each zone of the facility meet the design requirements according to safety regulations in nuclear facilities.

  18. Effect of the relationship between particle size, inter-particle distance, and metal loading of carbon supported fuel cell catalysts on their catalytic activity

    International Nuclear Information System (INIS)

    Gon Corradini, Patricia; Pires, Felipe I.; Paganin, Valdecir A.; Perez, Joelma; Antolini, Ermete

    2012-01-01

    The effect of the relationship between particle size (d), inter-particle distance (x i ), and metal loading (y) of carbon supported fuel cell Pt or PtRu catalysts on their catalytic activity, based on the optimum d (2.5–3 nm) and x i /d (>5) values, was evaluated. It was found that for y i /d can be always obtained. For y ≥ 30 wt%, instead, the positive effect of a thinner catalyst layer of the fuel cell electrode than that using catalysts with y i /d compared to their optimum values, with in turns gives rise to a decrease in the catalytic activity. The effect of the x i /d ratio has been successfully verified by experimental results on ethanol oxidation on PtRu/C catalysts with same particle size and same degree of alloying but different metal loading. Tests in direct ethanol fuel cells showed that, compared to 20 wt% PtRu/C, the negative effect of the lower x i /d on the catalytic activity of 30 and 40 wt% PtRu/C catalysts was superior to the positive effect of the thinner catalyst layer.

  19. The determination of uranium distribution homogeneity in the fuel plates with the uranium loading of 4.80 and 5.20 g/cm3 by X-Ray attenuation

    International Nuclear Information System (INIS)

    Supardjo; Rojak, A.; Boybul; Suyoto; Datam, A. S.

    2000-01-01

    The calibration of X-Ray intensity of the U 3 Si 2 -AI fuel plates with the uranium loading between 3.60 up to 5.20 g/cm 3 and varied thickness of AIMgSi1 reference block have been performed. The measurement with changing variable slit diameter and energy of X-Ray attenuation, are produced enough representative X-Ray intensity at 18 mm slit diameter and energy of 43 kV. From the correlation of X-ray intensities vs variation of uranium loading in the fuel plates and thickness of the AIMgSi1 materials, the equivalence of thickness of the AIMgSi1 block to the uranium loading of fuel plates are determined. By assuming that the tolerance of the homogeneity measurement is + 20 % from normal thickness staircase of the AIMgSi1 standard could be determined and than together with fuel plate were scanned to determine the uranium homogeneity. The test result on the U 3 Si 2 -AI fuel plates with uranium loading of 4.80 and 5.20 g/cm 3 (each 4 fuel plates) indicated that uranium distribution in the fuel plates is relatively homogeneous, with each maximum deviation being 6.30 % and 6.90%. It is showed that measurement method is relatively good, easy, and fast so that this method is suitable to control the uranium homogeneity in the fuel plate. (author)

  20. A highly order-structured membrane electrode assembly with vertically aligned carbon nanotubes for ultra-low Pt loading PEM fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Tian, Zhi Qun; Lim, San Hua; Poh, Chee Kok; Lin, Jianyi [Institute of Chemical and Engineering Sciences, 1 Pesek Road, Jurong Island, Singapore 627833 (Singapore); Tang, Zhe; Chua, Daniel [Department of Materials Science and Engineering, National University of Singapore, Singapore 117542 (Singapore); Xia, Zetao [Institute of Materials Research and Engineering, 3 Research Link, Singapore 117602 (Singapore); Luo, Zhiqiang; Shen, Zexiang [Division of Physics and Applied Physics, School of Physical and Mathematical Sciences, Nanyang Technological University, 637371 Singapore (Singapore); Shen, Pei Kang [State Key Laboratory of Optoelectronic Materials and Technologies, and Key Laboratory of Low-carbon Chemistry and Energy Conservation of Guangdong Province, School of Physics and Engineering, Sun Yat-sen University, Guangzhou, 510275 (China); Feng, Yuan Ping [Department of Physics, National University of Singapore, Singapore 117542 (Singapore)

    2011-11-15

    A simple method was developed to prepare ultra-low Pt loading membrane electrode assembly (MEA) using vertically aligned carbon nanotubes (VACNTs) as highly ordered catalyst support for PEM fuel cells application. In the method, VACNTs were directly grown on the cheap household aluminum foil by plasma enhanced chemical vapor deposition (PECVD), using Fe/Co bimetallic catalyst. By depositing a Pt thin layer on VACNTs/Al and subsequent hot pressing, Pt/VACNTs can be 100% transferred from Al foil onto polymer electrolyte membrane for the fabrication of MEA. The whole transfer process does not need any chemical removal and destroy membrane. The PEM fuel cell with the MEA fabricated using this method showed an excellent performance with ultra-low Pt loading down to 35 {mu}g cm{sup -2} which was comparable to that of the commercial Pt catalyst on carbon powder with 400 {mu}g cm{sup -2}. To the best of our knowledge, for the first time, we identified that it is possible to substantially reduce the Pt loading one order by application of order-structured electrode based on VACNTs as Pt catalysts support, compared with the traditional random electrode at a comparable performance through experimental and mathematical methods. (Copyright copyright 2011 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  1. A study on the optimal fuel loading pattern design in pressurized water reactor using the artificial neural network and the fuzzy rule based system

    International Nuclear Information System (INIS)

    Kim, Han Gon; Chang, Soon Heung; Lee, Byung

    2004-01-01

    The Optimal Fuel Shuffling System (OFSS) is developed for optimal design of PWR fuel loading pattern. In this paper, an optimal loading pattern is defined that the local power peaking factor is lower than predetermined value during one cycle and the effective multiplication factor is maximized in order to extract maximum energy. OFSS is a hybrid system that a rule based system, a fuzzy logic, and an artificial neural network are connected each other. The rule based system classifies loading patterns into two classes using several heuristic rules and a fuzzy rule. A fuzzy rule is introduced to achieve more effective and fast searching. Its membership function is automatically updated in accordance with the prediction results. The artificial neural network predicts core parameters for the patterns generated from the rule based system. The back-propagation network is used for fast prediction of core parameters. The artificial neural network and the fuzzy logic can be used as the tool for improvement of existing algorithm's capabilities. OFSS was demonstrated and validated for cycle 1 of Kori unit 1 PWR. (author)

  2. A study on the optimal fuel loading pattern design in pressurized water reactor using the artificial neural network and the fuzzy rule based system

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Han Gon; Chang, Soon Heung; Lee, Byung [Department of Nuclear Engineering, Korea Advanced Institute of Science and Technology, Yusong-gu, Taejon (Korea, Republic of)

    2004-07-01

    The Optimal Fuel Shuffling System (OFSS) is developed for optimal design of PWR fuel loading pattern. In this paper, an optimal loading pattern is defined that the local power peaking factor is lower than predetermined value during one cycle and the effective multiplication factor is maximized in order to extract maximum energy. OFSS is a hybrid system that a rule based system, a fuzzy logic, and an artificial neural network are connected each other. The rule based system classifies loading patterns into two classes using several heuristic rules and a fuzzy rule. A fuzzy rule is introduced to achieve more effective and fast searching. Its membership function is automatically updated in accordance with the prediction results. The artificial neural network predicts core parameters for the patterns generated from the rule based system. The back-propagation network is used for fast prediction of core parameters. The artificial neural network and the fuzzy logic can be used as the tool for improvement of existing algorithm's capabilities. OFSS was demonstrated and validated for cycle 1 of Kori unit 1 PWR. (author)

  3. Stall/surge dynamics of a multi-stage air compressor in response to a load transient of a hybrid solid oxide fuel cell-gas turbine system

    Science.gov (United States)

    Azizi, Mohammad Ali; Brouwer, Jacob

    2017-10-01

    A better understanding of turbulent unsteady flows in gas turbine systems is necessary to design and control compressors for hybrid fuel cell-gas turbine systems. Compressor stall/surge analysis for a 4 MW hybrid solid oxide fuel cell-gas turbine system for locomotive applications is performed based upon a 1.7 MW multi-stage air compressor. Control strategies are applied to prevent operation of the hybrid SOFC-GT beyond the stall/surge lines of the compressor. Computational fluid dynamics tools are used to simulate the flow distribution and instabilities near the stall/surge line. The results show that a 1.7 MW system compressor like that of a Kawasaki gas turbine is an appropriate choice among the industrial compressors to be used in a 4 MW locomotive SOFC-GT with topping cycle design. The multi-stage radial design of the compressor enhances the ability of the compressor to maintain air flow rate during transient step-load changes. These transient step-load changes are exhibited in many potential applications for SOFC/GT systems. The compressor provides sustained air flow rate during the mild stall/surge event that occurs due to the transient step-load change that is applied, indicating that this type of compressor is well-suited for this hybrid application.

  4. Used Nuclear Fuel Loading and Structural Performance Under Normal Conditions of Transport- Demonstration of Approach and Results on Used Fuel Performance Characterization

    Energy Technology Data Exchange (ETDEWEB)

    Adkins, Harold [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Geelhood, Ken [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Koeppel, Brian [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Coleman, Justin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bignell, John [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Flores, Gregg [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Wang, Jy-An [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Sanborn, Scott [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Spears, Robert [Idaho National Lab. (INL), Idaho Falls, ID (United States); Klymyshyn, Nick [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2013-09-30

    This document addresses Oak Ridge National Laboratory milestone M2FT-13OR0822015 Demonstration of Approach and Results on Used Nuclear Fuel Performance Characterization. This report provides results of the initial demonstration of the modeling capability developed to perform preliminary deterministic evaluations of moderate-to-high burnup used nuclear fuel (UNF) mechanical performance under normal conditions of storage (NCS) and normal conditions of transport (NCT) conditions. This report also provides results from the sensitivity studies that have been performed. Finally, discussion on the long-term goals and objectives of this initiative are provided.

  5. Preliminary decay heat calculations for the fuel loaded irradiation loop device of the RMB multipurpose Brazilian reactor

    Energy Technology Data Exchange (ETDEWEB)

    Campolina, Daniel; Costa, Antonio Carlos L. da; Andrade, Edison P., E-mail: campolina@cdtn.br, E-mail: aclp@cdtn.br, E-mail: epa@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (SETRE/CDTN/CNEN-MG), Belo Horizonte, MG (Brazil). Servico de Tecnologia de Reatores

    2017-07-01

    The structuring project of the Brazilian Multipurpose Reactor (RMB) is responsible for meeting the capacity to develop and test materials and nuclear fuel for the Brazilian Nuclear Program. An irradiation test device (Loop) capable of performing fuel test for power reactor rods is being conceived for RMB reflector. In this work preliminary neutronic calculations have been carried out in order to determine parameters to the cooling system of the Loop basic design. The heat released as a result of radioactive decay of fuel samples was calculated using ORIGEN-ARP and it resulted less than 200 W after 1 hour of irradiation interruption. (author)

  6. Time/motion observations and dose analysis of reactor loading, transportation, and dry unloading of an overweight truck spent fuel shipment

    International Nuclear Information System (INIS)

    Hostick, C.J.; Lavender, J.C.; Wakeman, B.H.

    1992-04-01

    This document presents observed activity durations and radiation dose analyses for an overweight truck shipment of pressurized water reactor (PWR) spent fuel from the Surry Power Station in Virginia to the Idaho National Engineering Laboratory. The shipment consisted of a TN-8L shipping cask carrying three 9-year-old PWR spent fuel assemblies. Handling times and dose analyses for at-reactor activities were completed by Virginia Electric and Power Company (Virginia Power) personnel. Observations of in-transit and unloading activities were made by Pacific Northwest Laboratory (PNL) personnel, who followed the shipment for approximately 2800 miles and observed cask unloading activities. In-transit dose estimates were calculated using dose rate maps provided by Virginia Power for a fully loaded TN-8L shipping cask. The dose analysis for the cask unloading operations is based on the observations of PNL personnel

  7. Analysis and effect of multi-fuel and practical constraints on economic load dispatch in the presence of Unified Power Flow Controller using UDTPSO

    Directory of Open Access Journals (Sweden)

    Chintalapudi V. Suresh

    2015-09-01

    Full Text Available This paper presents an attempt to analyze the effect of multi-fuel and practical constraints on economic load dispatch problem using a novel uniform distributed two-stage particle swarm optimization (UDTPSO algorithm without and with unified power flow controller (UPFC while satisfying equality, inequality, practical constraints such as ramp-rate and prohibited operating zone (POZ limits and device operating limits. A Novel severity function is formulated based on the transmission line overloads and bus voltage violations to identify an optimal location to install UPFC. A multi-objective optimization problem is solved for multi-fuel non-convex cost and transmission power loss objectives. Obtained results for considered standard test functions and electrical systems indicate the effectiveness of the proposed algorithm and can obtain efficient solution when compared to existing methods. Hence, the proposed method is a promising method and can be easily applied to optimize the power system objectives.

  8. Utilization of Plutonium and Higher Actinides in the HTGR as Possibility to Maintain Long-Term Operation on One Fuel Loading

    International Nuclear Information System (INIS)

    Tsvetkova, Galina V.; Peddicord, Kenneth L.

    2002-01-01

    Promising existing nuclear reactor concepts together with new ideas are being discussed worldwide. Many new studies are underway in order to identify prototypes that will be analyzed and developed further as systems for Generation IV. The focus is on designs demonstrating full inherent safety, competitive economics and proliferation resistance. The work discussed here is centered on a modularized small-size High Temperature Gas-cooled Reactor (HTGR) concept. This paper discusses the possibility of maintaining long-term operation on one fuel loading through utilization of plutonium and higher actinides in the small-size pebble-bed reactor (PBR). Acknowledging the well-known flexibility of the PBR design with respect to fuel composition, the principal limitations of the long-term burning of plutonium and higher actinides are considered. The technological challenges and further research are outlined. The results allow the identification of physical features of the PBR that significantly influence flexibility of the design and its applications. (authors)

  9. Sizewell B nuclear power station: the basis for the decision by the Health and Safety Executive to grant consent to load fuel into the reactor

    International Nuclear Information System (INIS)

    1994-01-01

    The licensing and consent process and the basis for granting a consent for Nuclear Electric to load fuel into the Sizewell B reactor in the United Kingdom are explained. Consent was granted by the UK Nuclear Installations Inspectorate on behalf of the Health and Safety Executive on satisfactory completion of construction and those commissioning stages needed to proceed safely, and the production of a satisfactory safety case. A summary of the assessment of the safety case is appended. It covers the reactor core, coolant system structural integrity, engineered safety features, main and essential electrical system, control and instrumentation, radioactive waste management, radiological protection, fuel storage and handling, civil works and structures, fault analysis, human factors, hazard analysis, quality assurance, and decommissioning. (UK)

  10. Fuel management inside the reactor. Report of generation of the nuclear bank for the fuel of the initial load of the Laguna Verde U-1 reactor with the FMS codes; Administracion de combustible dentro del reactor. Reporte de generacion del banco nuclear para el combustible de la carga inicial del reactor de Laguna Verde U-1 con los codigos del FMS

    Energy Technology Data Exchange (ETDEWEB)

    Alonso V, G. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico); Torres A, C. [CFE, Veracruz (Mexico)

    1991-06-15

    In this work in a general way the form in that it was generated the database of the initial fuel load of the Laguna Verde Unit 1 reactor is described. The initial load is formed with fuel of the GE6 type. The obtained results during the formation of the database in as much as to the behavior of the different cell parameters regarding the one burnt of the fuel and the variation of vacuums in the coolant channel its are compared very favorably with those reported by the General Electric fuel supplier and reported in the design documents of the same one. (Author)

  11. Effects of high density dispersion fuel loading on the kinetic parameters of a low enriched uranium fueled material test research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Muhammad, Farhan [Department of Nuclear Engineering, Pakistan Institute of Engineering and Applied Sciences, P.O. Nilore, Islamabad 45650 (Pakistan)], E-mail: mfarhan_73@yahoo.co.uk; Majid, Asad [Department of Nuclear Engineering, Pakistan Institute of Engineering and Applied Sciences, P.O. Nilore, Islamabad 45650 (Pakistan)

    2008-09-15

    The effects of using high density low enriched uranium on the neutronic parameters of a material test research reactor were studied. For this purpose, the low density LEU fuel of an MTR was replaced with high density LEU fuels currently being developed under the RERTR program. Since the alloying elements have different cross-sections affecting the reactor in different ways, therefore fuels U-Mo (9 w/o) which contain the same elements in same ratio were selected for analysis. Simulations were carried out to calculate core excess reactivity, neutron flux spectrum, prompt neutron generation time, effective delayed neutron fraction and feedback coefficients including Doppler feedback coefficient, and reactivity coefficients for change of water density and temperature. Nuclear reactor analysis codes including WIMS-D4 and CITATION were employed to carry out these calculations. It is observed that the excess reactivity at the beginning of life does not increase as the uranium density of fuel. Both the prompt neutron generation time and the effective delayed neutron fraction decrease as the uranium density increases. The absolute value of Doppler feedback coefficient increases while the absolute values of reactivity coefficients for change of water density and temperature decrease.

  12. Effects of high density dispersion fuel loading on the kinetic parameters of a low enriched uranium fueled material test research reactor

    International Nuclear Information System (INIS)

    Muhammad, Farhan; Majid, Asad

    2008-01-01

    The effects of using high density low enriched uranium on the neutronic parameters of a material test research reactor were studied. For this purpose, the low density LEU fuel of an MTR was replaced with high density LEU fuels currently being developed under the RERTR program. Since the alloying elements have different cross-sections affecting the reactor in different ways, therefore fuels U-Mo (9 w/o) which contain the same elements in same ratio were selected for analysis. Simulations were carried out to calculate core excess reactivity, neutron flux spectrum, prompt neutron generation time, effective delayed neutron fraction and feedback coefficients including Doppler feedback coefficient, and reactivity coefficients for change of water density and temperature. Nuclear reactor analysis codes including WIMS-D4 and CITATION were employed to carry out these calculations. It is observed that the excess reactivity at the beginning of life does not increase as the uranium density of fuel. Both the prompt neutron generation time and the effective delayed neutron fraction decrease as the uranium density increases. The absolute value of Doppler feedback coefficient increases while the absolute values of reactivity coefficients for change of water density and temperature decrease

  13. Determining the platinum loading and distribution of industrial scale polymer electrolyte membrane fuel cell electrodes using low energy X-ray imaging

    DEFF Research Database (Denmark)

    Holst, T.; Vassiliev, Anton; Kerr, R.

    2014-01-01

    Low energy X-ray imaging (E <25 keV) is herein demonstrated to be a rapid, effective and non-destructive tool for the quantitative determination of the platinum loading and distribution over the entire geometric area of gas diffusion electrodes for polymer electrolyte membrane fuel cells. A linea...... of electrodes fabricated using an industrial spraying process. This technique proves to be an attractive option for the electrode performance study, the process optimization and quality control of electrode fabrication on an industrial scale....

  14. Using super-capacitors in combination with Bi-directional DC/DC converters for active load management in residential fuel cell applications

    Energy Technology Data Exchange (ETDEWEB)

    Cacciato, M.; Giulii Capponi, F. [Rome Univ., ' La Sapienza' , Dept. of Electrical Engineering (Italy)

    2004-07-01

    Among innovative conversion systems for alternative energy, Fuel Cells (FCs) are ideal in applications as distributed power generation or automotive. The connection of FCs to domestic or industrial loads requires a DC/AC converter also acting as a energy buffer to match the different dynamics of FCs and loads. In the last years, a new type of electrolytic capacitors called Super- Capacitors (SCs), has been designed using double layers technology. Such components are able to store more energy than electrolytic capacitors maintaining the capability to swap it at high power levels. Firstly, different solution used to connect SCs to a FC based conversion system are considered. Then, a comparison of bi-directional DC/DC converters designed to manage SCs energy is performed. Finally, the converter design and a laboratory prototype of the adopted solution are reported. (authors)

  15. Overview of the fast reactors fuels program

    International Nuclear Information System (INIS)

    Evans, E.A.; Cox, C.M.; Hayward, B.R.; Rice, L.H.; Yoshikawa, H.H.

    1980-04-01

    Each nation involved in LMFBR development has its unique energy strategies which consider energy growth projections, uranium resources, capital costs, and plant operational requirements. Common to all of these strategies is a history of fast reactor experience which dates back to the days of the Manhatten Project and includes the CLEMENTINE Reactor, which generated a few watts, LAMPRE, EBR-I, EBR-II, FERMI, SEFOR, FFTF, BR-1, -2, -5, -10, BOR-60, BN-350, BN-600, JOYO, RAPSODIE, Phenix, KNK-II, DFR, and PFR. Fast reactors under design or construction include PEC, CRBR, SuperPhenix, SNR-300, MONJU, and Madras (India). The parallel fuels and materials evolution has fully supported this reactor development. It has involved cermets, molten plutonium alloy, plutonium oxide, uranium metal or alloy, uranium oxide, and mixed uranium-plutonium oxides and carbides

  16. A Study of the Failure of Joints in Composite Material Fuel Cells Due to Hydraulic Ram Loading

    Science.gov (United States)

    1976-06-01

    H co PSw Z QW <H W CO 33 PS4 o CO O CM \\ Q> 00 vO m CO CM N ra Figure VI.B.l THICKNESS MODEL 55 it acts upon on the membrane, gives the force to be...ability of the joint to carry the loads created by hydraulic ram loading. It would also make the manufacturing procedure easier, less time consuming , and...70 less expensive. Cutting holes and channels in a composite plate not only alters the behavior and load carrying capa- bility of the plate, but it is

  17. Effects of high density dispersion fuel loading on the uncontrolled reactivity insertion transients of a low enriched uranium fueled material test research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Muhammad, Farhan [Department of Nuclear Engineering, Pakistan Institute of Engineering and Applied Sciences, Nilore, Islamabad 45650 (Pakistan)], E-mail: farhan73@hotmail.com; Majid, Asad [Department of Nuclear Engineering, Pakistan Institute of Engineering and Applied Sciences, Nilore, Islamabad 45650 (Pakistan)

    2009-08-15

    The effects of using high density low enriched uranium on the uncontrolled reactivity insertion transients of a material test research reactor were studied. For this purpose, the low density LEU fuel of an MTR was replaced with high density U-Mo (9w/o) LEU fuels currently being developed under the RERTR program having uranium densities of 6.57 gU/cm{sup 3}, 7.74 gU/cm{sup 3} and 8.57 gU/cm{sup 3}. Simulations were carried out to determine the reactor performance under reactivity insertion transients with totally failed control rods. Ramp reactivities of 0.25$/0.5 s and 1.35$/0.5 s were inserted with reactor operating at full power level of 10 MW. Nuclear reactor analysis code PARET was employed to carry out these calculations. It was observed that when reactivity insertion was 0.25$/0.5 s, the new power level attained increased by 5.8% as uranium density increases from 6.57 gU/cm{sup 3} to 8.90 gU/cm{sup 3}. This results in increased maximum temperatures of fuel, clad and coolant outlet, achieved at the new power level, by 4.7 K, 4.4 K and 2.4 K, respectively. When reactivity insertion was 1.35$/0.5 s, the feedback reactivities were unable to control the reactor which resulted in the bulk boiling of the coolant; the one with the highest fuel density was the first to reach the boiling point.

  18. Applying burnable poison particles to reduce the reactivity swing in high temperature reactors with batch-wise fuel loading

    International Nuclear Information System (INIS)

    Kloosterman, J.L.; Dam, H. van; Hagen, T.H.J.J. van der

    2003-01-01

    Burnup calculations have been performed on a standard HTR fuel pebble with a radius of 3 cm containing 9 g of 8% enriched uranium and burnable poison particles (BPP) made of B 4 C highly enriched in 10 B. The radius of the BPP and the number of particles per fuel pebble have been varied to find the flattest reactivity-to-time curve. It was found that for a k∞ of 1.1, a reactivity swing as low as 2% can be obtained when each fuel pebble contains about 1070 BPP with a radius of 75 μm. For coated BPP that consist of a graphite kernel with a radius of 300 μm covered with a B 4 C burnable poison layer, a similar value for the reactivity swing can be obtained. Cylindrical particles seem to perform worse. In general, the modification of the geometry of BPP is an effective means to tailor the reactivity curve of HTRs

  19. Statistical analysis of the vibration loading of the reactor internals and fuel assemblies of reactor units type WWER-440 from deferent projects

    International Nuclear Information System (INIS)

    Ovcharov, O.; Pavelko, V.; Usanov, A.; Arkadov, G.; Dolgov, A.; Molchanov, V.; Anikeev, J.; Pljush, A.

    2006-01-01

    In this paper the following items have been presented: 1) Vibration noise instrument channels; 2) Vibration loading characteristics of control assemblies, internals and design peculiarities of internals of WWER-440 deferent projects; 3) Coolant flow rate through the reactor, reactor core, fuel assemblies and control assemblies for different projects WWER-440 and 4) Noise measurements of coolant speed per channel. The change of auto power spectrum density of absolute displacement detector signal for the last 12 years of SUS monitoring of the Kola NPP unit 2; the coherence functions groups between two SPND of the same level for the Kola NPP unit 1; the measured coolant flow rate at Paks NPP and the auto power spectrum density group of SPND signals from 11 neutron measuring channels of the Kola NPP unit 1 are given. The main factors of vibration loading of internals and fuel assemblies for Kola NPP units 1-4, Bohunice NPP units 1 and 2 and Novovoronezh NPP units 3 and 4 are also discussed

  20. Part-Load Performance Prediction and Operation Strategy Design of Organic Rankine Cycles with a Medium Cycle Used for Recovering Waste Heat from Gaseous Fuel Engines

    Directory of Open Access Journals (Sweden)

    Xuan Wang

    2016-07-01

    Full Text Available The Organic Rankine Cycle (ORC is regarded as a suitable way to recover waste heat from gaseous fuel internal combustion engines. As waste heat recovery systems (WHRS have always been designed based on rated working conditions, while engines often work under part-load conditions, it is quite significant to analyze the part-load performance and corresponding operation strategy of ORC systems. This paper presents a dynamic model of ORC with a medium cycle used for a large gaseous fuel engine and analyzes the effect of adjustable parameters on the system performance, giving effective control directions under various conditions. The results indicate that the intermediary fluid mass flow rate has nearly no effect on the output power and thermal efficiency of the ORC, while the mass flow rate of working fluid has a great effect on them. In order to get a better system performance under different working conditions, the system should be operated with the working fluid mass flow rate as large as possible, but with a slight degree of superheating. Then, with the control of constant superheat degree at the end of the heating process, the performance of the combined system that consists of ORC and the engine at steady state under seven typical working conditions is also analyzed. The results indicate that the energy-saving effect of WHRS becomes worse and worse as the working condition decreases. Especially at 40% working condition the WHRS nearly has no energy-saving effect anymore.

  1. A study on physical characteristics of supercritical light - water reactor loaded with (232U-238Th-238U) oxide fuel

    International Nuclear Information System (INIS)

    Kulikov, E. G.; Shmelev, A. N.; Apse, V. A.; Kulikov, G. G.

    2007-01-01

    The attractiveness of using (U-Th)-fuel in supercritical light water reactor is considered. The dilution of 2 33U in 2 38U is proposed with the purpose of increasing non-proliferation of this fissile isotope. Comparison of different fuel compositions is accomplished from the point of view of fissile isotope breeding and achieved burn-up; parasitic neutron absorption cross-sections are also compared. It is analyzed the impact for neutron balance of both cladding materials: zirconium alloy and stainless steel

  2. Molten fuel behaviour during slow overpower transients

    International Nuclear Information System (INIS)

    Guerin, Y.; Boidron, M.

    1985-01-01

    In large commercial reactors as Super-Phenix, if we take into account all the uncertainties on the pins and on the core, it is no longer possible to guarantee the absence of fuel melting during incidental events such as slow overpower transients. We have then to explain what happens in the pins when fuel melting occurs and to demonstrate that a limited amount of molten fuel generates no risk of clad failure. For that purpose, we may use the results of a great number of experiments (about 40) that have been performed at C.E.A., most of them in thermal reactor, but some experiments have also been performed in Rapsodie, especially during the last run of this reactor. In a great part of these experiments, fuel melting occurred at beginning of life, but we have also some results at different burnups up to 5 at %. It is not the aim of this paper to describe all these experiments and the results of their post irradiation examination, but to summarize the main conclusions that have been set out of them and that have enabled us to determine the main characteristics of fuel element behaviour when fuel melting occurs

  3. Comparison of riparian and upland forest stand structure and fuel loads in beetle infested watersheds, southern Rocky Mountains

    Science.gov (United States)

    Kathleen A. Dwire; Robert Hubbard; Roberto Bazan

    2015-01-01

    Extensive outbreaks of mountain pine beetle (MPB), spruce beetle (SB), and other insects are altering forest stand structure throughout western North America, and thereby contributing to the heterogeneity of fuel distribution. In forested watersheds, conifer-dominated riparian forests frequently occur as narrow linear features in the landscape mosaic and contribute to...

  4. The IRSN has appraised the security of the TN 117 parcel model loaded with new or irradiated fuels

    International Nuclear Information System (INIS)

    2009-10-01

    This document reports the examination made by the IRSN on the security of a parcel design which is to be used for the transportation of new or irradiated nuclear fuels and more precisely for the transportation of fissile radionuclides with a high radioactivity. According to this content, this packaging must therefore comply with some radioactive material transport security regulations which correspond to specific mechanical, thermal and immersion requirements. After a description of the parcel design, this report comments the experimental and numerical results of the parcel structural analysis, and recommends some further investigations. It also briefly comments the results of a thermal analysis, and of a security-criticality analysis. The parcel security appears to depend on the fuel pencil watertightness

  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. Method for controlling a nuclear fueled electric power generating unit and interfacing the same with a load dispatching system

    International Nuclear Information System (INIS)

    Mueller, N.P.; Meyer, C.E.

    1984-01-01

    A pressurized water reactor (PWR) nuclear fueled, electric power generating unit is controlled through the use of on-line calculations of the rapid, step and ramp, power change capabilities of the unit made from measured values of power level, axial offset, coolant temperature and rod position taking into account operator generated, safety and control, and balance of plant limits. The power change capabilities so generated may be fed to an automatic dispatch system which provides closed loop control of a power grid system. (author)

  7. Exploring the Future of Fuel Loads in Tasmania, Australia: Shifts in Vegetation in Response to Changing Fire Weather, Productivity, and Fire Frequency

    Directory of Open Access Journals (Sweden)

    Rebecca Mary Bernadette Harris

    2018-04-01

    Full Text Available Changes to the frequency of fire due to management decisions and climate change have the potential to affect the flammability of vegetation, with long-term effects on the vegetation structure and composition. Frequent fire in some vegetation types can lead to transformational change beyond which the vegetation type is radically altered. Such feedbacks limit our ability to project fuel loads under future climatic conditions or to consider the ecological tradeoffs associated with management burns. We present a “pathway modelling” approach to consider multiple transitional pathways that may occur under different fire frequencies. The model combines spatial layers representing current and future fire danger, biomass, flammability, and sensitivity to fire to assess potential future fire activity. The layers are derived from a dynamically downscaled regional climate model, attributes from a regional vegetation map, and information about fuel characteristics. Fire frequency is demonstrated to be an important factor influencing flammability and availability to burn and therefore an important determinant of future fire activity. Regional shifts in vegetation type occur in response to frequent fire, as the rate of change differs across vegetation type. Fire-sensitive vegetation types move towards drier, more fire-adapted vegetation quickly, as they may be irreversibly impacted by even a single fire, and require very long recovery times. Understanding the interaction between climate change and fire is important to identify appropriate management regimes to sustain fire-sensitive communities and maintain the distribution of broad vegetation types across the landscape.

  8. Pd-Pt loaded graphene aerogel on nickel foam composite as binder-free anode for a direct glucose fuel cell unit

    Science.gov (United States)

    Tsang, Chi Him A.; Leung, D. Y. C.

    2017-09-01

    Fabrication of electrocatalyst for direct glucose fuel cell (DGFC) operation involves destructive preparation methods with the use of stabilizer like binder, which may cause activity depreciation. Binder-free electrocatalytic electrode becomes a possible solution to the above problem. Binder-free bimetallic Pd-Pt loaded graphene aerogel on nickel foam plates with different Pd/Pt ratios (1:2.32, 1:1.62, and 1:0.98) are successfully fabricated through a green one-step mild reduction process producing a Pd-Pt/GO/nickel form plate (NFP) composite. Anode with the binder-free electrocatalysts exhibit a strong activity in a batch type DGFC unit under room temperature. The effects of glucose and KOH concentrations, and the Pd/Pt ratios of the electrocatalyst on the DGFC performance are also studied. Maximum power density output of 1.25 mW cm-2 is recorded with 0.5 M glucose/3 M KOH as the anodic fuel, and Pd1Pt0.98/GA/NFP as catalyst, which is the highest obtained so far among other types of electrocatalyst.

  9. Estimation of average burnup of damaged fuels loaded in Fukushima Dai-ichi reactors by using the 134Cs/137Cs ratio method

    International Nuclear Information System (INIS)

    Endo, T.; Sato, S.; Yamamoto, A.

    2012-01-01

    Average burnup of damaged fuels loaded in Fukushima Dai-ichi reactors is estimated, using the 134 Cs/ 137 Cs ratio method for measured radioactivities of 134 Cs and 137 Cs in contaminated soils within the range of 100 km from the Fukushima Dai-ichi nuclear power plants. As a result, the measured 134 Cs/ 137 Cs ratio from the contaminated soil is 0.996±0.07 as of March 11, 2011. Based on the 134 Cs/ 137 Cs ratio method, the estimated burnup of damaged fuels is approximately 17.2±1.5 [GWd/tHM]. It is noted that the numerical results of various calculation codes (SRAC2006/PIJ, SCALE6.0/TRITON, and MVP-BURN) are almost the same evaluation values of 134 Cs/ 137 Cs ratio with same evaluated nuclear data library (ENDF-B/VII.0). The void fraction effect in depletion calculation has a major impact on 134 Cs/ 137 Cs ratio compared with the differences between JENDL-4.0 and ENDF-B/VII.0. (authors)

  10. High Pt Loading on Polydopamine Functionalized Graphene as a High Performance Cathode Electrocatalyst for Proton Exchange Membrane Fuel Cells

    Directory of Open Access Journals (Sweden)

    Monireh Faraji

    2016-04-01

    Full Text Available Morphology and size of platinum nanoparticles are a crucial factor in improving their catalytic activity and stability. Here, we firstly report the synthesis of high loading Pt nanoparticles on polydopamine reduced Graphene. The loading concentration of Pt (nanoparticles NPs on Graphene can be adjusted in the range of 60-70%.With the insertion of polydopamine between Graphene oxide sheets, stacking of Graphene can be effectively prevented, promoting diffusion of oxygen molecules through the Graphene sheets and enhancing the oxygen reduction reaction electrocatalytic activity. Compared to commercial catalysts (i.e., state-of-the-art Pt/C catalyst the as synthesized Pt supported polydopamine grafted reduced graphite oxide (Pt@PDA-rGO hybrid displays very high oxygen reduction reaction catalytic activities. We propose a unique 2D profile of the polydopamine-rGO role as a barrier preventing leaching of Pt into the electrolyte. The fabricated electrodes were evaluated with electrochemical techniques for oxygen reduction reaction and the obtained results were further verified by the transmission electron microscopy micrographs on the microstructure of the integrated pt@PDA-rGO structures. It has been revealed that the electrochemical impedance spectroscopy technique can provide more explicit information than polarization curves on the performance dependence on charge-transfer and mass transport processes at different overpotential regions.

  11. Reversible Operation of Solid Oxide Cells for Sustainable Fuel Production and Solar/Wind Load-Balancing

    DEFF Research Database (Denmark)

    Graves, Christopher R.; Villarreal, D.; Mýrdal, Jón Steinar Garðarsson

    2016-01-01

    with focus onfundamentals or applications of bi-directional operation. This presentation will highlight ourrecent developments in applying reversible SOCs (RSOCs) for renewable energy storagewith respect to cell and stack testing, cell and system design, and techno-economicanalysis.At the cell level, long...... exceeds the wind power supply.At the system level, techno-economic analyses and system designs for different scalesand applications have been realized. A simulation of an RSOC system that uses real-worldtime-series market prices for electricity and natural gas in Denmark to decide when tooperate...... in electrolysis mode (buying electricity and selling methane) or fuel-cell mode(buying gas and selling electricity) shows the advantage of a reversible system and thechanging operating profile as the fraction of wind power supply grows. Finally, we discussthe potential for systems with novel chemistries...

  12. Combustion and emissions characteristics of a spark-ignition engine fueled with hydrogen–methanol blends under lean and various loads conditions

    International Nuclear Information System (INIS)

    Zhang, Bo; Ji, Changwei; Wang, Shuofeng; Liu, Xiaolong

    2014-01-01

    Methanol is a promising alternative fuel for the spark-ignition engines. This paper experimentally investigated the performance of a hydrogen-blended methanol engine at lean and various load conditions. The test was conducted on a four-cylinder commercial spark-ignition engine equipped with an electronically controlled hydrogen port injection system. The test was conducted under a typical city driving speed of 1400 rpm and a constant excess air ratio of 1.20. Two hydrogen volume fractions in the intake of 0 and 3% were adopted to investigate the effect of hydrogen addition on combustion and emissions performance of the methanol engine. The test results showed that brake thermal efficiency was improved after the hydrogen addition. When manifolds absolute pressure increased from about 38 to 83 kPa, brake thermal efficiencies after the hydrogen addition were increased by 6.5% and 4.2%. The addition of hydrogen availed shortening flame development and propagation periods. The peak cylinder temperature was raised whereas cylinder temperature at the exhaust valve opening was decreased after the hydrogen addition. The addition of hydrogen contributed to the dropped hydrocarbon and carbon monoxide. However, nitrogen oxides were slightly raised after the hydrogen enrichment. - Highlights: • Load characteristics of a H 2 -blended methanol engine are experimentally studied. • H 2 addition is more effective on raising engine efficiency at low loads. • Flame development and propagation periods are shortened after H 2 addition. • H 2 enrichment contributes to the smooth operation of the methanol engine. • HC and CO emissions from the methanol engine are reduced after H 2 addition

  13. Influence of Compression Ratio on High Load Performance and Knock Behavior for Gasoline Port-Fuel Injection, Natural Gas Direct Injection and Blended Operation in a Spark Ignition Engine

    Energy Technology Data Exchange (ETDEWEB)

    Pamminger, Michael; Sevik, James; Scarcelli, Riccardo; Wallner, Thomas; Hall, Carrie

    2017-03-28

    Natural Gas (NG) is an alternative fuel which has attracted a lot of attention recently, in particular in the US due to shale gas availability. The higher hydrogen-to-carbon (H/C) ratio, compared to gasoline, allows for decreasing carbon dioxide emissions throughout the entire engine map. Furthermore, the high knock resistance of NG allows increasing the efficiency at high engine loads compared to fuels with lower knock resistance. NG direct injection (DI) allows for fuel to be added after intake valve closing (IVC) resulting in an increase in power density compared to an injection before IVC. Steady-state engine tests were performed on a single-cylinder research engine equipped with gasoline (E10) port-fuel injection (PFI) and NG DI to allow for in-cylinder blending of both fuels. Knock investigations were performed at two discrete compression ratios (CR), 10.5 and 12.5. Operating conditions span mid-load, wide-open-throttle and boosted conditions, depending on the knock response of the fuel blend. Blended operation was performed using E10 gasoline and NG. An additional gasoline type fuel (E85) with higher knock resistance than E10 was used as a high-octane reference fuel, since the octane rating of E10-NG fuel blends is unknown. Spark timing was varied at different loads under stoichiometric conditions in order to study the knock response as well as the effects on performance and efficiency. As anticipated, results suggest that the knock resistance can be increased significantly by increasing the NG amount. Comparing the engine operation with the least knock resistant fuel, E10 PFI, and the fuel blend with the highest knock resistance, 75% NG DI, shows an increase in indicated mean effective pressure of about 9 bar at CR 12.5. The usage of reference fuels with known knock characteristics allowed an assessment of knock characteristic of intermediate E10-NG blend levels. Mathematical correlations were developed allowing characterizing the occurrence of knocking

  14. Effects of carbon brush anode size and loading on microbial fuel cell performance in batch and continuous mode

    KAUST Repository

    Lanas, Vanessa; Ahn, Yongtae; Logan, Bruce E.

    2014-01-01

    Larger scale microbial fuel cells (MFCs) require compact architectures to efficiently treat wastewater. We examined how anode-brush diameter, number of anodes, and electrode spacing affected the performance of the MFCs operated in fed-batch and continuous flow mode. All anodes were initially tested with the brush core set at the same distance from the cathode. In fed-batch mode, the configuration with three larger brushes (25 mm diameter) produced 80% more power (1240 mW m-2) than reactors with eight smaller brushes (8 mm) (690 mW m-2). The higher power production by the larger brushes was due to more negative and stable anode potentials than the smaller brushes. The same general result was obtained in continuous flow operation, although power densities were reduced. However, by moving the center of the smaller brushes closer to the cathode (from 16.5 to 8 mm), power substantially increased from 690 to 1030 mW m-2 in fed batch mode. In continuous flow mode, power increased from 280 to 1020 mW m-2, resulting in more power production from the smaller brushes than the larger brushes (540 mW m-2). These results show that multi-electrode MFCs can be optimized by selecting smaller anodes, placed as close as possible to the cathode. © 2013 Elsevier B.V. All rights reserved.

  15. Effects of carbon brush anode size and loading on microbial fuel cell performance in batch and continuous mode

    KAUST Repository

    Lanas, Vanessa

    2014-02-01

    Larger scale microbial fuel cells (MFCs) require compact architectures to efficiently treat wastewater. We examined how anode-brush diameter, number of anodes, and electrode spacing affected the performance of the MFCs operated in fed-batch and continuous flow mode. All anodes were initially tested with the brush core set at the same distance from the cathode. In fed-batch mode, the configuration with three larger brushes (25 mm diameter) produced 80% more power (1240 mW m-2) than reactors with eight smaller brushes (8 mm) (690 mW m-2). The higher power production by the larger brushes was due to more negative and stable anode potentials than the smaller brushes. The same general result was obtained in continuous flow operation, although power densities were reduced. However, by moving the center of the smaller brushes closer to the cathode (from 16.5 to 8 mm), power substantially increased from 690 to 1030 mW m-2 in fed batch mode. In continuous flow mode, power increased from 280 to 1020 mW m-2, resulting in more power production from the smaller brushes than the larger brushes (540 mW m-2). These results show that multi-electrode MFCs can be optimized by selecting smaller anodes, placed as close as possible to the cathode. © 2013 Elsevier B.V. All rights reserved.

  16. An experimental study of a hydrogen-enriched ethanol fueled Wankel rotary engine at ultra lean and full load conditions

    International Nuclear Information System (INIS)

    Amrouche, F.; Erickson, P.A.; Varnhagen, S.; Park, J.W.

    2016-01-01

    Highlights: • H_2 was added at the intake of a single-rotor ethanol fueled Wankel engine. • The engine was operating at ultra-lean condition, WOT and 3000 rpm. • H_2 enrichment helps shortening the burn duration, enhance the thermal efficiency and reduce the BSEC. • H_2 addition helps to reduce HC, CO and CO_2 emissions. - Abstract: In this paper, the effect of hydrogen addition to ethanol in a monorotor Wankel engine at wide open throttle position and in an ultra-lean operating regime was experimentally investigated. For this aim, variation of hydrogen enrichment levels on the ethanol engine performance and emissions were considered. Experiments were carried out under a constant engine speed of 3000 rpm and fixed spark timing of 15 °BTDC. The test results showed that hydrogen enrichment improved the combustion process through shortening of the flame development and flame propagation periods and reducing the cyclic variation. Furthermore, the reduction of burn duration with the increase of hydrogen fraction enhances the thermal efficiency, reducing the brake-specific energy consumption, as well as reducing the unburned hydrocarbons emissions of the Wankel engine.

  17. Overview of the fast reactors fuels program. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Evans, E.A.; Cox, C.M.; Hayward, B.R.; Rice, L.H.; Yoshikawa, H.H.

    1980-04-01

    Each nation involved in LMFBR development has its unique energy strategies which consider energy growth projections, uranium resources, capital costs, and plant operational requirements. Common to all of these strategies is a history of fast reactor experience which dates back to the days of the Manhatten Project and includes the CLEMENTINE Reactor, which generated a few watts, LAMPRE, EBR-I, EBR-II, FERMI, SEFOR, FFTF, BR-1, -2, -5, -10, BOR-60, BN-350, BN-600, JOYO, RAPSODIE, Phenix, KNK-II, DFR, and PFR. Fast reactors under design or construction include PEC, CRBR, SuperPhenix, SNR-300, MONJU, and Madras (India). The parallel fuels and materials evolution has fully supported this reactor development. It has involved cermets, molten plutonium alloy, plutonium oxide, uranium metal or alloy, uranium oxide, and mixed uranium-plutonium oxides and carbides.

  18. Electrochemical Deposition of Platinum and Palladium on Gold Nanoparticles Loaded Carbon Nanotube Support for Oxidation Reactions in Fuel Cell

    Directory of Open Access Journals (Sweden)

    Surin Saipanya

    2014-01-01

    Full Text Available Pt and Pd sequentially electrodeposited Au nanoparticles loaded carbon nanotube (Au-CNT was prepared for the electrocatalytic study of methanol, ethanol, and formic acid oxidations. All electrochemical measurements were carried out in a three-electrode cell. A platinum wire and Ag/AgCl were used as auxiliary and reference electrodes, respectively. Suspension of the Au-CNT, phosphate buffer, isopropanol, and Nafion was mixed and dropped on glassy carbon as a working electrode. By sequential deposition method, PdPtPt/Au-CNT, PtPdPd/Au-CNT, and PtPdPt/Au-CNT catalysts were prepared. Cyclic voltammograms (CVs of those catalysts in 1 M H2SO4 solution showed hydrogen adsorption and hydrogen desorption reactions. CV responses for those three catalysts in methanol, ethanol, and formic acid electrooxidations studied in 2 M CH3OH, CH3CH2OH, and HCOOH in 1 M H2SO4 show characteristic oxidation peaks. The oxidation peaks at anodic scan contribute to those organic substance oxidations while the peaks at cathodic scan are related with the reoxidation of the adsorbed carbonaceous species. Comparing all those three catalysts, it can be found that the PdPtPt/Au-CNT catalyst is good at methanol oxidation; the PtPdPt/Au-CNT effectively enhances ethanol oxidation while the PtPdPd/Au-CNT exceptionally catalyzes formic acid oxidation. Therefore, a different stoichiometry affects the electrochemical active surface area of the catalysts to achieve the catalytic oxidation reactions.

  19. Reflector modelization in neutronic and optimization methods applied to fuel loading pattern; Modelisation du reflecteur en neutronique et methodes d`optimisation appliquees aux plans de rechargement

    Energy Technology Data Exchange (ETDEWEB)

    Argaud, J P

    1995-12-01

    I Physical description of P.W.R nuclear core can be handled by multigroup neutronic diffusion model. We are interested in two problems, using the same approach for the optimization aspect. To deal with some differences between calculations and measurements, the question of their reduction is then introduced. A reflector parameters identification from core measurements is then purposed, the reflector being at the present time the less known part of core diffusion model. This approach conducts to study the reflector model, in particular by an analysis of its transport origin. It leads finally to a new model of reflector described by boundary operators using an integral formulation on the core/reflector interface. That is on this new model that a parameter identification formulation of calculations-measurements differences reduction is given, using an adjoint state formulation to minimize errors by a gradient method. Furthermore, nuclear fuel reload of P.W.R core needs an optimal distribution of fuel assemblies, namely a loading pattern. This combinatorial optimization problem is then expressed as a cost function minimization, the cost function describing the power spatial distribution. Various methods (linear programming, simulated annealing,...), used to solve this problem, are detailed, given in particular a practical search example. A new approach is then proposed, using the gradient of the cost function to direct the search in the patterns discrete space. Final results of complete patterns search trials are presented, and compared to those obtained by other methods. In particular the results are obtained very quickly. (author). 81 refs., 55 figs., 5 appends.

  20. Seal performance of thermal aged metal gasket of dual purpose metal cask for interim spent fuel storage after external impact load

    International Nuclear Information System (INIS)

    Takeshi Yokoyama; Masami Kato; Satoshi Itooka

    2005-01-01

    As for interim storage for spent nuclear fuels using dual purpose dry metal cask in Japan, we recognize one of the important technical issues that there is a possibility for the cask with degraded metal gasket during storage to apply to transportation. In our study until 2003 focused on degradation of important components for the cask safety performance during storage and application to transportation, for metal gasket, we conducted property tests for degradation and influence of lid movement on seal performance, and furthermore verification tests. From the results, we developed the method to evaluate leak rate from lid with degraded metal gasket at accidents during transportation and in addition, we found as follows: Lid would hardly move and leak rate would not increase seriously during fire event. Leak rate from lid with degraded metal gasket could be evaluated by using results of leak rate trend depending on horizontal displacement of lid by external impact load. So, with regard to influence of lid movement on seal performance, we conducted additional test for extending horizontal displacement in lid moving in 2004. In addition, seal performance was discussed from the results, both previous and latest test. (authors)

  1. A Simple Power Management Scheme with Enhanced Stability for a Solar PV/Wind/Fuel Cell/Grid Fed Hybrid Power Supply Designed for Industrial Loads

    Directory of Open Access Journals (Sweden)

    S. Saravanan

    2014-01-01

    Full Text Available This paper proposes a new power conditioner topology with an intelligent power management controller that integrates multiple renewable energy sources such as solar energy, wind energy, and fuel cell energy with battery and AC grid supply as backup to make the best use of their operating characteristics with better reliability than that could be obtained by single renewable energy source based power supply. The proposed embedded controller is programmed to perform MPPT for solar PV panel and WTG, SOC estimation and battery, maintaining a constant voltage at PCC and power flow control by regulating the reference currents of the controller in an instantaneous basis. The instantaneous variation in reference currents of the controller enhances the controller response as it accommodates the effect of continuously varying solar insolation and wind speed in the power management. It also prioritizes the sources for consumption to achieve maximum usage of green energy than grid energy. The simulation results of the proposed power management system with real-time solar radiation and wind velocity data collected from solar centre, KEC, and experimental results for a sporadically varying load demand are presented in this paper and the results are encouraging from reliability and stability perspectives.

  2. Completion of UO{sub 2} pellets production and fuel rods load for the RA-8 critical facility; Finalizacion de la produccion de pastillas y carga de barras combustibles de UO{sub 2} para el conjunto critico RA-8

    Energy Technology Data Exchange (ETDEWEB)

    Marajofsky, Adolfo; Perez, Lidia E; Thern, Gerardo G; Altamirano, Jorge S; Benitez, Ana M; Cardenas, Hugo R; Becerra, Fabian A; Perez, Aldo E; Fuente, Mariano de la [Comision Nacional de Energia Atomica, General San Martin (Argentina). Dept. de Combustibles Nucleares

    1999-07-01

    The Advanced Fuels Division produced fuel pellets of {sup 235}U with 1.8% and 3.6% enrichment and Zry-4 cladding loads for the RA-8 reactor at Pilcaniyeu Technological Unit. For economical and availability reasons, the powder acquired was initially UO{sub 2} with 3.4% enrichment in {sup 235}U, therefore the {sup 235}U powder with 1.8% enrichment was produced by mechanical mixture. The production of fuel pellets for both enrichments was carried out by cold pressing and sintering processes in reducing atmosphere. The load of Zry-4 claddings was performed manually. The production stages can be divided into setup, qualification and production. This production allows not only to fulfill satisfactorily the new fuel rods supply for the RA-8 reactor but also to count with a new equipment and skilled personnel as well as to meet quality and assurance control methods for future pilot-scale production and even new fuel elements production. (author)

  3. 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.)

  4. Fuel assembly

    International Nuclear Information System (INIS)

    Nomata, Terumitsu.

    1993-01-01

    Among fuel pellets to be loaded to fuel cans of a fuel assembly, fuel pellets having a small thermal power are charged in a region from the end of each of spacers up to about 50mm on the upstream of coolants that flow vertically at the periphery of fuel rods. Coolants at the periphery of fuel rods are heated by the heat generation, to result in voids. However, since cooling effect on the upstream of the spacers is low due to influences of the spacers. Further, since the fuel pellets disposed in the upstream region have small thermal power, a void coefficient is not increased. Even if a thermal power exceeding cooling performance should be generated, there is no worry of causing burnout in the upstream region. Even if burnout should be caused, safety margin and reliability relative to burnout are improved, to increase an allowable thermal power, thereby enabling to improve integrity and reliability of fuel rods and fuel assemblies. (N.H.)

  5. Thermal conductivity of a graphite bipolar plate (BPP) and its thermal contact resistance with fuel cell gas diffusion layers: Effect of compression, PTFE, micro porous layer (MPL), BPP out-of-flatness and cyclic load

    Science.gov (United States)

    Sadeghifar, Hamidreza; Djilali, Ned; Bahrami, Majid

    2015-01-01

    This paper reports on measurements of thermal conductivity of a graphite bipolar plate (BPP) as a function of temperature and its thermal contact resistance (TCR) with treated and untreated gas diffusion layers (GDLs). The thermal conductivity of the BPP decreases with temperature and its thermal contact resistance with GDLs, which has been overlooked in the literature, is found to be dominant over a relatively wide range of compression. The effects of PTFE loading, micro porous layer (MPL), compression, and BPP out-of-flatness are also investigated experimentally. It is found that high PTFE loadings, MPL and even small BPP out-of-flatness increase the BPP-GDL thermal contact resistance dramatically. The paper also presents the effect of cyclic load on the total resistance of a GDL-BPP assembly, which sheds light on the behavior of these materials under operating conditions in polymer electrolyte membrane fuel cells.

  6. The thermal-mechanical behavior of fuel pins during power's maneuvering regime at stationary core loading on 2nd unit of KHNPP

    International Nuclear Information System (INIS)

    Ieremenko, M.; Ovdiyenko, Y.; Khalimonchuk, V.

    2007-01-01

    Results of thermal-mechanical behaviour of fuel pins during daily power's maneuvering regime that were proposed for second unit of Khmelnitsky NPP are presented. Calculations were performed for campaign's moments 100 and 160 fpd and for different type of regulation. Additionally calculations were performed for campaign 7. It is the design variant of the campaign and reactor core contains the high burnt fuel. Calculations of macro-core parameters (Kq, Kv) was performed by spatial computer code DYN3D. Calculations of micro-core parameters (fuel pin power) was performed by computer code DERAB. Calculations of thermal-mechanical behaviour of fuel pins was performed by computer code TRANSURANUS (Authors)

  7. PWR fuel thermomechanics

    International Nuclear Information System (INIS)

    Traccucci, R.; Leclercq, J.

    1986-01-01

    Fuel thermo-mechanics means the studies of mechanical and thermal effects, and more generally, the studies of the behavior of the fuel assembly under stresses including thermal and mechanical loads, hydraulic effects and phenomena induced by materials irradiation. This paper describes the studies dealing with the fuel assembly behavior, first in normal operating conditions, and then in accidental conditions. 43 refs [fr

  8. Uranium and thorium loadings determined by chemical and nondestructive methods in HTGR fuel rods for the Fort St. Vrain Early Validation Irradiation Experiment

    International Nuclear Information System (INIS)

    Angelini, P.; Rushton, J.E.

    1979-01-01

    The Fort St. Vrain Early Validation Irradiation Experiment is an irradiation test of reference and of improved High-Temperature Gas-Cooled Reactor fuels in the Fort St. Vrain Reactor. The irradiation test includes fuel rods fabricated at ORNL on an engineering scale fuel rod molding machine. Fuel rods were nondestructively assayed for 235 U content by a technique based on the detection of prompt-fission neutrons induced by thermal-neutron interrogation and were later chemically assayed by using the modified Davies Gray potentiometric titration method. The chemical analysis of the thorium content was determined by a volumetric titration method. The chemical assay method for uranium was evaluated and the results from the as-molded fuel rods agree with those from: (1) large samples of Triso-coated fissile particles, (2) physical mixtures of the three particle types, and (3) standard solutions to within 0.05%. Standard fuel rods were fabricated in order to evaluate and calibrate the nondestructive assay device. The agreement of the results from calibration methods was within 0.6%. The precision of the nondestructive assay device was established as approximately 0.6% by repeated measurements of standard rods. The precision was comparable to that estimated by Poisson statistics. A relative difference of 0.77 to 1.5% was found between the nondestructive and chemical determinations on the reactor grade fuel rods

  9. Design Package for Fuel Retrieval System Fuel Handling Tool Modification

    International Nuclear Information System (INIS)

    TEDESCHI, D.J.

    2000-01-01

    This is a design package that contains the details for a modification to a tool used for moving fuel elements during loading of MCO Fuel Baskets for the Fuel Retrieval System. The tool is called the fuel handling tool (or stinger). This document contains requirements, development design information, tests, and test reports

  10. High loading uranium plate

    International Nuclear Information System (INIS)

    Wiencek, T.C.; Domagala, R.F.; Thresh, H.R.

    1990-01-01

    Two embodiments of a high uranium fuel plate are disclosed which contain a meat comprising structured uranium compound confined between a pari of diffusion bonded ductile metal cladding plates uniformly covering the meat, the meat hiving a uniform high fuel loading comprising a content of uranium compound greater than about 45 Vol. % at a porosity not greater than about 10 Vol. %. In a first embodiment, the meat is a plurality of parallel wires of uranium compound. In a second embodiment, the meat is a dispersion compact containing uranium compound. The fuel plates are fabricated by a hot isostatic pressing process

  11. 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.)

  12. Optimal shifting of Photovoltaic and load fluctuations from fuel cell and electrolyzer to lead acid battery in a Photovoltaic/hydrogen standalone power system for improved performance and life time

    Science.gov (United States)

    Tesfahunegn, S. G.; Ulleberg, Ø.; Vie, P. J. S.; Undeland, T. M.

    Cost reduction is very critical in the pursuit of realizing more competitive clean and sustainable energy systems. In line with this goal a control method that enables minimization of the cost associated with performance and life time degradation of fuel cell and electrolyzer, and cost of battery replacement in PV/hydrogen standalone power systems is developed. The method uses the advantage of existing peak shaving battery to suppress short-term PV and load fluctuations while reducing impact on the cycle life of the battery itself. This is realized by diverting short-term cyclic charge/discharge events induced by PV/load power fluctuations to the upper band of the battery state of charge regime while operating the fuel cell and electrolyzer systems along stable (smooth) power curves. Comparative studies of the developed method with two other reference cases demonstrate that the proposed method fares better with respect to defined performance indices as fluctuation suppression rate and mean state of charge. Modeling of power electronics and design of controllers used in the study are also briefly discussed in Appendix A.

  13. Status of load management

    Energy Technology Data Exchange (ETDEWEB)

    Juchymenko, A

    1983-08-01

    A summary is presented of the status of load management, defined as any activity by an electric utility to affect the size and characteristics of its load. Load management is currently viewed by electric utilities as an important tool for marketing electricity in a competitive fuel situation. A major aim of the National Energy Program is to reduce Canada's dependence on oil by 1990 to 10% of the energy used by all markets. As a result, electricity may play a greater role in the supply of primary energy. Research in load management has been directed mostly towards the residential market, especially direct control of domestic hot water heaters and air conditioners. Studies conducted in Canada and the U.S. to determine user's receptiveness to direct control of loads and thermal energy storage systems indicate that these load management techniques are in most cases not acceptable to customers, who prefer voluntary reduction in demand. The potential exists in the industrial market to use load management to assist in electrifying many of the fossil fuel-fired processes at competitive energy prices. Some of the more important applications include an industrial heat pump to heat liquids to 120{degree}C, induction heating for melting and heat treating of metals, and mechanical vapor recompression equipment to produce proces steam. 21 refs., 2 figs., 2 tabs.

  14. Optimized core loading sequence for Ukraine WWER-1000 reactors

    International Nuclear Information System (INIS)

    Dye, M.; Shah, H.

    2015-01-01

    Fuel Assemblies (WFAs) experienced mechanical damage of the grids during loading at both South Ukraine 2 (SU2) and South Ukraine 3 (SU3). The grids were damaged due to high lateral loads exceeding their strength limit. The high lateral loads were caused by a combination of distortion and stiffness of the mixed core fuel assemblies and significant fuel assembly-to-fuel assembly interaction combined with the core loading sequence being used. To prevent damage of the WFA grids during core loading, Westinghouse has developed a loading sequence technique and loading aides (smooth sided dummies and top nozzle loading guides) designed to minimize fuel assembly-to-fuel assembly interaction while maximizing the potential for successful loading (i.e., no fuel assembly damage and minimized loading time). The loading sequence technique accounts for cycle-specific core loading patterns and is based on previous Westinghouse WWER core loading experience and fundamental principles. The loading aids are developed to “open-up” the target core location or to provide guidance into a target core location. The Westinghouse optimized core loading sequence and smooth sided dummies were utilized during the successful loading of SU3 Cycle 25 mixed core in March 2015, with no instances of fuel assembly damage and yet still provided considerable time savings relative to the 2012 and 2013 SU3 reload campaigns. (authors)

  15. Forest attributes and fuel loads of riparian vs. upland stands in mountain pine beetle infested watersheds, southern Rocky Mountains [Chapter 13

    Science.gov (United States)

    Kathleen A. Dwire; Roberto A. Bazan; Robert Hubbard

    2015-01-01

    Extensive outbreaks of mountain pine beetle (MPB), spruce beetle (SB), and other insects are altering forest stand structure throughout the Western United States, and thereby increasing the natural heterogeneity of fuel distribution. Riparian forests frequently occur as narrow linear features in the landscape mosaic and can contribute to the spatial complexity of...

  16. Fuel cycle cost study with HEU and LEU fuels

    International Nuclear Information System (INIS)

    Matos, J.E.; Freese, K.E.

    1984-01-01

    Fuel cycle costs are compared for a range of 235 U loadings with HEU and LEU fuels using the IAEA generic 10 MW reactor as an example. If LEU silicide fuels are successfully demonstrated and licensed, the results indicate that total fuel cycle costs can be about the same or lower than those with the HEU fuels that are currently used in most research reactors

  17. A fuel cycle cost study with HEU and LEU fuels

    International Nuclear Information System (INIS)

    Matos, J.E.; Freese, K.E.

    1985-01-01

    Fuel cycle costs are compared for a range of 235 U loadings with HEU and LEU fuels using the IAEA generic 10 MW reactor as an example. If LEU silicide fuels are successfully demonstrated and licensed, the results indicate that total fuel cycle costs can be about the same or lower than those with the HEU fuels that are currently used in most research reactors. (author)

  18. A fuel cycle cost study with HEU and LEU fuels

    Energy Technology Data Exchange (ETDEWEB)

    Matos, J E; Freese, K E [Argonne National Laboratory, Argonne, IL (United States)

    1985-07-01

    Fuel cycle costs are compared for a range of {sup 235}U loadings with HEU and LEU fuels using the IAEA generic 10 MW reactor as an example. If LEU silicide fuels are successfully demonstrated and licensed, the results indicate that total fuel cycle costs can be about the same or lower than those with the HEU fuels that are currently used in most research reactors. (author)

  19. Three years re-use of low burned fuel assemblies from units 1 and 2 in core loading of units 3 and 4 WWER-440 at Kozloduy NPP

    International Nuclear Information System (INIS)

    Stoyanova, I.; Antov, A.; Spasova, V.

    2006-01-01

    At the end of 2002 Units 1 and 2 of Kozloduy NPP were shutdown before the end of their designed lifetime. This left a large number of assemblies yet with a significant energy resource. In 2003, 43 assemblies from Unit 1 and 55 assemblies from Unit 2 are loaded in the cores of Units 3 and 4 respectively. In 2004, new 49 assemblies from Unit 1 and new 55 assemblies from Unit 2 are loaded in the cores of Units 3 and 4 respectively. In 2005, the next new 25 assemblies from Unit 1 and 66 assemblies from Unit 2 are loaded in the cores of Units 3 and 4 respectively. The realized core loadings of Units 3 and 4 are illustrated. The large number of low burned assemblies after the shut down of Units 1 and 2 and their utilization during 4 consecutive cycles in the cores of Units 3 and 4 both do not permit to obtain in the last 3 cycles the typical mean burn up 35-36 MWd/kg U of 3.6% discharged assemblies

  20. Effect of bundle junction face and misalignment on the pressure drops across a randomly loaded and aligned 12 bundles in CANDU fuel channel

    Energy Technology Data Exchange (ETDEWEB)

    Suk, H. C.; Sim, K. S.; Chang, C. H.; Lee, Y. O. [Korea Atomic Energy Reaearch Institute, Taejon (Korea, Republic of)

    1996-06-01

    The pressure drop of twelve fuel bundle string in the CANDU-6 fuel channel is equal to the sum of the eleven junction pressure losses, the bundle string entrance and exit pressure losses, the skin friction pressure loss, and other appendage pressure losses, where the junction loss is dependent on the bundle and faces and angular alignments of the junctions. The results of the single junction pressure drop tests in a short rig show that the most probable pressure drop of the eleven junction was analytically equal to the eleven times of average pressure drop of all the possible single junction pressure drops, and also that the largest and smallest junction pressure drops across the eleven junctions probably occurred only with BA and BB type junctions, respectively, where A and B denote the bundle end sides with an end-plates on which a company monogram is stamped and unstamped, respectively. 5 refs., 7 figs., 1 tab. (author).

  1. 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.)

  2. Loads imposed on dual purpose casks in German on-site-storage facilities for long term intermediate storage of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Wetzel, N.; Rabe, O. [TUeV NORD EnSys Hannover GmbH und Co. KG, Hanover (Germany)

    2004-07-01

    In accordance with recent changes of the atomic energy act and in order to secure reliable removal of spent fuel from the nuclear power plants' fuel storage ponds the German utilities filed license applications for a total of 12 onsite- storage facilities for spent fuel assemblies. By the end of 2003 the last of these storage facilities were licensed and are currently under construction. The first on-site-storage facility of that line became operational in late 2002. There are several design lines of storage facilities with different handling procedures or possible accident conditions. Short term interim storage facilities for a few casks are characterized by individual concrete hoods shielding the casks in horizontal position whereas long term intermediate storage facilities currently erected for large numbers of casks typically feature a condensed pattern of casks stored in upright position and massive structures of reinforced concrete. TUeV Hannover/Sachsen-Anhalt e. V. (now TUeV NORD EnSys Hannover GmbH and Co. KG) has been contracted as a body of independent experts for the assessment of all related safety requirements on behalf of the national licensing authority, the federal office for radiation protection (BfS).

  3. Loads imposed on dual purpose casks in German on-site-storage facilities for long term intermediate storage of spent nuclear fuel

    International Nuclear Information System (INIS)

    Wetzel, N.; Rabe, O.

    2004-01-01

    In accordance with recent changes of the atomic energy act and in order to secure reliable removal of spent fuel from the nuclear power plants' fuel storage ponds the German utilities filed license applications for a total of 12 onsite- storage facilities for spent fuel assemblies. By the end of 2003 the last of these storage facilities were licensed and are currently under construction. The first on-site-storage facility of that line became operational in late 2002. There are several design lines of storage facilities with different handling procedures or possible accident conditions. Short term interim storage facilities for a few casks are characterized by individual concrete hoods shielding the casks in horizontal position whereas long term intermediate storage facilities currently erected for large numbers of casks typically feature a condensed pattern of casks stored in upright position and massive structures of reinforced concrete. TUeV Hannover/Sachsen-Anhalt e. V. (now TUeV NORD EnSys Hannover GmbH and Co. KG) has been contracted as a body of independent experts for the assessment of all related safety requirements on behalf of the national licensing authority, the federal office for radiation protection (BfS)

  4. Fuel rods

    International Nuclear Information System (INIS)

    Hattori, Shinji; Kajiwara, Koichi.

    1980-01-01

    Purpose: To ensure the safety for the fuel rod failures by adapting plenum springs to function when small forces such as during transportation of fuel rods is exerted and not to function the resilient force when a relatively great force is exerted. Constitution: Between an upper end plug and a plenum spring in a fuel rod, is disposed an insertion member to the lower portion of which is mounted a pin. This pin is kept upright and causes the plenum spring to function resiliently to the pellets against the loads due to accelerations and mechanical vibrations exerted during transportation of the fuel rods. While on the other hand, if a compression force of a relatively high level is exerted to the plenum spring during reactor operation, the pin of the insertion member is buckled and the insertion member is inserted to the inside of the plenum spring, whereby the pellets are allowed to expand freely and the failures in the fuel elements can be prevented. (Moriyama, K.)

  5. 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.)

  6. Preliminary design and analysis on nuclear fuel cycle for fission-fusion hybrid spent fuel burner

    International Nuclear Information System (INIS)

    Chen Yan; Wang Minghuang; Jiang Jieqiong

    2012-01-01

    A wet-processing-based fuel cycle and a dry-processing were designed for a fission-fusion hybrid spent fuel burner (FDS-SFB). Mass flow of SFB was preliminarily analyzed. The feasibility analysis of initial loaded fuel inventory, recycle fuel fabrication and spent fuel reprocessing were preliminarily evaluated. The results of mass flow of FDS-SFB demonstrated that the initial loaded fuel inventory, recycle fuel fabrication and spent fuel reprocessing of nuclear fuel cycle of FDS-SFB is preliminarily feasible. (authors)

  7. Postirradiation examination of high-U-loaded, low-enriched U3O8, UAl2, and U3Si test fuel plates

    International Nuclear Information System (INIS)

    Gomez, J.; Morando, R.; Perez, E.E.; Giorsetti, D.R.; Copeland, G.L.; Hofman, G.L.; Snelgrove, J.L.

    1985-01-01

    The scope of this work is to present an evaluation of the postirradiation examination of the second set of high-U-loaded, low-enriched U 3 O 8 , UAl 2 and U 3 Si miniature plates manufactured by the Comision Nacional de Energia Atomica (CNEA) of Argentina, and irradiated and examined, within the framework of the Reduced Enrichment Research and Test Reactor (RERTR) Program, at Oak Ridge National Laboratory and Argonne National Laboratory. This paper includes fabrication details of the plates, their irradiation history and the results of postirradiation examination which are compared to those of the previous test and to present results from other laboratories participating in the RERTR Program. Postirradiation examination of these plates showed satisfactory performance for the oxides, aluminides and silicides (except for the highest-loaded U 3 Si plate) with the only indication of detrimental behavior being the slight bowing of some plates at about 80% burnup

  8. Postirradiation examination of high-U-loaded, low-enriched U3O8, UAl2, and U3Si test fuel plates

    International Nuclear Information System (INIS)

    Gomez, J.; Morando, R.; Perez, E.E.; Giorsetti, D.R.; Copeland, G.L.; Hofman, G.L.; Snelgrove, J.L.

    1985-01-01

    The scope of this work is to present an evaluation of the postirradiation examination of the second set of high-U-loaded, low-enriched U 3 O 8 , UAl 2 and U 3 Si miniature plates manufactured by the Comision Nacional de Energia Atomica (CNEA) of Argentina, and irradiated and examined, within the framework of the Reduced Enrichment Research and Test Reactor (RERTR) Program, at Oak Ridge National Laboratory and Argonne National Laboratory. This paper includes fabrication details of the plates, their irradiation history and the results of postirradiation examination which are compared to those of the previous test and to present results from other laboratories participating in the REM Program. Postirradiation examination of these plates showed satisfactory performance for the oxides, aluminides and silicides (except for the highest-loaded U 3 Si plate) with the only indication of detrimental behavior being the slight bowing of some plates at about 80% burnup. (author)

  9. Postirradiation examination of high-U-loaded low-enriched U3O8, UAl2, and U3Si test fuel plates

    International Nuclear Information System (INIS)

    Gomez, J.; Morando, R.; Perez, E.E.; Giorsetti, D.R.; Copeland, G.L.; Hofmann, G.; Snelgrove, J.L.

    1984-01-01

    The scope of this work is to present an evaluation of the postirradiation examination of the second set of high-U-loaded low-enriched U 3 O 8 , UAl 2 and U 3 Si miniature plates manufactured by the Comision Nacional de Energia Atomica (CNEA) of Argentina, and irradiated and examinated, within the framework of the Reduced Enrichment Research and Test Reactor (RERTR) Program, at Oak Ridge National Laboratory and Argonne National Laboratory. This paper includes fabrication details of the plates, their irradiation history and the results of postirradiation examination which are compared to those of the previous test and to present results from other laboratories participating in the RERTR Program. Postirradiation examination of these plates showed satisfactory poerformance for the oxides, aluminides and silicides (except for the highest-loaded U 3 Si plate) with the only indication of detrimental behavior during the slight bowing of some plates at about 80% burnup

  10. Nuclear fuel accounting

    International Nuclear Information System (INIS)

    Aisch, D.E.

    1977-01-01

    After a nuclear power plant has started commercial operation the actual nuclear fuel costs have to be demonstrated in the rate making procedure. For this purpose an accounting system has to be developed which comprises the following features: 1) All costs associated with nuclear fuel shall be correctly recorded; 2) it shall be sufficiently flexible to cover also deviations from proposed core loading patterns; 3) it shall be applicable to different fuel cycle schemes. (orig./RW) [de

  11. Power to Fuels: Dynamic Modeling of a Slurry Bubble Column Reactor in Lab-Scale for Fischer Tropsch Synthesis under Variable Load of Synthesis Gas

    Directory of Open Access Journals (Sweden)

    Siavash Seyednejadian

    2018-03-01

    Full Text Available This research developed a comprehensive computer model for a lab-scale Slurry Bubble Column Reactor (SBCR (0.1 m Dt and 2.5 m height for Fischer–Tropsch (FT synthesis under flexible operation of synthesis gas load flow rates. The variable loads of synthesis gas are set at 3.5, 5, 7.5 m3/h based on laboratory adjustments at three different operating temperatures (483, 493 and 503 K. A set of Partial Differential Equations (PDEs in the form of mass transfer and chemical reaction are successfully coupled to predict the behavior of all the FT components in two phases (gas and liquid over the reactor bed. In the gas phase, a single-bubble-class-diameter (SBCD is adopted and the reduction of superficial gas velocity through the reactor length is incorporated into the model by the overall mass balance. Anderson Schulz Flory distribution is employed for reaction kinetics. The modeling results are in good agreement with experimental data. The results of dynamic modeling show that the steady state condition is attained within 10 min from start-up. Furthermore, they show that step-wise syngas flow rate does not have a detrimental influence on FT product selectivity and the dynamic modeling of the slurry reactor responds quite well to the load change conditions.

  12. The spent fuel safety experiment

    International Nuclear Information System (INIS)

    Harmms, G.A.; Davis, F.J.; Ford, J.T.

    1995-01-01

    The Department of Energy is conducting an ongoing investigation of the consequences of taking fuel burnup into account in the design of spent fuel transportation packages. A series of experiments, collectively called the Spent Fuel Safety Experiment (SFSX), has been devised to provide integral benchmarks for testing computer-generated predictions of spent fuel behavior. A set of experiments is planned in which sections of unirradiated fuel rods are interchanged with similar sections of spent PWR fuel rods in a critical assembly. By determining the critical size of the arrays, one can obtain benchmark data for comparison with criticality safety calculations. The integral reactivity worth of the spent fuel can be assessed by comparing the measured delayed critical fuel loading with and without spent fuel. An analytical effort to model the experiments and anticipate the core loadings required to yield the delayed critical conditions runs in parallel with the experimental effort

  13. Safety analysis of MOX fuels by fuel performance code

    Energy Technology Data Exchange (ETDEWEB)

    Suzuki, Motoe [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2002-12-01

    Performance of plutonium rick mixed oxide fuels specified for the Reduced-Moderation Water Reactor (RMWR) has been analysed by modified fuel performance code. Thermodynamic properties of these fuels up to 120 GWd/t burnup have not been measured and estimated using existing uranium fuel models. Fission product release, pressure rise inside fuel rods and mechanical loads of fuel cans due to internal pressure have been preliminarily assessed based on assumed axial power distribution history, which show the integrity of fuel performance. Detailed evaluation of fuel-cladding interactions due to thermal expansion or swelling of fuel pellets due to high burnup will be required for safety analysis of mixed oxide fuels. Thermal conductivity and swelling of plutonium rich mixed oxide fuels shall be taken into consideration. (T. Tanaka)

  14. Romanian nuclear fuel program

    International Nuclear Information System (INIS)

    Budan, O.

    1999-01-01

    The paper presents and comments the policy adopted in Romania for the production of CANDU-6 nuclear fuel before and after 1990. The CANDU-6 nuclear fuel manufacturing started in Romania in December 1983. Neither AECL nor any Canadian nuclear fuel manufacturer were involved in the Romanian industrial nuclear fuel production before 1990. After January 1990, the new created Romanian Electricity Authority (RENEL) assumed the responsibility for the Romanian Nuclear Power Program. It was RENEL's decision to stop, in June 1990, the nuclear fuel production at the Institute for Nuclear Power Reactors (IRNE) Pitesti. This decision was justified by the Canadian specialists team findings, revealed during a general, but well enough technically founded analysis performed at IRNE in the spring of 1990. All fuel manufactured before June 1990 was quarantined as it was considered of suspect quality. By that time more than 31,000 fuel bundles had already been manufactured. This fuel was stored for subsequent assessment. The paper explains the reasons which provoked this decision. The paper also presents the strategy adopted by RENEL after 1990 regarding the Romanian Nuclear Fuel Program. After a complex program done by Romanian and Canadian partners, in November 1994, AECL issued a temporary certification for the Romanian nuclear fuel plant. During the demonstration manufacturing run, as an essential milestone for the qualification of the Romanian fuel supplier for CANDU-6 reactors, 202 fuel bundles were produced. Of these fuel bundles, 66 were part of the Cernavoda NGS Unit 1 first fuel load (the balance was supplied by Zircatec Precision Industries Inc. ZPI). The industrial nuclear fuel fabrication re-started in Romania in January 1995 under AECL's periodical monitoring. In December 1995, AECL issued a permanent certificate, stating the Romanian nuclear fuel plant as a qualified and authorised CANDU-6 fuel supplier. The re-loading of the Cernavoda NGS Unit 1 started in the middle

  15. Power assisted fuel cell

    Energy Technology Data Exchange (ETDEWEB)

    Jarvis, L P; Atwater, T B; Plichta, E J; Cygan, P J [US Army CECOM, Fort Monmouth, NJ (United States). Research Development and Engineering Center

    1998-02-01

    A hybrid fuel cell demonstrated pulse power capability at pulse power load simulations synonymous with electronics and communications equipment. The hybrid consisted of a 25.0 W Proton Exchange Membrane Fuel Cell (PEMFC) stack in parallel with a two-cell lead-acid battery. Performance of the hybrid PEMFC was superior to either the battery or fuel cell stack alone at the 18.0 W load. The hybrid delivered a flat discharge voltage profile of about 4.0 V over a 5 h radio continuous transmit mode of 18.0 W. (orig.)

  16. TMI-2 spent fuel shipping

    International Nuclear Information System (INIS)

    Quinn, G.J.; Burton, H.M.

    1985-01-01

    TMI-2 failed fuel will be shipped to the Idaho National Engineering Laboratory for use in the DOE Core Examination Program. The fuel debris will be loaded into three types of canisters during defueling and dry loaded into a spent fuel shipping cask. The cask design accommodates seven canisters per cask and has two separate containment vessels with ''leaktight'' seals. Shipments are expectd to begin in early 1986

  17. Conceptual evaluation of hybrid energy system comprising wind-biomass-nuclear plants for load balancing and for production of renewable synthetic transport fuels

    International Nuclear Information System (INIS)

    Carlsson, Johan; Purvins, Arturs; Papaioannou, Ioulia T.; Shropshire, David; Cherry, Robert S.

    2014-01-01

    Future energy systems will increasingly need to integrate variable renewable energy in order to reduce greenhouse gas emissions from power production. Addressing this trend the present paper studies how a hybrid energy systems comprising aggregated wind farms, a biomass processing plant, and a nuclear cogeneration plant could support high renewable energy penetration. The hybrid energy system operates so that its electrical output tends to meet demand. This is achieved mainly through altering the heat-to-power ratio of the nuclear reactor and by using excess electricity for hydrogen production through electrolysis. Hybrid energy systems with biomass treatment processes, i.e. drying, torrefaction, pyrolysis and synthetic fuel production were evaluated. It was shown that the studied hybrid energy system comprising a 1 GWe wind farm and a 347 MWe nuclear reactor could closely follow the power demand profile with a standard deviation of 34 MWe. In addition, on average 600 m"3 of bio-gasoline and 750 m"3 bio-diesel are produced daily. The reduction of greenhouse gas emissions of up to 4.4 MtCO_2eq annually compared to power generation and transport using conventional fossil fuel sources. (author)

  18. Sensor system for fuel transport vehicle

    Science.gov (United States)

    Earl, Dennis Duncan; McIntyre, Timothy J.; West, David L.

    2016-03-22

    An exemplary sensor system for a fuel transport vehicle can comprise a fuel marker sensor positioned between a fuel storage chamber of the vehicle and an access valve for the fuel storage chamber of the vehicle. The fuel marker sensor can be configured to measure one or more characteristics of one or more fuel markers present in the fuel adjacent the sensor, such as when the marked fuel is unloaded at a retail station. The one or more characteristics can comprise concentration and/or identity of the one or more fuel markers in the fuel. Based on the measured characteristics of the one or more fuel markers, the sensor system can identify the fuel and/or can determine whether the fuel has been adulterated after the marked fuel was last measured, such as when the marked fuel was loaded into the vehicle.

  19. Application of a Coated Film Catalyst Layer Model to a High Temperature Polymer Electrolyte Membrane Fuel Cell with Low Catalyst Loading Produced by Reactive Spray Deposition Technology

    Directory of Open Access Journals (Sweden)

    Timothy D. Myles

    2015-10-01

    Full Text Available In this study, a semi-empirical model is presented that correlates to previously obtained experimental overpotential data for a high temperature polymer electrolyte membrane fuel cell (HT-PEMFC. The goal is to reinforce the understanding of the performance of the cell from a modeling perspective. The HT-PEMFC membrane electrode assemblies (MEAs were constructed utilizing an 85 wt. % phosphoric acid doped Advent TPS® membranes for the electrolyte and gas diffusion electrodes (GDEs manufactured by Reactive Spray Deposition Technology (RSDT. MEAs with varying ratios of PTFE binder to carbon support material (I/C ratio were manufactured and their performance at various operating temperatures was recorded. The semi-empirical model derivation was based on the coated film catalyst layer approach and was calibrated to the experimental data by a least squares method. The behavior of important physical parameters as a function of I/C ratio and operating temperature were explored.

  20. High performance polymer electrolyte fuel cells with ultra-low Pt loading electrodes prepared by dual ion-beam assisted deposition

    International Nuclear Information System (INIS)

    Saha, Madhu Sudan; Gulla, Andrea F.; Allen, Robert J.; Mukerjee, Sanjeev

    2006-01-01

    Ultra-low pure Pt-based electrodes (0.04-0.12 mg Pt /cm 2 ) were prepared by dual ion-beam assisted deposition (dual IBAD) method on the surface of a non-catalyzed gas diffusion layer (GDL) substrate. Film thicknesses ranged between 250 and 750 A, these are compared with a control, a conventional Pt/C (1.0 mg Pt(MEA) /cm 2 , E-TEK). The IBAD electrode constituted a significantly different morphology, where low density Pt deposits (largely amorphous) were formed with varying depths of penetration into the gas diffusion layer, exhibiting a gradual change towards increasing crystalline character (from 250 to 750 A). Mass specific power density of 0.297 g Pt /kW is reported with 250 A IBAD deposit (0.04 mg Pt /cm 2 for a total MEA loading of 0.08 mg Pt /cm 2 ) at 0.65 V. This is contrasted with the commercial MEA with a loading of 1 mg Pt(MEA) /cm 2 where mass specific power density obtained was 1.18 g Pt /kW (at 0.65 V), a value typical of current state of the art commercial electrodes containing Pt/C. The principal shortcoming in this effort is the area specific power density which was in the range of 0.27-0.43 W/cm 2 (for 250-750 A IBAD) at 0.65 V, hence much below the automotive target value of 0.8-0.9 W/cm 2 (at 0.65 V). An attempt to mitigate these losses is reported with the use of patterning. In this context a series of patterns ranging from 45 to 80% Pt coverage were used in conjunction with a hexagonal hole geometry. Up to 30% lowering of mass transport losses were realized

  1. Development and evaluation of carbon and binder loading in low-cost activated carbon cathodes for air-cathode microbial fuel cells

    KAUST Repository

    Wei, Bin; Tokash, Justin C.; Chen, Guang; Hickner, Michael A.; Logan, Bruce E.

    2012-01-01

    Activated carbon (AC) air cathodes were constructed using variable amounts of carbon (43-171 mg cm-2) and an inexpensive binder (10 wt% polytetrafluoroethylene, PTFE), and with or without a porous cloth wipe-based diffusion layer (DL) that was sealed with PDMS. The cathodes with the highest AC loading of 171 mg cm-2, and no diffusion layer, produced 1255 ± 75 mW m-2 and did not appreciably vary in performance after 1.5 months of operation. Slightly higher power densities were initially obtained using 100 mg cm-2 of AC (1310 ± 70 mW m-2) and a PDMS/wipe diffusion layer, although the performance of this cathode decreased to 1050 ± 70 mW m-2 after 1.5 months, and 1010 ± 190 mW m-2 after 5 months. AC loadings of 43 mg cm-2 and 100 mg cm-2 did not appreciably affect performance (with diffusion layers). MFCs with the Pt catalyst and Nafion binder initially produced 1295 ± 13 mW m-2, but the performance decreased to 930 ± 50 mW m -2 after 1.5 months, and then to 890 ± 20 mW m-2 after 5 months. Cathode performance was optimized for all cathodes by using the least amount of PTFE binder (10%, in tests using up to 40%). These results provide a method to construct cathodes for MFCs that use only inexpensive AC and a PTFE, while producing power densities similar to those of Pt/C cathodes. The methods used here to make these cathodes will enable further tests on carbon materials in order to optimize and extend the lifetime of AC cathodes in MFCs. © 2012 The Royal Society of Chemistry.

  2. Behavior and failure of uniformly hydrided Zircaloy-4 fuel claddings between 25 C and 480 C under various stress states, including RIA loading conditions

    International Nuclear Information System (INIS)

    Le Saux, M.; Carassou, S.; Averty, X.; Le Saux, M.; Besson, J.; Poussard, C.

    2010-01-01

    The anisotropic plastic behavior and the fracture of as-received and hydrided Cold-Worked Stress Relieved Zircaloy-4 cladding tubes are investigated under thermal-mechanical loading conditions representative of Pellet-Clad Mechanical Interaction during Reactivity Initiated Accidents in Pressurized Water Reactors. In order to study the combined effects of temperature, hydrogen content, loading direction and stress state, Axial Tensile, Hoop Tensile, Expansion Due to Compression and hoop Plane Strain Tensile tests are performed at room temperature, 350 C and 480 C on the material containing various hydrogen contents up to 1200 wt. ppm (hydrides are circumferential and homogeneously distributed). These tests are combined with digital image correlation and metallographic and fractographic observations at different scales. The flow stress of the material decreases with increasing temperature. The material is either strengthened or softened by hydrogen depending on temperature and hydrogen content. Plastic anisotropy depends on temperature but not on hydrogen content. The ductility of the material decreases with increasing hydrogen content at room temperature due to damage nucleation by hydride cracking. The plastic strain that leads to hydride fracture at room temperature decreases with increasing hydrogen content. The influence of stress triaxiality on hydride cracking is negligible in the studied range. The influence of hydrogen on material ductility is negligible at 350 C and 480 C since hydrides do not crack at these temperatures. The ductility of the material increases with increasing temperature. The evolution of material ductility is associated with a change in both the macroscopic fracture mode of the specimens and the microscopic failure mechanisms. (authors)

  3. Fuel cycle management

    International Nuclear Information System (INIS)

    Herbin, H.C.

    1977-01-01

    The fuel cycle management is more and more dependent on the management of the generation means among the power plants tied to the grid. This is due mainly because of the importance taken by the nuclear power plants within the power system. The main task of the fuel cycle management is to define the refuelling pattern of the new and irradiated fuel assemblies to load in the core as a function of: 1) the differences which exist between the actual conditions of the core and what was expected for the present cycle, 2) the operating constraints and the reactor availability, 3) the technical requirements in safety and the technological limits of the fuel, 4) the economics. Three levels of fuel cycle management can be considered: 1) a long term management: determination of enrichments and expected cycle lengths, 2) a mid term management whose aim corresponds to the evaluation of the batch to load within the core as a function of both: the next cycle length to achieve and the integrated power history of all the cycles up to the present one, 3) a short term management which deals with the updating of the loaded fuel utilisations to take into account the operation perturbations, or with the alteration of the loading pattern of the next batch to respect unexpected conditions. (orig.) [de

  4. 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.

  5. Ground measurements of fuel and fuel consumption from experimental and operational prescribed fires at Eglin Air Force Base, Florida

    Science.gov (United States)

    Roger D. Ottmar; Robert E. Vihnanek; Clinton S. Wright; Andrew T. Hudak

    2014-01-01

    Ground-level measurements of fuel loading, fuel consumption, and fuel moisture content were collected on nine research burns conducted at Eglin Air Force Base, Florida in November, 2012. A grass or grass-shrub fuelbed dominated eight of the research blocks; the ninth was a managed longleaf pine (Pinus palustrus) forest. Fuel loading ranged from 1.7 Mg ha-1 on a...

  6. Program on MOX fuel utilization in light water reactors

    International Nuclear Information System (INIS)

    Kenda, Hirofumi

    2000-01-01

    MOX fuel utilization program by the Japanese electric power companies was released in February, 1997. Principal philosophy for MOX fuel design is that MOX fuel shall be compatible with Uranium fuel and behavior of core loaded with MOX fuel shall be similar to that of conventional core. MOX fuel is designed so that geometry and nuclear capability of MOX fuel are equivalent to Uranium fuel. (author)

  7. Alternative Fuel Reduction Treatments in the Gunflint Corridor of the Superior National Forest: Second year results and sampling recommendations

    Science.gov (United States)

    Daniel W. Gilmore; Douglas N. Kastendick; John C. Zasada; Paula J. Anderson

    2003-01-01

    Fuel loadings need to be considered in two ways: 1) the total fuel loadings of various size classes and 2) their distribution across a site. Fuel treatments in this study affected both. We conclude that 1) mechanical treatments of machine piling and salvage logging reduced fine and heavy fuel loadings and 2) prescribed fire was successful in reducing fine fuel...

  8. Load forecasting

    International Nuclear Information System (INIS)

    Mak, H.

    1995-01-01

    Slides used in a presentation at The Power of Change Conference in Vancouver, BC in April 1995 about the changing needs for load forecasting were presented. Technological innovations and population increase were said to be the prime driving forces behind the changing needs in load forecasting. Structural changes, market place changes, electricity supply planning changes, and changes in planning objectives were other factors discussed. It was concluded that load forecasting was a form of information gathering, that provided important market intelligence

  9. Advanced fuel cycles for WWER-1000 reactors

    International Nuclear Information System (INIS)

    Semchenkov, Y. M.; Pavlovichev, A. M.; Pavlov, V. I.; Spirkin, E. I.; Styrin, Y. A.; Kosourov, E. K.

    2007-01-01

    Main stages of Russian uranium fuel development regarding improvement of safety and economics of fuel load operation are presented. Intervals of possible changes in fuel cycle duration have been demonstrated for the use of current and perspective fuel. Examples of equilibrium fuel load patterns have been demonstrated and main core neutronics parameters have been presented. Problems on the use of axial blankets with reduced enrichment in WWER-1000 fuel assemblies are considered. Some results are presented regarding core neutronic characteristics of WWER-1000 at the use of regenerated uranium and uranium-plutonium fuel. Examples of equilibrium fuel cycles for the core partially loaded with MOX fuel from weapon-grade plutonium are also considered (Authors)

  10. Enhancement of the efficiency of the automatic control system to control the thermal load of steam boilers fired with fuels of several types

    Science.gov (United States)

    Ismatkhodzhaev, S. K.; Kuzishchin, V. F.

    2017-05-01

    An automatic control system to control the thermal load (ACS) in a drum-type boiler under random fluctuations in the blast-furnace and coke-oven gas consumption rates and to control action on the natural gas consumption is considered. The system provides for use of a compensator by the basic disturbance, the blast-furnace gas consumption rate. To enhance the performance of the system, it is proposed to use more accurate mathematical second-order delay models of the channels of the object under control in combination with calculation by frequency methods of the controller parameters as well as determination of the structure and parameters of the compensator considering the statistical characteristics of the disturbances and using simulation. The statistical characteristics of the random blast-furnace gas consumption signal based on experimental data are provided. The random signal is presented in the form of the low-frequency (LF) and high-frequency (HF) components. The models of the correlation functions and spectral densities are developed. The article presents the results of calculating the optimal settings of the control loop with the controlled variable in the form of the "heat" signal with the restricted frequency variation index using three variants of the control performance criteria, viz., the linear and quadratic integral indices under step disturbance and the control error variance under random disturbance by the blastfurnace gas consumption rate. It is recommended to select a compensator designed in the form of series connection of two parts, one of which corresponds to the operator inverse to the transfer function of the PI controller, i.e., in the form of a really differentiating element. This facilitates the realization of the second part of the compensator by the invariance condition similar to transmitting the compensating signal to the object input. The results of simulation under random disturbance by the blast-furnace gas consumption are reported

  11. Impact of high soot-loaded and regenerated diesel particulate filters on the emissions of persistent organic pollutants from a diesel engine fueled with waste cooking oil-based biodiesel

    International Nuclear Information System (INIS)

    Chen, Chia-Yang; Lee, Wen-Jhy; Wang, Lin-Chi; Chang, Yu-Cheng; Yang, Hsi-Hsien; Young, Li-Hao; Lu, Jau-Huai; Tsai, Ying I.; Cheng, Man-Ting; Mwangi, John Kennedy

    2017-01-01

    Highlights: • WCO-based biodiesel blends cannot stimulate POPs formation in uncatalyzed DPF. • Formation mechanism of POPs in diesel engines is homogeneous gas-phase formation. • The gas-phase POPs are highly dominant in the raw exhausts of diesel engines. • The regeneration of the DPF can drastically reduce the formation potential of POPs in the DPFs. - Abstract: This study evaluated the impact on persistent organic pollutant (POP) emissions from a diesel engine when deploying a diesel oxidation catalyst (DOC) combined with an uncatalyzed diesel particulate filter (DPF), as well as fueling with conventional diesel (B2) and waste cooking oil-based (WCO-based) biodiesel blends (B10 and B20). When the engine was fueled with WCO-based biodiesel blends (B10 and B20) in combination with deploying DOC+A-DPF, their levels of the chlorine and potassium contents could not stimulate the formation of chlorinated POPs (PCDD/Fs and PCBs), although previous studies had warned that happened on diesel engines fueled with biodiesel and deployed with iron-catalyzed DPFs. In contrast, the WCO-based biodiesel with a lower aromatic content reduced the precursors for POP formation, and its higher oxygen content compared to diesel promoted more complete combustion, and thus using WCO-based biodiesel could reduce both PM_2_._5 and POP emissions from diesel engines. This study also evaluated the impact of DPF conditions on the POP emissions from a diesel engine; that is, the difference in POP emissions before and just after the regeneration of the DPF. In comparison to the high soot-loaded DPF scenario, the regeneration of the DPF can drastically reduce the formation potential of POPs in the DPFs. An approach was developed to correct the effects of sampling artifacts on the partitioning of gas- and particle-phase POPs in the exhaust. The gas-phase POPs are highly dominant (89.7–100%) in the raw exhausts of diesel engines, indicating that the formation mechanism of POPs in diesel

  12. Method of transporting fuel assemblies

    International Nuclear Information System (INIS)

    Okada, Katsutoshi.

    1979-01-01

    Purpose: To enable safety transportation of fuel assemblies for FBR type reactors by surrounding each of fuel elements in a wrapper tube by a rubbery, hollow cylindrical container and by sealing medium such as air to the inside of the container. Method: A fuel element is contained in a hollow cylindrical rubber-like tube. The fuel element has an upper end plug, a lower end plug and a wire spirally wound around the outer periphery. Upon transportation of the fuel assemblies, each of the fuel elements is covered with the container and arranged in the wrapper tube and then the fuel assemblies are assembled. Then, medium such as air is sealed for each of the fuel elements by way of an opening and then the opening is tightly closed. Before loading the transported fuel assemblies in the reactor, the medium is discharged through the opening and the container is completely extracted and removed from the inside of the wrapper tube. (Seki, T.)

  13. Apparatus for fuel replacement

    International Nuclear Information System (INIS)

    Imada, Takahiko.

    1974-01-01

    Object: To support a telescope mast such that no deforming load is applied to it even during massive vibration, it is held fixed at the time of fuel replacement to permit satisfactory remote control operation by automatic operation. Structure: The body of the fuel replacement apparatus is provided with telescope mast fixing means comprising a slide base supported for reciprocal movement with respect to a telescope mast, an operating arm pivoted at the slide base, a wrist member mounted on the free end of the operating arm and an engagement member for restricting the slide base and operating arm at the time of loading and unloading the fuel. When loading and unloading the fuel, the slide base and operating arm are restrained by the engagement member to reliably restrict the vibration of the telescope mast. When the fuel replacement apparatus is moved, the means provided on the operating arm is smoothly displaced to follow the swing (vibration) of the telescope mast to prevent the deforming load from being applied to the support portion or other areas. The wrist member supports the telescope mast such that it can be rotated while restraining movement in the axial direction, and it is provided with revolution drive means for rotating the telescope mast under remote control. (Kamimura, M.)

  14. Increasing TRIGA fuel lifetime with 12 wt.% U TRIGA fuel

    Energy Technology Data Exchange (ETDEWEB)

    Naughton, W F; Cenko, M J; Levine, S H; Witzig, W F [Pennsylvania State University (United States)

    1974-07-01

    In-core fuel management studies have been performed for the Penn State Breazeale Reactor (PSBR) wherein 12 wt % U fuel elements are used to replace the standard 8.5 wt % U TRIGA fuel. The core configuration used to develop a calculational model was a 90-element hexagonal array, which is representative of the PSBR core, and consists of five hexagonal rings surrounding a central thimble containing water. The technique employed for refueling the core fully loaded with 8.5 wt % U fuel involves replacing 8.5 wt % U fuel with 12 wt % U fuel using an in-out reloading scheme. A batch reload consists of 6 new 12 wt % U fuel elements. Placing the 12 wt % U fuel in the B ring produces fuel temperatures ({approx}450 {sup o}C) that are well below the 800{sup o}C maximum limitation when the PSBR is operating at its maximum allowed power of 1 Megawatt. The advantages of using new 12 wt % U fuel to replace the burned up 8.5 wt % U fuel in the B ring over refueling strictly with 8.5 wt % U-Zr TRIGA fuel are clearly delineated in Table 1 where cost calculations used the General Atomic pre-1972 prices for TRIGA fuel, i.e., $1500 and $1650 for an 8.5 and 12 wt % U fuel element, respectively. Experimental results obtained to date utilizing the 12 wt % U fuel elements agree with the computed results. (author)

  15. Fuel rod behaviour at high burnup WWER fuel cycles

    International Nuclear Information System (INIS)

    Medvedev, A.; Bogatyr, S.; Kouznetsov, V.; Khvostov, G.; Lagovsky; Korystin, L.; Poudov, V.

    2003-01-01

    The modernisation of WWER fuel cycles is carried out on the base of complete modelling and experimental justification of fuel rods up to 70 MWd/kgU. The modelling justification of the reliability of fuel rod and fuel rod with gadolinium is carried out with the use of certified START-3 code. START-3 code has a continuous experimental support. The thermophysical and strength reliability of WWER-440 fuel is justified for fuel rod and pellet burnups 65 MWd/kgU and 74 MWd/U, accordingly. Results of analysis are demonstrated by the example of uranium-gadolinium fuel assemblies of second generation under 5-year cycle with a portion of 6-year assemblies and by the example of successfully completed pilot operation of 5-year cycle fuel assemblies during 6 years at unit 3 of Kolskaja NPP. The thermophysical and strength reliability of WWER-1000 fuel is justified for a fuel rod burnup 66 MWd/kgU by the example of fuel operation under 4-year cycles and 6-year test operation of fuel assemblies at unit 1 of Kalininskaya NPP. By the example of 5-year cycle at Dukovany NPP Unit 2 it was demonstrated that WWER fuel rod of a burnup 58 MWd/kgU ensure reliable operation under load following conditions. The analysis has confirmed sufficient reserves of Russian fuel to implement program of JSC 'TVEL' in order to improve technical and economical parameters of WWER fuel cycles

  16. Nuclear fuel pellet charging device

    International Nuclear Information System (INIS)

    Komuro, Kojiro.

    1990-01-01

    The present invention concerns a nuclear fuel pellet loading device, in which nuclear fuel pellets are successively charged from an open end of a fuel can while rotating the can. That is, a fuel can sealed at one end with an end plug and opened at the other end is rotated around its pipe axis as the center on a rotationally diriving table. During rotation of the fuel can, nuclear fuel pellets are successively charged by means of a feed rod of a feeding device to the inside of the fuel can. The fuel can is rotated while being supported horizontally and the fuel pellets are charged from the open end thereof. Alternatively, the fuel can is rotated while being supported obliquely and the fuel pellets are charged gravitationally into the fuel can. In this way, the damages to the barrier of the fuel can can be reduce. Further, since the fuel pellets can be charged gravitationally by rotating the fuel can while being supported obliquely, the damages to the barrier can be reduced remarkably. (I.S.)

  17. Nuclear criticality safety studies applicable to spent fuel shipping cask designs and spent fuel storage

    International Nuclear Information System (INIS)

    Tang, J.S.

    1980-11-01

    Criticality analyses of water-moderated and reflected arrays of LWR fresh and spent fuel assemblies were carried out in this study. The calculated results indicate that using the assumption of fresh fuel loading in spent fuel shipping cask design leads to assembly spacings which are about twice the spacings of spent fuel loadings. Some shipping cask walls of composite lead and water are more effective neutron reflectors than water of 30.48 cm

  18. Beam loading

    OpenAIRE

    Boussard, Daniel

    1987-01-01

    We begin by giving a description of the radio-frequency generator-cavity-beam coupled system in terms of basic quantities. Taking beam loading and cavity detuning into account, expressions for the cavity impedance as seen by the generator and as seen by the beam are derived. Subsequently methods of beam-loading compensation by cavity detuning, radio-frequency feedback and feedforward are described. Examples of digital radio-frequency phase and amplitude control for the special case of superco...

  19. Effect of pilot fuel quantity on the performance of a dual fuel engine

    Energy Technology Data Exchange (ETDEWEB)

    Abd Alla, G.H.; Soliman, H.A.; Badr, O.A.; Abd Rabbo, M.F. [Zagazig University, Cairo (Egypt). Shoubra Faculty of Engineering

    2000-04-01

    It is well known that the operation of dual fuel engines at lower loads suffers from lower thermal efficiency and higher unburned percentages of fuel. To rectify this problem, tests have been conducted on a special single cylinder compression ignition research engine (Ricardo E6) to investigate the effect of pilot fuel quantity on the performance of an indirect injection diesel engine fuelled with gaseous fuel. Diesel fuel was used as the pilot fuel and methane or propane was used as the main fuel which was inducted into the intake manifold to mix with the intake air. Through experimental investigations, it is shown that, the low efficiency and excess emissions at light loads can be improved significantly by increasing the amount of pilot fuel, while increasing the amount of pilot fuel at high loads led to early knocking. (author)

  20. Metallic Reactor Fuel Fabrication for SFR

    Energy Technology Data Exchange (ETDEWEB)

    Song, Hoon; Kim, Jong-Hwan; Ko, Young-Mo; Woo, Yoon-Myung; Kim, Ki-Hwan; Lee, Chan-Bock [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    The metal fuel for an SFR has such advantages such as simple fabrication procedures, good neutron economy, high thermal conductivity, excellent compatibility with a Na coolant, and inherent passive safety 1. U-Zr metal fuel for SFR is now being developed by KAERI as a national R and D program of Korea. The fabrication technology of metal fuel for SFR has been under development in Korea as a national nuclear R and D program since 2007. The fabrication process for SFR fuel is composed of (1) fuel slug casting, (2) loading and fabrication of the fuel rods, and (3) fabrication of the final fuel assemblies. Fuel slug casting is the dominant source of fuel losses and recycled streams in this fabrication process. Fabrication on the rod type metallic fuel was carried out for the purpose of establishing a practical fabrication method. Rod-type fuel slugs were fabricated by injection casting. Metallic fuel slugs fabricated showed a general appearance was smooth.

  1. Progress of the DUPIC Fuel Compatibility Analysis (IV) - Fuel Performance

    International Nuclear Information System (INIS)

    Choi, Hang Bok; Ryu, Ho Jin; Roh, Gyu Hong; Jeong, Chang Joon; Park, Chang Je; Song, Kee Chan; Lee, Jung Won

    2005-10-01

    This study describes the mechanical compatibility of the direct use of spent pressurized water reactor (PWR) fuel in Canada deuterium uranium (CANDU) reactors (DUPIC) fuel, when it is loaded into a CANDU reactor. The mechanical compatibility can be assessed for the fuel management, primary heat transport system, fuel channel, and the fuel handling system in the reactor core by both the experimental and analytic methods. Because the physical dimensions of the DUPIC fuel bundle adopt the CANDU flexible (CANFLEX) fuel bundle design which has already been demonstrated for a commercial use in CANDU reactors, the experimental compatibility analyses focused on the generation of material property data and the irradiation tests of the DUPIC fuel, which are used for the computational analysis. The intermediate results of the mechanical compatibility analysis have shown that the integrity of the DUPIC fuel is mostly maintained under the high power and high burnup conditions even though some material properties like the thermal conductivity is a little lower compared to the uranium fuel. However it is required to slightly change the current DUPIC fuel design to accommodate the high internal pressure of the fuel element. It is also strongly recommended to perform more irradiation tests of the DUPIC fuel to accumulate a database for the demonstration of the DUPIC fuel performance in the CANDU reactor

  2. Progress of the DUPIC Fuel Compatibility Analysis (IV) - Fuel Performance

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hang Bok; Ryu, Ho Jin; Roh, Gyu Hong; Jeong, Chang Joon; Park, Chang Je; Song, Kee Chan; Lee, Jung Won

    2005-10-15

    This study describes the mechanical compatibility of the direct use of spent pressurized water reactor (PWR) fuel in Canada deuterium uranium (CANDU) reactors (DUPIC) fuel, when it is loaded into a CANDU reactor. The mechanical compatibility can be assessed for the fuel management, primary heat transport system, fuel channel, and the fuel handling system in the reactor core by both the experimental and analytic methods. Because the physical dimensions of the DUPIC fuel bundle adopt the CANDU flexible (CANFLEX) fuel bundle design which has already been demonstrated for a commercial use in CANDU reactors, the experimental compatibility analyses focused on the generation of material property data and the irradiation tests of the DUPIC fuel, which are used for the computational analysis. The intermediate results of the mechanical compatibility analysis have shown that the integrity of the DUPIC fuel is mostly maintained under the high power and high burnup conditions even though some material properties like the thermal conductivity is a little lower compared to the uranium fuel. However it is required to slightly change the current DUPIC fuel design to accommodate the high internal pressure of the fuel element. It is also strongly recommended to perform more irradiation tests of the DUPIC fuel to accumulate a database for the demonstration of the DUPIC fuel performance in the CANDU reactor.

  3. Nuclear fuel banks

    International Nuclear Information System (INIS)

    Anon.

    2010-01-01

    In december 2010 IAEA gave its agreement for the creation of a nuclear fuel bank. This bank will allow IAEA to help member countries that renounce to their own uranium enrichment capacities. This bank located on one or several member countries will belong to IAEA and will be managed by IAEA and its reserve of low enriched uranium will be sufficient to fabricate the fuel for the first load of a 1000 MW PWR. Fund raising has been successful and the running of the bank will have no financial impact on the regular budget of the IAEA. Russia has announced the creation of the first nuclear fuel bank. This bank will be located on the Angarsk site (Siberia) and will be managed by IAEA and will own 120 tonnes of low-enriched uranium fuel (between 2 and 4.95%), this kind of fuel is used in most Russian nuclear power plants. (A.C.)

  4. Nuclear fuel element

    International Nuclear Information System (INIS)

    Yamanaka, Tsuneyasu.

    1976-01-01

    Purpose: To provide a mechanism for the prevention of fuel pellet dislocation in fuel can throughout fuel fablication, fuel transportation and reactor operation. Constitution: A plenum spacer as a mechanism for the prevention of fuel pellet dislocation inserted into a cladding tube comprises split bodies bundled by a frame and an expansion body being capable of inserting into the central cavity of the split bodies. The expansion body is, for example, in a conical shape and the split bodies are formed so that they define in the center portion, when disposed along the inner wall of the cladding tube, a gap capable of inserting the conical body. The plenum spacer is assembled by initially inserting the split bodies in a closed state into the cladding tube after the loading of the pellets, pressing their peripheral portions and then inserting the expansion body into the space to urge the split bodies to the inner surface of the cladding tube. (Kawakami, Y.)

  5. Nuclear fuel elements

    International Nuclear Information System (INIS)

    Kawada, Toshiyuki; Hirayama, Satoshi; Yoneya, Katsutoshi.

    1980-01-01

    Purpose: To enable load-depending operation as well as moderation for the restriction of operation conditions in the present nuclear reactors, by specifying the essential ingredients and the total weight of the additives to UO 2 fuel substances. Constitution: Two or more additives selected from Al 2 O 3 , B 2 O, CaO, MgO, SiO 2 , Na 2 O and P 2 O 5 are added by the total weight of 2 - 5% to fuel substances consisting of UO 2 or a mixture of UO 2 and PuO 2 . When the mixture is sintered, the strength of the fuel elements is decreased and the fuel-cladding interactions due to the difference in the heat expansion coefficients between the ceramic fuel elements and the metal claddings are decreased to a substantially harmless degree. (Horiuchi, T.)

  6. Nuclear fuel assembly

    International Nuclear Information System (INIS)

    Domoto, Noboru; Masuda, Hiroyuki

    1989-01-01

    In a nuclear fuel assembly loaded with a plurality of fuel rods, the inside of a fuel rod disposed at a high neutron flux region is divided into an inner region and an outer region, and more burnable poisons are mixed in the inner region than in the outer region. Alternatively, the central portion of a pellet disposed in a high neutron flux region is made hollow, in which burnable poisons are charged. This can prevent neutron infinite multiplication factor from decreasing extremely at the initial burning stage. Further, the burnable poisons are not rapidly burnt completely and local peaking coefficient can be controlled. Accordingly, in a case of suppressing a predetermined excess reactivity by using a fuel rod incorporated with the burnable poison, the fuel economy can be improved more and the reactor core controllability can also be improved as compared with the usual case. (T.M.)

  7. Fuel assemblies

    International Nuclear Information System (INIS)

    Mukai, Hideyuki

    1987-01-01

    Purpose: To prevent bending of fuel rods caused by the difference of irradiation growth between coupling fuel rods and standards fuel rods thereby maintain the fuel rod integrity. Constitution: The f value for a fuel can (the ratio of pole of zirconium crystals in the entire crystals along the axial direction of the fuel can) of a coupling fuel rod secured by upper and lower tie plates is made smaller than the f value for the fuel can of a standard fuel rod not secured by the upper and the lower tie plates. This can make the irradiation growth of the fuel can of the coupling fuel rod greater than the irradiation growth of the fuel can of the standard fuel rod and, accordingly, since the elongation of the standard fuel rod can always by made greater, bending of the standard fuel rod can be prevented. (Yoshihara, M.)

  8. Fuel assembly

    International Nuclear Information System (INIS)

    Wataumi, Kazutoshi; Tajiri, Hiroshi.

    1992-01-01

    In a fuel assembly of a BWR type reactor, a pellet to be loaded comprises an external layer of fissile materials containing burnable poisons and an internal layer of fissile materials not containing burnable poison. For example, there is provided a dual type pellet comprising an external layer made of UO 2 incorporated with Gd 2 O 3 at a predetermined concentration as the burnable poisons and an internal layer made of UO 2 not containing Gd 2 O 3 . The amount of the burnable poisons required for predetermined places is controlled by the thickness of the ring of the external layer. This can dissipate an unnecessary poisoning effect at the final stage of the combustion cycle. Further, since only one or a few kinds of powder mixture of the burnable poisons and the fissile materials is necessary, production and product control can be facilitated. (I.N.)

  9. Fuel development studies

    International Nuclear Information System (INIS)

    Michel, F.

    1986-12-01

    This paper describes the main lines of the studies carried out to develop the Fast Neutron Fuel Element, from the ''SPX1-first load'' version, to progress to high performance which will be required for the project 1500 and for the fast neutron series [fr

  10. Beam loading

    CERN Document Server

    Gamp, Alexander

    2013-01-01

    We begin by giving a description of the radio-frequency generator-cavity-beam coupled system in terms of basic quantities. Taking beam loading and cavity detuning into account, expressions for the cavity impedance as seen by the generator and as seen by the beam are derived. Subsequently methods of beam-loading compensation by cavity detuning, radio-frequency feedback and feedforward are described. Examples of digital radio-frequency phase and amplitude control for the special case of superconducting cavities are also given. Finally, a dedicated phase loop for damping synchrotron oscillations is discussed.

  11. Beam loading

    International Nuclear Information System (INIS)

    Gamp, Alexander

    2013-01-01

    We begin by giving a description of the radio-frequency generator-cavity-beam coupled system in terms of basic quantities. Taking beam loading and cavity detuning into account, expressions for the cavity impedance as seen by the generator and as seen by the beam are derived. Subsequently methods of beam-loading compensation by cavity detuning, radio-frequency feedback and feedforward are described. Examples of digital radio-frequency phase and amplitude control for the special case of superconducting cavities are also given. Finally, a dedicated phase loop for damping synchrotron oscillations is discussed. (author)

  12. Fuel economy handbook

    Energy Technology Data Exchange (ETDEWEB)

    Short, W [ed.

    1979-01-01

    An overview of the UK's energy situation from 1950 to 2020 is presented. Problems are discussed and recommendations are made. A strong argument is presented for energy conservation, greater use of nuclear energy, and restrained production of North Sea oil. Specific recommendations are made for financial and operational considerations of (1) new or replacement boiler plants; (2) space heating of factories, offices and similar buildings; and (3) possible use of various fuels including duel-fuel economics and use of wastes. Tariffs and charges are discussed as well as services (e.g. compressed air, cooling water, sources of waste, etc.). Standby considerations (peak load lopping, turbines-engines, parallel or sectioned operation, etc.) and heat distribution (steam, condensate return and uses) are discussed. Throughout, the emphasis is on fuel economy. Savings in process such as recovering waste heat and the storage of heat are considered. For small industrial furnaces, intermittent heating, heat recovery, and the importance of furnace loading are discussed. (MJJ)

  13. Fuel cell research: Towards efficient energy

    CSIR Research Space (South Africa)

    Rohwer, MB

    2008-11-01

    Full Text Available fuel cells by optimising the loading of catalyst (being expensive noble metals) and ionomer; 2) Improving conventional acidic direct alcohol fuel cells by developing more efficient catalysts and by investigating other fuels than methanol; 3... these components add significantly to the overall cost of a PEMFC. 1 We focused our research activities on: 1) The effect of the loading of catalytic ink on cell performance; 2) The effect of the ionomer content in the catalytic ink; 3) Testing...

  14. My fuel treatment planner: a user guide.

    Science.gov (United States)

    Robin L. Biesecker; Roger D. Fight

    2006-01-01

    My Fuel Treatment Planner (MyFTP) is a tool for calculating and displaying the financial costs and potential revenues associated with forest fuel reduction treatments. It was designed for fuel treatment planners including those with little or no background in economics, forest management, or timber sales. This guide provides the information needed to acquire, load, and...

  15. Fuel Cell Equivalent Electric Circuit Parameter Mapping

    DEFF Research Database (Denmark)

    Jeppesen, Christian; Zhou, Fan; Andreasen, Søren Juhl

    In this work a simple model for a fuel cell is investigated for diagnostic purpose. The fuel cell is characterized, with respect to the electrical impedance of the fuel cell at non-faulty conditions and under variations in load current. Based on this the equivalent electrical circuit parameters can...

  16. The management status of the spent fuel in HANARO(1995-2009)

    International Nuclear Information System (INIS)

    Choi, Ho Young; Lim, Kyeng Hwan; Kim, Hyung Wook; Lee, Choong Sung; Ahn, Guk Hoon

    2009-11-01

    In HANARO, the spent fuels are stored in the spent fuel storage pool of the reactor hall. The capacity of the spent fuel storage pool was designed to store 600 bundles for 36 rods fuel, 432 bundles for 18 rods fuel, 315 rods for TRIGA reactor fuel and the fuels loaded in the reactor core. The spent fuel storage pool can store spent fuels discharged from the reactor core for 20 years normal operation. As for July 2009, the spent fuel 337 bundles are stored in the spent fuel storage pool. There are 217 bundles of 36 rods fuel and 120 bundles of 18 rods fuel. In this report, the information of the spent fuel about the loading date in the reactor core, discharged date, burnup, invisible inspection results and loading position in the spent fuel storage pool are described

  17. Concept and experimental studies on fuel and target for minor actinides and fission products transmutation

    Energy Technology Data Exchange (ETDEWEB)

    Prunier, C; Guerin, Y [CEA Centre d` Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. d` Etudes des Combustibles; Salvatores, M [CEA Centre d` Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Direction des Reacteurs Nucleaires; Zaetta, A [CEA Centre d` Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. d` Etudes des Reacteurs

    1994-12-31

    High activity long-lived radionuclides in nuclear wastes, namely minor actinides (americium and neptunium) are in large amount generated by current nuclear reactive. The destruction of these radionuclides is a part of the French SPIN (Partitioning and Burning) program consistent with the determination to send a minimum amount of harmful products for final storage. Transmutation concepts are defined for neptunium and americium taking into account fuel cycle strategies. Neptunium destruction does not pose any major problems. It`s a by-product of uranium consumption, as plutonium and in despite of a slight gamma activity due to the protactinium 233 it`s quite easy to handle. Diluting neptunium in the mixed oxide fuels (MOX) should not be an obstacle for fabrication, in-pile behaviour and reprocessing either. Consequently we make the proposal of homogeneous mode of neptunium in MOX which should be soon explored in the experimental OSIRIS reactor and in the Phenix and Superphenix reactors. The analysis is more complex for the multi isotope americium. Its destruction is difficult because of gamma radioactivity which complicates fabrication. Experiments in Phenix and calculation showed that Phenix reactor offers a good potential for americium incineration, but similar data do not exist for PWR. It will remain a well known difficulty for fabrication and reprocessing. In this case we have to put a real new face to the fabrication flow-sheet of americium compounds and we propose to develop the heterogeneous mode. Targets choice are defined in term of: -safety, considering fuel reaction with cladding and water sodium, -transmutation rate, limited by target behaviour, in FR`s (Phenix), PWR`s (OSIRIS) and HFR (Petten), -reprocessing, checking the solubility of such targets by Purex process. So, at the beginning of our program the account has been on improving fuel and targets properties related to safety and fuel cycle. (authors). 4 figs.

  18. 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.

  19. 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.

  20. Load sensor

    NARCIS (Netherlands)

    Van den Ende, D.; Almeida, P.M.R.; Dingemans, T.J.; Van der Zwaag, S.

    2007-01-01

    The invention relates to a load sensor comprising a polymer matrix and a piezo-ceramic material such as PZT, em not bedded in the polymer matrix, which together form a compos not ite, wherein the polymer matrix is a liquid crystalline resin, and wherein the piezo-ceramic material is a PZT powder

  1. Carbohydrate Loading.

    Science.gov (United States)

    Csernus, Marilyn

    Carbohydrate loading is a frequently used technique to improve performance by altering an athlete's diet. The objective is to increase glycogen stored in muscles for use in prolonged strenuous exercise. For two to three days, the athlete consumes a diet that is low in carbohydrates and high in fat and protein while continuing to exercise and…

  2. Low Loss Advanced Metallic Fuel Casting Evaluation

    International Nuclear Information System (INIS)

    Kim, Kihwan; Ko, Youngmo; Kim, Jonghwan; Song, Hoon; Lee Chanbock

    2014-01-01

    The fabrication process for SFR fuel is composed of fuel slug casting, loading and fabrication of the fuel rods, and the fabrication of the final fuel assemblies. Fuel slug casting is the dominant source of fuel losses and recycles streams in the fabrication process. Recycle streams include fuel slug reworks, returned scraps, and fuel casting heels, which are a special concern in the counter gravity injection casting process because of the large masses involved. Large recycle and waste streams result in lowering the productivity and the economic efficiency of fuel production. To increase efficiency the fuel losses in the furnace chamber, crucible, and the mold, after casting a considerable amount of fuel alloy in the casting furnace, will be quantitatively evaluated. After evaluation the losses will be identified and minimized. It is expected that this study will contribute to the minimization of fuel losses and the wastes streams in the fabrication process of the fuel slugs. Also through this study the technical readiness level of the metallic fuel fabrication process will be further enhanced. In this study, U-Zr alloy system fuel slugs were fabricated by a gravity casting method. Metallic fuel slugs were successfully fabricated with 19 slugs/batch with diameter of 5mm and length of 300mm. Fuel losses was quantitatively evaluated in casting process for the fuel slugs. Fuel losses of the fuel slugs were so low, 0.1∼1.0%. Injection casting experiments have been performed to reduce the fuel loss and improve the casting method. U-Zr fuel slug having φ5.4-L250mm was soundly fabricated with 0.1% in fuel loss. The fuel losses could be minimized to 0.1%, which showed that casting technology of fuel slugs can be a feasible approach to reach the goal of the fuel losses of 0.1% or less in commercial scale

  3. Alternative Fuels

    Science.gov (United States)

    Alternative fuels include gaseous fuels such as hydrogen, natural gas, and propane; alcohols such as ethanol, methanol, and butanol; vegetable and waste-derived oils; and electricity. Overview of alternative fuels is here.

  4. Advanced fuel cycles and burnup increase of WWER-440 fuel

    International Nuclear Information System (INIS)

    Proselkov, V.; Saprykin, V.; Scheglov, A.

    2003-01-01

    Analyses of operational experience of 4.4% enriched fuel in the 5-year fuel cycle at Kola NPP Unit 3 and fuel assemblies with Uranium-Gadolinium fuel at Kola NPP Unit 4 are made. The operability of WWER-440 fuel under high burnup is studied. The obtained results indicate that the fuel rods of WWER-440 assemblies intended for operation within six years of the reviewed fuel cycle totally preserve their operability. Performed analyses have demonstrated the possibility of the fuel rod operability during the fuel cycle. 12 assemblies were loaded into the reactor unit of Kola 3 in 2001. The predicted burnup in six assemblies was 59.2 MWd/kgU. Calculated values of the burnup after operation for working fuel assemblies were ∼57 MWd/kgU, for fuel rods - up to ∼61 MWd/kgU. Data on the coolant activity, specific activity of the benchmark iodine radionuclides of the reactor primary circuit, control of the integrity of fuel rods of the assemblies that were operated for six years indicate that not a single assembly has reached the criterion for the early discharge

  5. Fuel cells : a viable fossil fuel alternative

    Energy Technology Data Exchange (ETDEWEB)

    Paduada, M.

    2007-02-15

    This article presented a program initiated by Natural Resources Canada (NRCan) to develop proof-of-concept of underground mining vehicles powered by fuel cells in order to eliminate emissions. Recent studies on American and Canadian underground mines provided the basis for estimating the operational cost savings of switching from diesel to fuel cells. For the Canadian mines evaluated, the estimated ventilation system operating cost reductions ranged from 29 per cent to 75 per cent. In order to demonstrate the viability of a fuel cell-powered vehicle, NRCan has designed a modified Caterpillar R1300 loader with a 160 kW hybrid power plant in which 3 stacks of fuel cells deliver up to 90 kW continuously, and a nickel-metal hydride battery provides up to 70 kW. The battery subsystem transiently boosts output to meet peak power requirements and also accommodates regenerative braking. Traction for the loader is provided by a brushless permanent magnet traction motor. The hydraulic pump motor is capable of a 55 kW load continuously. The loader's hydraulic and traction systems are operated independently. Future fuel cell-powered vehicles designed by the program may include a locomotive and a utility vehicle. Future mines running their operations with hydrogen-fueled equipment may also gain advantages by employing fuel cells in the operation of handheld equipment such as radios, flashlights, and headlamps. However, the proton exchange membrane (PEM) fuel cells used in the project are prohibitively expensive. The catalytic content of a fuel cell can add hundreds of dollars per kW of electric output. Production of catalytic precious metals will be strongly connected to the scale of use and acceptance of fuel cells in vehicles. In addition, the efficiency of hydrogen production and delivery is significantly lower than the well-to-tank efficiency of many conventional fuels. It was concluded that an adequate hydrogen infrastructure will be required for the mining industry

  6. 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.)

  7. Load sensor

    OpenAIRE

    Van den Ende, D.; Almeida, P.M.R.; Dingemans, T.J.; Van der Zwaag, S.

    2007-01-01

    The invention relates to a load sensor comprising a polymer matrix and a piezo-ceramic material such as PZT, em not bedded in the polymer matrix, which together form a compos not ite, wherein the polymer matrix is a liquid crystalline resin, and wherein the piezo-ceramic material is a PZT powder forming 30-60% by volume of the composite, and wherein the PZT powder forms 40-50% by volume of the composite.

  8. Seismic behaviour of fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Song, Heuy Gap; Jhung, Myung Jo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1993-11-01

    A general approach for the dynamic time-history analysis of the reactor core is presented in this paper as a part of the fuel assembly qualification program. Several detailed core models are set up to reflect the placement of the fuel assemblies within the core shroud. Peak horizontal responses are obtained for each model for the motions induced from earthquake. The dynamic responses such as fuel assembly shear force, bending moment and displacement, and spacer grid impact loads are carefully investigated. Also, the sensitivity responses are obtained for the earthquake motions and the fuel assembly non-linear response characteristics are discussed. (Author) 9 refs., 24 figs., 1 tab.

  9. Method for loading resin beds

    International Nuclear Information System (INIS)

    Notz, K.J.; Rainey, R.H.; Greene, C.W.; Shockley, W.E.

    1978-01-01

    An improved method of preparing nuclear reactor fuel by carbonizing a uranium loaded cation exchange resin provided by contacting a H+ loaded resin with a uranyl nitrate solution deficient in nitrate, comprises providing the nitrate deficient solution by a method comprising the steps of reacting in a reaction zone maintained between about 145 to 200 0 C, a first aqueous component comprising a uranyl nitrate solution having a boiling point of at least 145 0 C with a second aqueous component to provide a gaseous phase containing HNO 3 and a reaction product comprising an aqueous uranyl nitrate solution deficient in nitrate

  10. Nuclear fuels

    International Nuclear Information System (INIS)

    Gangwani, Saloni; Chakrabortty, Sumita

    2011-01-01

    Nuclear fuel is a material that can be consumed to derive nuclear energy, by analogy to chemical fuel that is burned for energy. Nuclear fuels are the most dense sources of energy available. Nuclear fuel in a nuclear fuel cycle can refer to the fuel itself, or to physical objects (for example bundles composed of fuel rods) composed of the fuel material, mixed with structural, neutron moderating, or neutron reflecting materials. Long-lived radioactive waste from the back end of the fuel cycle is especially relevant when designing a complete waste management plan for SNF. When looking at long-term radioactive decay, the actinides in the SNF have a significant influence due to their characteristically long half-lives. Depending on what a nuclear reactor is fueled with, the actinide composition in the SNF will be different. The following paper will also include the uses. advancements, advantages, disadvantages, various processes and behavior of nuclear fuels

  11. 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.)

  12. Preliminary study on direct recycling of spent PWR fuel in PWR system

    International Nuclear Information System (INIS)

    Waris, Abdul; Nuha; Novitriana; Kurniadi, Rizal; Su'ud, Zaki

    2012-01-01

    Preliminary study on direct recycling of PWR spent fuel to support SUPEL (Straight Utilization of sPEnt LWR fuel in LWR system) scenario has been conducted. Several spent PWR fuel compositions in loaded PWR fuel has been evaluated to obtain the criticality of reactor. The reactor can achieve it criticality for U-235 enrichment in the loaded fresh fuel is at least 4.0 a% with the minimum fraction of the spent fuel in the core is 15.0 %. The neutron spectra become harder with the escalating of U-235 enrichment in the loaded fresh fuel as well as the amount of the spent fuel in the core.

  13. Spent fuel management

    International Nuclear Information System (INIS)

    2005-01-01

    The production of nuclear electricity results in the generation of spent fuel that requires safe, secure and efficient management. Appropriate management of the resulting spent fuel is a key issue for the steady and sustainable growth of nuclear energy. Currently about 10,000 tonnes heavy metal (HM) of spent fuel are unloaded every year from nuclear power reactors worldwide, of which 8,500 t HM need to be stored (after accounting for reprocessed fuel). This is the largest continuous source of civil radioactive material generated, and needs to be managed appropriately. Member States have referred to storage periods of 100 years and even beyond, and as storage quantities and durations extend, new challenges arise in the institutional as well as in the technical area. The IAEA gives high priority to safe and effective spent fuel management. As an example of continuing efforts, the 2003 International Conference on Storage of Spent Fuel from Power Reactors gathered 125 participants from 35 member states to exchange information on this important subject. With its large number of Member States, the IAEA is well-positioned to gather and share information useful in addressing Member State priorities. IAEA activities on this topic include plans to produce technical documents as resources for a range of priority topics: spent fuel performance assessment and research, burnup credit applications, cask maintenance, cask loading optimization, long term storage requirements including records maintenance, economics, spent fuel treatment, remote technology, and influence of fuel design on spent fuel storage. In addition to broader topics, the IAEA supports coordinated research projects and technical cooperation projects focused on specific needs

  14. Disposal of spent fuel

    International Nuclear Information System (INIS)

    Blomeke, J.O.; Ferguson, D.E.; Croff, A.G.

    1978-01-01

    Based on preliminary analyses, spent fuel assemblies are an acceptable form for waste disposal. The following studies appear necessary to bring our knowledge of spent fuel as a final disposal form to a level comparable with that of the solidified wastes from reprocessing: 1. A complete systems analysis is needed of spent fuel disposition from reactor discharge to final isolation in a repository. 2. Since it appears desirable to encase the spent fuel assembly in a metal canister, candidate materials for this container need to be studied. 3. It is highly likely that some ''filler'' material will be needed between the fuel elements and the can. 4. Leachability, stability, and waste-rock interaction studies should be carried out on the fuels. The major disadvantages of spent fuel as a disposal form are the lower maximum heat loading, 60 kW/acre versus 150 kW/acre for high-level waste from a reprocessing plant; the greater long-term potential hazard due to the larger quantities of plutonium and uranium introduced into a repository; and the possibility of criticality in case the repository is breached. The major advantages are the lower cost and increased near-term safety resulting from eliminating reprocessing and the treatment and handling of the wastes therefrom

  15. Hydrogen as automotive fuel

    International Nuclear Information System (INIS)

    Ambrosini, G.; Ciancia, A.; Pede, G.; Brighigna, M.

    1993-01-01

    Hydrogen fueled vehicles may just be the answer to the air pollution problem in highly polluted urban environments where the innovative vehicle's air pollution abatement characteristics would justify its high operating costs as compared with those of conventional automotive alternatives. This paper examines the feasibility of hydrogen as an automotive fuel by analyzing the following aspects: the chemical-physical properties of hydrogen in relation to its use in internal combustion engines; the modifications necessary to adapt internal combustion engines to hydrogen use; hydrogen fuel injection systems; current production technologies and commercialization status of hydrogen automotive fuels; energy efficiency ratings; environmental impacts; in-vehicle storage systems - involving the use of hydrides, high pressure systems and liquid hydrogen storage systems; performance in terms of pay-load ratio; autonomous operation; and operating costs. With reference to recent trial results being obtained in the USA, an assessment is also made of the feasibility of the use of methane-hydrogen mixtures as automotive fuels. The paper concludes with a review of progress being made by ENEA (the Italian Agency for New Technology, Energy and the Environment) in the development of fuel storage and electronic fuel injection systems for hydrogen powered vehicles

  16. Thermal Cycling of Uranium Dioxide - Tungsten Cermet Fuel Specimens

    Energy Technology Data Exchange (ETDEWEB)

    Gripshover, P.J.; Peterson, J.H.

    1969-12-08

    In phase I tungsten clad cermet fuel specimens were thermal cycled, to study the effects of fuel loading, fuel particle size, stablized fuel, duplex coatings, and fabrication techniques on dimensional stability during thermal cycling. In phase II the best combination of the factors studies in phase I were combined in one specimen for evaluation.

  17. Conventional fuel tank blunt impact tests : test and analysis results

    Science.gov (United States)

    2014-04-02

    The Federal Railroad Administrations Office of Research : and Development is conducting research into fuel tank : crashworthiness. A series of impact tests are planned to : measure fuel tank deformation under two types of dynamic : loading conditi...

  18. High density dispersion fuel

    International Nuclear Information System (INIS)

    Hofman, G.L.

    1996-01-01

    A fuel development campaign that results in an aluminum plate-type fuel of unlimited LEU burnup capability with an uranium loading of 9 grams per cm 3 of meat should be considered an unqualified success. The current worldwide approved and accepted highest loading is 4.8 g cm -3 with U 3 Si 2 as fuel. High-density uranium compounds offer no real density advantage over U 3 Si 2 and have less desirable fabrication and performance characteristics as well. Of the higher-density compounds, U 3 Si has approximately a 30% higher uranium density but the density of the U 6 X compounds would yield the factor 1.5 needed to achieve 9 g cm -3 uranium loading. Unfortunately, irradiation tests proved these peritectic compounds have poor swelling behavior. It is for this reason that the authors are turning to uranium alloys. The reason pure uranium was not seriously considered as a dispersion fuel is mainly due to its high rate of growth and swelling at low temperatures. This problem was solved at least for relatively low burnup application in non-dispersion fuel elements with small additions of Si, Fe, and Al. This so called adjusted uranium has nearly the same density as pure α-uranium and it seems prudent to reconsider this alloy as a dispersant. Further modifications of uranium metal to achieve higher burnup swelling stability involve stabilization of the cubic γ phase at low temperatures where normally α phase exists. Several low neutron capture cross section elements such as Zr, Nb, Ti and Mo accomplish this in various degrees. The challenge is to produce a suitable form of fuel powder and develop a plate fabrication procedure, as well as obtain high burnup capability through irradiation testing

  19. Development of MOX fuel database

    International Nuclear Information System (INIS)

    Ikusawa, Yoshihisa; Ozawa, Takayuki

    2007-03-01

    We developed MOX Fuel Database, which included valuable data from several irradiation tests in FUGEN and Halden reactor, for help of LWR MOX use. This database includes the data of fabrication and irradiation, and the results of post-irradiation examinations for seven fuel assemblies, i.e. P06, P2R, E03, E06, E07, E08 and E09, irradiated in FUGEN. The highest pellet peak burn-up reached ∼48GWd/t in MOX fuels, of which the maximum plutonium content was ∼6 wt%, irradiated in E09 fuel assembly without any failure. Also the data from the instrumented MOX fuels irradiated in HBWR to study the irradiation behavior of BWR MOX fuels under the steady state condition (IFA-514/565 and IFA-529), under the load-follow operation condition (IFA-554/555) and under the transit condition (IFA-591) are included in this database. The highest assembly burn-up reached ∼56 GWd/t in IFA-565 steady state irradiation test, and the maximum linear power of MOX fuel rods was 58.3-68.4 kW/m without any failure in IFA-591 ramp test. In addition, valuable instrument data, i.e. cladding elongation, fuel stack elongation, fuel center temperature and rod inner pressure were obtained from IFA-554/555 load-follow test. (author)

  20. LEU fuel development at CERCA

    International Nuclear Information System (INIS)

    Durand, Jean Pierre; Ottone, J.C.; Mahe, M.; Ferraz, G.

    1998-01-01

    The aim of this paper is to detail the recent progress on both U 3 Si 2 high loaded fuels and new γ phase fuels. Concerning high density density silicide plates up to 6 g Ut/cm 3 , the CEA irradiation programme is completed. Data are still under analysis but one can state that the behaviour was globally similar to conventional fuels known in SILOE and OSIRIS reactors. From the new γ fuel point of view, after demonstration feasibility in 1997 of U Mo thermally stable plates loaded up to 8.3 g Ut/cm3, CERCA has analysed the technical ability of quality inspection means assuming that is of an utmost interest for the insurance of a proper use of high performances fuel in reactors. There are mainly two differences between U Mo fuels (and more generally γ fuels) and conventional ones. Firstly, X-ray diffraction analysis on the fuel powder are needed because the chemical analysis is not sufficient to characterise the γ structure requested. Secondly, the physical limits of the Ultrasonic inspection have been reached due to transitory effect between the meat and the edges. Therefore this technic can not applied in the transitory areas. From that knowledge, the manufacture specifications for a plate dedicated to an irradiation plan can be discussed with a clearer view of the main differences with the U 3 Si 2 fuel reference. (author)

  1. Economic aspects of Dukovany NPP fuel cycle

    International Nuclear Information System (INIS)

    Vesely, P.; Borovicka, M.

    2001-01-01

    The paper discusses some aspects of high burnup program implementation at Dukovany NPP and its influence on the fuel cycle costs. Dukovany internal fuel cycle is originally designed as a three years cycle of the Out-In-In fuel reloading patterns. These reloads are not only uneconomical but they additionally increased the radiation load of the reactor pressure vessel due to high neutron leakage typical for Out-In-In loading pattern. To avoid the high neutron leakage from the core a transition to 4-year fuel cycle is started in 1987. The neutron leakage from the core is sequentially decreased by insertion of older fuel assemblies at the core periphery. Other developments in fuel cycle are: 1) increasing of enrichment in control assemblies (3.6% of U-235); 2) improvement in fuel assembly design (reduce the assembly shroud thickness from 2.1 to 1.6 mm); 3) introduction of Zr spacer grid instead of stainless steel; 4) introduction of new type of assembly with profiled enrichment with average value of 3.82%. Due to increased reactivity of the new assemblies the transition to the partial 5-year fuel cycle is required. Typical fuel loading pattern for 3, 3.5, 4 and 5-year cycles are shown in the presented paper. An evaluation of fuel cost is also discussed by using comparative analysis of different fuel cycle options. The analysis shows that introduction of the high burnup program has decrease relative fuel cycle costs

  2. Fuel consumption organization at the Kola NPP

    International Nuclear Information System (INIS)

    Matveev, A.A.; Ignatenko, E.I.; Volkov, A.P.; Trofimov, B.A.

    1981-01-01

    Problems of using NPPs in the power systems including hydroelectric power plants and NPPs are considered on the example of the Kola power system. The methods of the WWER-440 reactor fuel loading formation, reactor power forcing, optimization of volumes and time of the NPP main equipment planned maintenance are discussed. It is concluded that the optimal methods for the WWER-440 reactor fuel loading formation are the following: reactor make-up with the lesser number of fuel assemblies with maximum designed enrichment; for the case of decreased loading energy capacity displacement of make-up fuel with 2.4% enrichment by the fuel with 3.6% enrichment when preserving the designed number of make-up fuel assemblies [ru

  3. Fuel assembly

    International Nuclear Information System (INIS)

    Sakuyama, Tadashi; Mukai, Hideyuki.

    1988-01-01

    Purpose: To prevent the bending of a fuel rod caused by the difference in the elongation between a joined fuel rod and a standard fuel rod thereby maintain the fuel rod integrity. Constitution: A joined fuel rod is in a thread engagement at its lower end plug thereof with a lower plate, while passed through at its upper end plug into an upper tie plate and secured with a nut. Further, a standard fuel rod is engaged at its upper end plug and lower end plug with the upper tie plate and the lower tie plate respectively. Expansion springs are mounted to the upper end plugs of these bonded fuel rods and the standard fuel rods for preventing this lifting. Each of the fuel rods comprises a plurality of sintered pellets of nuclear fuel materials laminated in a zircaloy fuel can. The content of the alloy ingredient in the fuel can of the bonded fuel rod is made greater than that of the alloy ingredient of the standard fuel rod. this can increase the elongation for the bonded fuel rod, and the spring of the standard fuel rod is tightly bonded to prevent the bending. (Yoshino, Y.)

  4. Economic impact analysis of load forecasting

    International Nuclear Information System (INIS)

    Ranaweera, D.K.; Karady, G.G.; Farmer, R.G.

    1997-01-01

    Short term load forecasting is an essential function in electric power system operations and planning. Forecasts are needed for a variety of utility activities such as generation scheduling, scheduling of fuel purchases, maintenance scheduling and security analysis. Depending on power system characteristics, significant forecasting errors can lead to either excessively conservative scheduling or very marginal scheduling. Either can induce heavy economic penalties. This paper examines the economic impact of inaccurate load forecasts. Monte Carlo simulations were used to study the effect of different load forecasting accuracy. Investigations into the effect of improving the daily peak load forecasts, effect of different seasons of the year and effect of utilization factors are presented

  5. Assessment of fretting wear in Hanaro fuel

    International Nuclear Information System (INIS)

    Chae, Hee Taek; Lim, Kyeong Hwan; Kim, Hark Rho

    1999-06-01

    Since the first fuel loading on Feb. 1995, various zero-power tests were performed in HANARO and power ascending tests followed. After the initial fuel loading, Hanaro operation staffs inspected only two fuel bundles which were evaluated to have the highest power at the end of each cycle and they did not recognize anything peculiar in the inspected bundles. At the end of 1996, Hanaro staffs found severe wear damages in the fuel components. After that, the 4th cycle core was re-arranged with fresh fuels only to investigate wear phenomena on the fuel components. The fuel inspections have been performed 25 times periodically since the core re-configuration. In this report, fretting wear characteristics of the fuel assemblies were evaluated and summarized. Wear damages of the improved fuel assembly to resolve the wear problem were compared with those of the original fuel assembly. Based on the results of the fuel inspections, we suggest that fuel inspection need not be done for the first 60 pump operation days in order to reduce the potential of damage by a fuel handling error and an operator's burden of the fuel inspection. (author). 6 refs., 10 tabs., 5 figs

  6. Loading method of core constituting elements

    International Nuclear Information System (INIS)

    Kasai, Shigeo

    1976-01-01

    Purpose: To provide a remote-controlled replacing method for core constituting elements in a liquid-metal cooling fast breeder, wherein particularly, the core constituting elements are prevented from being loaded on the core position other than as designated. Constitution: The method comprises a first step which determines a position of a suitable neutron shielding body in order to measure a reference level of complete insertion of the core constituting elements, a second step which inserts a gripper for a fuel exchanger, a third step which decides stroke dimensions of the complete insertion, and a fourth step which discriminates the core constituting elements to begin handling of fuel rods. The method further comprises a fifth step which determines a loading position of fuel rod, and a sixth step which inserts and loads fuel rods into the core. The method still further comprises a seventh step which compares and judges the dimension of loading stroke and the dimension of complete inserting stroke so that when coincided, loading is completed, and when not coincided, loading is not completed and then the cycle of the fourth step is repeated. (Kawakami, Y.)

  7. Fuel processing

    International Nuclear Information System (INIS)

    Allardice, R.H.

    1990-01-01

    The technical and economic viability of the fast breeder reactor as an electricity generating system depends not only upon the reactor performance but also on a capability to recycle plutonium efficiently, reliably and economically through the reactor and fuel cycle facilities. Thus the fuel cycle is an integral and essential part of the system. Fuel cycle research and development has focused on demonstrating that the challenging technical requirements of processing plutonium fuel could be met and that the sometimes conflicting requirements of the fuel developer, fuel fabricator and fuel reprocessor could be reconciled. Pilot plant operation and development and design studies have established both the technical and economic feasibility of the fuel cycle but scope for further improvement exists through process intensification and flowsheet optimization. These objectives and the increasing processing demands made by the continuing improvement to fuel design and irradiation performance provide an incentive for continuing fuel cycle development work. (author)

  8. BWRs with MOx fuel

    International Nuclear Information System (INIS)

    Demaziere, C.

    1999-01-01

    Calculations has been performed for loading BWRs with pure MOx or UOx/MOx fuel. It seems to be possible to load MOx bundles in BWRs, since most of the core characteristics are comparable with the ones of a full UOx core. Nevertheless two main problems arise: The shutdown margin at BOC is lower than 1%, this requires to have a new design for the control rods in order to increase their efficiency - but the problem can also be solved by modifying the Pu quality. The cores with MOx fuel are slightly less stable, unfortunately the simple model applied does not allow giving an absolute value for the decay ratio but only allows comparing the stability with the full UOx core

  9. 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

    Fuel is one of the essential components in a reactor. It is within that fuel that nuclear reactions take place, i.e. fission of heavy atoms, uranium and plutonium. Fuel is at the core of the reactor, but equally at the core of the nuclear system as a whole. Fuel design and properties influence reactor behavior, performance, and safety. Even though it only accounts for a small part of the cost per kilowatt-hour of power provided by current nuclear power plants, good utilization of fuel is a major economic issue. Major advances have yet to be achieved, to ensure longer in-reactor dwell-time, thus enabling fuel to yield more energy; and improve ruggedness. Aside from economics, and safety, such strategic issues as use of plutonium, conservation of resources, and nuclear waste management have to be addressed, and true technological challenges arise. This Monograph surveys current knowledge regarding in-reactor behavior, operating limits, and avenues for R and D. It also provides illustrations of ongoing research work, setting out a few noteworthy results recently achieved. Content: 1 - Introduction; 2 - Water reactor fuel: What are the features of water reactor fuel? 9 (What is the purpose of a nuclear fuel?, Ceramic fuel, Fuel rods, PWR fuel assemblies, BWR fuel assemblies); Fabrication of water reactor fuels (Fabrication of UO 2 pellets, Fabrication of MOX (mixed uranium-plutonium oxide) pellets, Fabrication of claddings); In-reactor behavior of UO 2 and MOX fuels (Irradiation conditions during nominal operation, Heat generation, and removal, The processes involved at the start of irradiation, Fission gas behavior, Microstructural changes); Water reactor fuel behavior in loss of tightness conditions (Cladding, the first containment barrier, Causes of failure, Consequences of a failure); Microscopic morphology of fuel ceramic and its evolution under irradiation; Migration and localization of fission products in UOX and MOX matrices (The ceramic under irradiation

  10. 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

    Fuel is one of the essential components in a reactor. It is within that fuel that nuclear reactions take place, i.e. fission of heavy atoms, uranium and plutonium. Fuel is at the core of the reactor, but equally at the core of the nuclear system as a whole. Fuel design and properties influence reactor behavior, performance, and safety. Even though it only accounts for a small part of the cost per kilowatt-hour of power provided by current nuclear power plants, good utilization of fuel is a major economic issue. Major advances have yet to be achieved, to ensure longer in-reactor dwell-time, thus enabling fuel to yield more energy; and improve ruggedness. Aside from economics, and safety, such strategic issues as use of plutonium, conservation of resources, and nuclear waste management have to be addressed, and true technological challenges arise. This Monograph surveys current knowledge regarding in-reactor behavior, operating limits, and avenues for R and D. It also provides illustrations of ongoing research work, setting out a few noteworthy results recently achieved. Content: 1 - Introduction; 2 - Water reactor fuel: What are the features of water reactor fuel? 9 (What is the purpose of a nuclear fuel?, Ceramic fuel, Fuel rods, PWR fuel assemblies, BWR fuel assemblies); Fabrication of water reactor fuels (Fabrication of UO{sub 2} pellets, Fabrication of MOX (mixed uranium-plutonium oxide) pellets, Fabrication of claddings); In-reactor behavior of UO{sub 2} and MOX fuels (Irradiation conditions during nominal operation, Heat generation, and removal, The processes involved at the start of irradiation, Fission gas behavior, Microstructural changes); Water reactor fuel behavior in loss of tightness conditions (Cladding, the first containment barrier, Causes of failure, Consequences of a failure); Microscopic morphology of fuel ceramic and its evolution under irradiation; Migration and localization of fission products in UOX and MOX matrices (The ceramic under

  11. Containing method for spent fuel and spent fuel containing vessel

    International Nuclear Information System (INIS)

    Maekawa, Hiromichi; Hanada, Yoshine.

    1996-01-01

    Upon containing spent fuels, a metal vessel main body and a support spacer having fuel containing holes are provided. The support spacer is disposed in the inside of the metal vessel main body, and spent fuel assemblies are loaded in the fuel containing holes. Then, a lid is welded at the opening of the metal vessel main body to provide a sealing state. In this state, heat released from the spent fuel assemblies is transferred to the wall of the metal vessel main body via the support spacer. Since the support spacer has a greater heat conductivity than gases, heat of the spent fuel assemblies tends to be released to the outside, thereby capable of removing heat of the spent fuel assemblies effectively. In addition, since the surfaces of the spent fuel assemblies are in contact with the inner surface of the fuel containing holes of the support spacer, impact-resistance and earthquake-resistance are ensured, and radiation from the spent fuel assemblies is decayed by passing through the layer of the support spacer. (T.M.)

  12. Spent Nuclear Fuel (SNF) Removal Campaign Plan

    International Nuclear Information System (INIS)

    PAJUNEN, A.L.

    2000-01-01

    The overall operation of the Spent Nuclear Fuel Project will include fuel removal, sludge removal, debris removal, and deactivation transition activities. Figure 1-1 provides an overview of the current baseline operating schedule for project sub-systems, indicating that a majority of fuel removal activities are performed over an approximately three-and-one-half year time period. The purpose of this document is to describe the strategy for operating the fuel removal process systems. The campaign plan scope includes: (1) identifying a fuel selection sequence during fuel removal activities, (2) identifying MCOs that are subjected to extra testing (process validation) and monitoring, and (3) discussion of initial MCO loading and monitoring in the Canister Storage Building (CSB). The campaign plan is intended to integrate fuel selection requirements for handling special groups of fuel within the basin (e.g., single pass reactor fuel), process validation activities identified for process systems, and monitoring activities during storage

  13. Stereo photo series for quantifying natural fuels. Volume XII: Post-hurricane fuels in forests of the Southeast United States.

    Science.gov (United States)

    Robert E. Vihnanek; Cameron S. Balog; Clinton S. Wright; Roger D. Ottmar; Jeffrey W. Kelly

    2009-01-01

    Two series of single and stereo photographs display a range of natural conditions and fuel loadings in post-hurricane forests in the southeastern United States. Each group of photos includes inventory information summarizing vegetation composition, structure and loading, woody material loading and density by size class, forest floor loading, and various site...

  14. Fuel assembly

    International Nuclear Information System (INIS)

    Fujibayashi, Toru.

    1970-01-01

    Herein disclosed is a fuel assembly in which a fuel rod bundle is easily detachable by rotating a fuel rod fastener rotatably mounted to the upper surface of an upper tie-plate supporting a fuel bundle therebelow. A locking portion at the leading end of each fuel rod protrudes through the upper tie-plate and is engaged with or separated from the tie-plate by the rotation of the fastener. The removal of a desired fuel rod can therefore be remotely accomplished without the necessity of handling pawls, locking washers and nuts. (Owens, K.J.)

  15. Nuclear fuel

    International Nuclear Information System (INIS)

    D Hondt, P.

    1998-01-01

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

  16. Preliminary analysis of aircraft fuel systems for use with broadened specification jet fuels

    Science.gov (United States)

    Pasion, A. J.; Thomas, I.

    1977-01-01

    An analytical study was conducted on the use of broadened specification hydrocarbon fuels in present day aircraft. A short range Boeing 727 mission and three long range Boeing 747 missions were used as basis of calculation for one-day-per-year extreme values of fuel loading, airport ambient and altitude ambient temperatures with various seasonal and climatic conditions. Four hypothetical fuels were selected; two high-vapor-pressure fuels with 35 kPa and 70 kPa RVP and two high-freezing-point fuels with -29 C and -18 C freezing points. In-flight fuel temperatures were predicted by Boeing's aircraft fuel tank thermal analyzer computer program. Boil-off rates were calculated for the high vapor pressure fuels and heating/insulation requirements for the high freezing point fuels were established. Possible minor and major heating system modifications were investigated with respect to heat output, performance and economic penalties for the high freezing point fuels.

  17. Load testing circuit

    DEFF Research Database (Denmark)

    2009-01-01

    A load testing circuit a circuit tests the load impedance of a load connected to an amplifier. The load impedance includes a first terminal and a second terminal, the load testing circuit comprising a signal generator providing a test signal of a defined bandwidth to the first terminal of the load...

  18. Fuel management

    International Nuclear Information System (INIS)

    Schwarz, E.R.

    1975-01-01

    Description of the operation of power plants and the respective procurement of fuel to fulfil the needs of the grid. The operation of the plants shall be optimised with respect to the fuel cost. (orig./RW) [de

  19. Fuel gases

    International Nuclear Information System (INIS)

    Anon.

    1996-01-01

    This paper gives a brief presentation of the context, perspectives of production, specificities, and the conditions required for the development of NGV (Natural Gas for Vehicle) and LPG-f (Liquefied Petroleum Gas fuel) alternative fuels. After an historical presentation of 80 years of LPG evolution in vehicle fuels, a first part describes the economical and environmental advantages of gaseous alternative fuels (cleaner combustion, longer engines life, reduced noise pollution, greater natural gas reserves, lower political-economical petroleum dependence..). The second part gives a comparative cost and environmental evaluation between the available alternative fuels: bio-fuels, electric power and fuel gases, taking into account the processes and constraints involved in the production of these fuels. (J.S.)

  20. Fuel cycles

    International Nuclear Information System (INIS)

    Hawley, N.J.

    1983-05-01

    AECL publications, from the open literature, on fuels and fuel cycles used in CANDU reactors are listed in this bibliography. The accompanying index is by subject. The bibliography will be brought up to date periodically