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Sample records for booster fuel assembly

  1. Fuel Conversion Effect on Neutronics Performance of YALINA-Booster Sub-Critical Assembly from HEU to LEU

    International Nuclear Information System (INIS)

    'YALINA-Booster' is a fast-thermal sub-critical facility intended for investigating the neutronics of accelerator driven systems (ADS) at different sub-criticality levels, different configurations, and fuel compositions for the ADS development. The conversion of the YALINA-Booster assembly with highly enriched uranium (HEU) fuel in fast zone (36 and 90% of 235U) to the low enriched uranium (LEU) with 235U of less than 20% without performance losses has been performed. The experimental research program have covered the measurements of sub-criticality levels, spatial distribution of neutron flux, time dependent neutron flux measurements from different neutron source pulse durations, threshold reaction rates, transmutation reaction rates, neutron spectrum, etc. One of the important issues is the validation of the current experimental methods and techniques and their adaptation for use in ADS experiments. In this paper, the main neutronics parameters of YALINA-Booster with HEU and LEU fuels are considered. (author)

  2. Gas Test Loop Booster Fuel Hydraulic Testing

    Energy Technology Data Exchange (ETDEWEB)

    Gas Test Loop Hydraulic Testing Staff

    2006-09-01

    The Gas Test Loop (GTL) project is for the design of an adaptation to the Advanced Test Reactor (ATR) to create a fast-flux test space where fuels and materials for advanced reactor concepts can undergo irradiation testing. Incident to that design, it was found necessary to make use of special booster fuel to enhance the neutron flux in the reactor lobe in which the Gas Test Loop will be installed. Because the booster fuel is of a different composition and configuration from standard ATR fuel, it is necessary to qualify the booster fuel for use in the ATR. Part of that qualification is the determination that required thermal hydraulic criteria will be met under routine operation and under selected accident scenarios. The Hydraulic Testing task in the GTL project facilitates that determination by measuring flow coefficients (pressure drops) over various regions of the booster fuel over a range of primary coolant flow rates. A high-fidelity model of the NW lobe of the ATR with associated flow baffle, in-pile-tube, and below-core flow channels was designed, constructed and located in the Idaho State University Thermal Fluids Laboratory. A circulation loop was designed and constructed by the university to provide reactor-relevant water flow rates to the test system. Models of the four booster fuel elements required for GTL operation were fabricated from aluminum (no uranium or means of heating) and placed in the flow channel. One of these was instrumented with Pitot tubes to measure flow velocities in the channels between the three booster fuel plates and between the innermost and outermost plates and the side walls of the flow annulus. Flow coefficients in the range of 4 to 6.5 were determined from the measurements made for the upper and middle parts of the booster fuel elements. The flow coefficient for the lower end of the booster fuel and the sub-core flow channel was lower at 2.3.

  3. Fuel assembly

    International Nuclear Information System (INIS)

    Purpose: To improve the thermal and mechanical safety of fuel rods and structural components by making the local power coefficient of jointed fuel rods greater than that of other fuel rods in a fuel assembly. Constitution: In a fuel assembly comprising a plurality of fuel rods bundled by a spacer and held at the upper and the lower positions with tie plates for insertion into a channel, the degree of enrichment of uranium 235 for uranium dioxide fuel pellets charged in jointed fuel rods is adjusted such that the local power coefficient of the jointed fuel rods is made greater than that of the other fuel rods. In the case if the upper tie plate is moved upwardly by the extension of the jointed fuel rods, other fuel rods axially free from the upper tie plate receives no tension, whereby the safety of the fuel assembly can be improved. (Moriyama, K.)

  4. Fuel assembly

    International Nuclear Information System (INIS)

    A fuel assembly of a BWR type reactor comprises a rectangular parallelopiped channel box and fuel bundles contained in the channel box. The fuel bundle comprises an upper tie plate, a lower tie plate, a plurality of spacers a plurality of fuel rods and a water rod. In each fuel rod, the amount of fission products is reduced at upper and lower end regions of an effective fuel portion than that in other regions of the effective fuel region. In a portion of the fuel rods, fuel pellets containing burnable poisons are disposed at the upper and lower end regions. In addition, the upper and lower portions are constituted with natural uranium. Each of the upper and lower end regions is not greater than 15% of the effective fuel length. Since this can enhance reactivity control effect without worsening fuel economy, the control amount for excess reactivity upon long-term cycle operation can be increased. (I.N.)

  5. Fuel assembly

    International Nuclear Information System (INIS)

    A fuel assembly is composed of a fuel bundle surrounded by a channel box. The fuel bundle comprises a large number of fuel rods and a water rod secured to upper and lower tie plate by way of a plurality of fuel spacers. Grooves (libretti) are formed in the direction along the flowing direction of coolants to at least one of the surface of the fuel rods, the inner surface of the channel box, the surface of the water rod and spacer constituting components. In this case, the lateral width of the libretto in the flowing direction is determined as the minimum thickness of the bottom layer of a layered flow determined by a coolant flow rate. With such a constitution, abrasion resistance relative to coolants is reduced to reduce the pressure loss of fuel assemblies. (I.N.)

  6. Fuel gas boosters in developing nations

    Energy Technology Data Exchange (ETDEWEB)

    Sumbles, Billy [VALERUS, Ahmadabad (India). South East Asia Regional; Paris, Mike [VALERUS, Houston, TX (United States)

    2008-07-01

    Regulations requiring the use of natural gas, as opposed to flaring, and the need for clean generation solutions are driving global demand for gas-fired generation and corresponding fuel gas booster stations. Gas-producing nations are looking for viable solutions to be able to capture and use natural gas. And nations in the Middle East, Asia and Europe are developing a gas infrastructure similar to the U.S. Gas transported across long distances in pipelines often requires compression to boost pressure and meet industrial needs. In many cases, the gas may need to be processed before it can be used. However, end-users cannot just order gas booster stations out of a catalogue. Several factors - such as turbine operation, gas quality, compression pressure requirements, and the extent of processing required - need to be considered before designing and installing such systems. (author)

  7. Fuel assembly

    International Nuclear Information System (INIS)

    Purpose: To reconstruct a BWR type reactor into a high conversion reactor with no substantial changes for the reactor inner structure such as control rod structure. Constitution: The horizontal cross sectional shape of a channel box is reformed into a square configuration and the arrangement of fuel rods is formed as a trigonal lattice-like configuration. As a method of improving the conversion ratio, there is considered to use a dense lattice by narrowing the distance between fuel rods and trigonal lattice arrangement for fuel rod is advantageous therefor. A square shape cross sectional configuration having equal length both in the lateral and longitudinal directions is suitable for the channel box as a guide upon movement of the control rod. Fuel rods can be arranged with no loss by the trigonal lattice configuration, by which it is possible to improve the neutron moderation, increase the reactor core reactivity and conduct effective fuel combustion. In this way, it is possible to attain the object by inserting the follower portion of the control rod at the earier half and extracting the same at the latter half during the operation period in the reactor core comprising fuel assemblies suitable to a high conversion BWR type reactor having average conversion ratio of about 0.8. (Kamimura, M.)

  8. Fuel assembly

    International Nuclear Information System (INIS)

    The object of the present invention is to improve the hydrodynamic stability in the fuel channels of BWR type reactors and effectively utilize the coolant driving power corresponding to the reduction due to pressure loss. That is, in a fuel assembly having usual fuel rods and, in addition, water rods and short fuel rods, the structures of water rods, upper tie plates and the spacers are designed from a hydrodynamic point of view, to reduce the pressure loss. On the other hand, a lattice-like flow channel resistance member is disposed to a lower tie plate. The bundle flow rate is made uniform by the flow channel resistance member, and the pressure loss of the tie plate is increased by the reduction of the pressure loss by the arrangement of the short fuel rod and the reduction of the pressure loss described above. Since this increases the ratio of the single phase stream pressure loss in the total reactor core pressure loss, the hydrodynamic stability in the fuel channel is improved. (I.J.)

  9. Fuel assembly

    International Nuclear Information System (INIS)

    Since the neutron flux distribution and the power distribution of a fuel assembly in which short fuel rods vary greatly in the vicinity of a boundary where the distribution of uranium amount is different, the reading value of local power range monitors, having the detectors positioned in the vicinity of the boundary is varied. Then in the present invention, the upper end of the effective axial length of fuel rod is so made as not approaching with the detection position of the local power range monitor in a reactor core. Further, the upper end of the effective axial length of fuel rods in a 4 x 4 fuel rod lattice positioned at the corner on the side of the local power range monitor is so made as not approaching the detection position of the local power range monitor. As a result, the change of the neutron flux distribution and power distribution in the vicinity of the position where the detector of the local power range monitor is situated can be extremely reduced. Accordingly, there is no scattering and fluctuation for the reading value by the local power range monitor, to improve the monitoring performance for thermal characteristics in the reactor core. (N.H.)

  10. Fuel assembly

    International Nuclear Information System (INIS)

    The cross section of a fuel assembly is divided to a first region containing corner portions at which channel fasteners are situated and a second region not containing corner portions. The average enrichment degree of plutonium in the first region is decreased than that of the second region, and the number of fuel rods containing burnable poisons is increased at the first region than that of the second region. In the first region of the fuel assembly, the effect of moderating neutrons is enhanced since the cross section of a moderator flow channel at the outer side of the channel box is large. Therefore, local power peaking is increased in the first region while it is decreased in the second region that opposes to a narrow gap. The average enrichment degree of plutonium in the first region is decreased and that in the second region is increased by so much, to flatten the power distribution. Then, the reduction of the reactivity worth of gadolinia, as burnable poisons, can be suppressed. (N.H.)

  11. Fuel assembly

    International Nuclear Information System (INIS)

    The present invention concerns a fuel assembly of a BWR type reactor, and prevents aging change of flow rate of coolants leaked from a gap between a lower tie plate and a channel box. That is, in the fuel assembly, a great number of fuel rods and a plurality of water rods are bundled by a plurality of spacers, the upper and the lower ends thereof are supported by upper and lower tie plates, and they are contained in a channel box. Plate-like protrusions are disposed rotatably to the lower tie plate at a position corresponding to the lower end of the channel box. In addition, through holes are disposed on the side wall of the lower tie plate. With such a constitution, the protrusions rotate at a connection portion by hydraulic pressure of leaking coolants, and urge the channel box by the other end to control leakage of coolants. Further, since the through holes are disposed on the side wall of the lower tie plate, pressure difference is caused between the upper and the lower surfaces of the plate of the protrusion, to rotate the protrusions at the connection portion, and the other end of the protrusions presses the channel box to obtain the same effect. (I.S.)

  12. Fuel assembly

    International Nuclear Information System (INIS)

    Fuel rods are arranged in a lattice-like structure by way of a plurality of spacers and the lower ends thereof are fixed to a lower tie plate for assembling a fuel rod bundle. The outer circumference is surrounded by a basket having a plurality of openings and the basket is surrounded by a channel box. The basket is connected to a handle at the upper end and to a lower tie plate at the lower end and, further, defined with a scraper at each of openings. Coolants flown from the lower tie plate to the channel box flow the channels between the channel box and the basket and a fuel rod bundle, uprise while forming a two-phase flow and flow out from the upper end of the channel box. Since no upper tie plate is present, pressure loss of coolants flow is reduced, and liquid membranes of coolants are peeled off by the scraper disposed at the opening of the basket, which contributes to the improvement of the limit power. In addition, fuel rods are inspected and cleaned easily. (N.H.)

  13. Fuel assembly

    International Nuclear Information System (INIS)

    Object: To divide fuel rods into several blocks so that fuels may be reversed vertically every block to leave sufficient allowance for reactor stoppage, thus enhancing taking-out combustion quality. Structure: A fuel inserting portion in upper and lower tie plates is designed so that a vertically symmetrical fuel may be inserted. That is, the construction of the fuel rod itself is entirely vertically symmetrical. Fuel regions are symmetrically arranged on uppper and lower ends, and expansion springs are also inserted at upper and lower parts. Outer springs of the fuel rods are always retained at plug portions on upper and lower ends. The fuel rods are of the sub-channel construction consisting of several rods, the fuel rods being separable from one another every sub-channel. Accordingly, the fuel may be reversed every sub-channel. (Kamimura, M.)

  14. Fuel assembly

    International Nuclear Information System (INIS)

    Fuel rods enriched with plutonium and fuel rods formed by incorporating combustible poisons in enriched uranium are arranged in square lattice like structure. MOX fuel pellets comprise PuO2 as a fuel material and contain 239Pu, 241Pu as fission products. The gadolinia-incorporated uranium fuel pellets comprise UO2 as a fuel material and gadolinia as a burnable poison incorporated therein and contains 235U as a fuel material. The axial distribution of the concentration of gadolinia contained in the uranium fuel rods is axially divided into three regions in a region less than 1/2 of a fuel effective length, and the concentration of gadolinia is highest at the lowest region, and the concentration of gadolinia is made lower toward the upper regions. With such a constitution, the degree of downward distortion of the axial power distribution is suppressed in a reactor core of a BWR type reactor having a large MOX loading rate. (I.N.)

  15. Fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Izutsu, Sadayuki; Fujita, Satoshi [Hitachi Engineering Co. Ltd., Ibaraki (Japan); Fujimaki, Shingo; Sasagawa, Masaru; Kaneto, Kunikazu; Mochida, Takaaki; Aoyama, Motoo; Shimada, Hidemitsu

    1997-09-09

    Fuel rods enriched with plutonium and fuel rods formed by incorporating combustible poisons in enriched uranium are arranged in square lattice like structure. MOX fuel pellets comprise PuO{sub 2} as a fuel material and contain {sup 239}Pu, {sup 241}Pu as fission products. The gadolinia-incorporated uranium fuel pellets comprise UO{sub 2} as a fuel material and gadolinia as a burnable poison incorporated therein and contains {sup 235}U as a fuel material. The axial distribution of the concentration of gadolinia contained in the uranium fuel rods is axially divided into three regions in a region less than 1/2 of a fuel effective length, and the concentration of gadolinia is highest at the lowest region, and the concentration of gadolinia is made lower toward the upper regions. With such a constitution, the degree of downward distortion of the axial power distribution is suppressed in a reactor core of a BWR type reactor having a large MOX loading rate. (I.N.)

  16. YALINA-booster subcritical assembly pulsed-neutron experiments : data processing and spatial corrections.

    Energy Technology Data Exchange (ETDEWEB)

    Cao, Y.; Gohar, Y.; Nuclear Engineering Division

    2010-10-11

    The YALINA-Booster experiments and analyses are part of the collaboration between Argonne National Laboratory of USA and the Joint Institute for Power & Nuclear Research - SOSNY of Belarus for studying the physics of accelerator driven systems for nuclear energy applications using low enriched uranium. The YALINA-Booster subcritical assembly is utilized for studying the kinetics of accelerator driven systems with its highly intensive D-T or D-D pulsed neutron source. In particular, the pulsed neutron methods are used to determine the reactivity of the subcritical system. This report examines the pulsed-neutron experiments performed in the YALINA-Booster facility with different configurations for the subcritical assembly. The 1141 configuration with 90% U-235 fuel and the 1185 configuration with 36% or 21% U-235 fuel are examined. The Sjoestrand area-ratio method is utilized to determine the reactivities of the different configurations. The linear regression method is applied to obtain the prompt neutron decay constants from the pulsed-neutron experimental data. The reactivity values obtained from the experimental data are shown to be dependent on the detector locations inside the subcritical assembly and the types of detector used for the measurements. In this report, Bell's spatial correction factors are calculated based on a Monte Carlo model to remove the detector dependences. The large differences between the reactivity values given by the detectors in the fast neutron zone of the YALINA-Booster are reduced after applying the spatial corrections. In addition, the estimated reactivity values after the spatial corrections are much less spatially dependent.

  17. Nuclear reactor fuel assembly

    International Nuclear Information System (INIS)

    A fuel assembly construction for liquid metal cooled fast breeder reactors is described in which the sub-assemblies carry a smaller proportion of parasitic material than do conventional sub-assemblies. (U.K.)

  18. Fuel Assembly Damping Summary

    International Nuclear Information System (INIS)

    This paper summary the fuel assembly damping data in air/in still water/under flow, released from foreign fuel vendors, compared our data with the published data. Some technical issues in fuel assembly damping measurement testing are also briefly discussed. Understanding of each fuel assembly damping mechanisms according to the surrounding medium and flow velocity can support the fuel design improvement in fuel assembly dynamics and structural integrity aspect. Because the upgraded requirements of the newly-developed advanced reactor system will demands to minimize fuel design margin in integrity evaluation, reduction in conservatism of fuel assembly damping can contribute to alleviate the fuel design margin for sure. Damping is an energy dissipation mechanism in a vibrating mechanical structure and prevents a resonant structure from having infinite vibration amplitudes. The sources of fuel assembly damping are various from support friction to flow contribution, and it can be increased by the viscosity or drag of surrounding fluid medium or the average velocity of water flowing. Fuel licensing requires fuel design evaluation in transient or accidental condition. Dynamic response analysis of fuel assembly is to show fuel integrity and requires information on assembly-wise damping in dry condition and under wet or water flowing condition. However, damping measurement test for the full-scale fuel assembly prototype is not easy to carry out because of the scale (fuel prototype, test facility), unsteadiness of test data (scattering, random sampling and processing), instrumentation under water flowing (water-proof response measurement), and noise. LWR fuel technology division in KAERI is preparing the infra structure for damping measurement test of full-scale fuel assembly, to support fuel industries and related research activities. Here is a preliminary summary of fuel assembly damping, published in the literature. Some technical issues in fuel assembly damping

  19. Reactor fuel assemblies

    International Nuclear Information System (INIS)

    A description is given of an improved spacer grid for a nuclear fuel assembly comprising fuel rods in a matrix wherein each rod is adapted to be enclosed by a spacer ''cell'' for positioning thereof relative to adjacent rods in the fuel assembly. 7 claims, 12 drawing figures

  20. Nuclear fuel assembly

    International Nuclear Information System (INIS)

    A nuclear fuel assembly includes and upper yoke, a base, an elongated, outer flow channel disposed substantially along the entire length of the fuel assembly and an elongated, internal, central water cross, formed by four, elongated metal angles, that divides the nuclear fuel assembly into four, separate, elongated fuel sections and that provides a centrally disposed path for the flow of subcooled neutron moderator along the length of the fuel assembly. A separate fuel bundle is located in each of the four fuel sections and includes an upper tie plate, a lower tie plate and a plurality of elongated fuel rods disposed therebetween. Preferably, each upper tie plate is formed from a plurality of interconnected thin metal bars and includes an elongated, axially extending pin that is received by the upper yoke of the fuel assembly for restraining lateral motion of the fuel bundle while permitting axial movement of the fuel bundle with respect to the outer flow channel. The outer flow channel is fixedly secured at its opposite longitudinal ends to the upper yoke and to the base to permit the fuel assembly to be lifted and handled in a vertical position without placing lifting loads or stresses on the fuel rods. The yoke, removably attached at the upper end of the fuel assembly to four structural ribs secured to the inner walls of the outer flow channel, includes, as integrally formed components, a lifting bail or handle, laterally extending bumpers, a mounting post for a spring assembly, four elongated apertures for receiving with a slip fit the axially extending pins mounted on the upper tie plates and slots for receiving the structural ribs secured to the outer flow channel. Locking pins securely attach the yoke to the structural ribs enabling the fuel assembly to be lifted as an entity

  1. Spent fuel assembly hardware

    International Nuclear Information System (INIS)

    When spent nuclear fuel is disposed of in a repository, the waste package will include the spent fuel assembly hardware, the structural portion of the fuel assembly, and the fuel pins. The spent fuel assembly hardware is the subject of this paper. The basic constituent parts of the fuel assembly will be described with particular attention on the materials used in their construction. The results of laboratory analyses performed to determine radionuclide inventories and trace impurities also will be described. Much of this work has been incorporated into a US Department of Energy (DOE) database maintained by Oak Ridge National Laboratory (ORNL). This database is documented in DOE/RW-0184 and can be obtained from Karl Notz at ORNL. The database provides a single source for information regarding wastes that may be sent to the repository

  2. Yalina-Booster Assembly: from HEU to LEU

    International Nuclear Information System (INIS)

    The YALINA facility is a unique facility which was designed as a zero power model of real ADS (Accelerator Driven System). It is intended to study ADS neutronics and kinetics of the subcritical reactors driven by external neutron sources. Accelerator-driven systems may play an important role in future nuclear fuel cycles to reduce the long-term radiotoxicity and volume of spent nuclear fuel. Successful operation of this facility is a scientific contribution from the Republic of Belarus, as well as the international community. The experimental data are used to benchmark and validate methods and computer codes for designing and licensing ADS. In this paper the investigation of spatial kinetics of the sub-critical systems with external neutron sources, validation of the experimental techniques for sub-criticality monitoring and estimation of probability of minor actinides and fission products transmutation is made for different configurations of Yalina-Booster during conversion from HEU to LEU: - 1st configuration – with HEU fuel in the fast zone (metallic uranium of 90% enrichment by 235U and UO2 of 36% enrichment by 235U) and uranium dioxide of 10% enrichment by 235U in thermal zone; - 2nd one – with 36% UO2 in fast zone and 10% in thermal zone; - 3rd one – with 21% UO2 in fast zone and 10% in thermal zone; - 4th one – with 21% UO2 in fast zone and 10% in thermal zone, differing from the 3rd configuration by rounded shape of fast zone and annular shape of the absorber zone with permanent number of the absorbing rods. (author)

  3. Fuel assembly reconstitution

    International Nuclear Information System (INIS)

    Fuel failures have been happened in Nuclear Power Plants worldwide, without lost of integrity and safety, mainly for the public, environment and power plants workers. The most common causes of these events are corrosion (CRUD), fretting and pellet cladding interaction. These failures are identified by increasing the activity of fission products, verified by chemical analyses of reactor coolant. Through these analyses, during the fourth operation cycle of Angra 2 Nuclear Power Plant, was possible to observe fuel failure indication. This indication was confirmed in the end of the cycle during the unloading of reactor core through leakage tests of fuel assembly, using the equipment called 'In Mast Sipping' and 'Box Sipping'. After confirmed, the fuel assembly reconstitution was scheduled, and happened in April, 2007, where was identified the cause and the fuel rod failure, which was substitute by dummy rods (zircaloy). The cause was fretting by 'debris'. The actions to avoid and prevent fuel assemblies failures are important. The goals of this work are to describe the methodology of fuel assembly reconstitution using the FARE (Fuel Assembly Reconstitution Equipment) system, to describe the results of this task in economic and security factors of the company and show how the fuel assembly failures are identified during operation and during the outage. (author)

  4. Nuclear fuel assembly spacer

    International Nuclear Information System (INIS)

    In a fuel assembly for a nuclear reactor a fuel element spacer formed of an array of laterally positioned cojoined tubular ferrules each providing a passage for one of the fuel elements, the elements being laterally supported in the ferrules between slender spring members and laterally oriented rigid stops

  5. Fuel nozzle assembly

    Science.gov (United States)

    Johnson, Thomas Edward; Ziminsky, Willy Steve; Lacey, Benjamin Paul; York, William David; Stevenson, Christian Xavier

    2011-08-30

    A fuel nozzle assembly is provided. The assembly includes an outer nozzle body having a first end and a second end and at least one inner nozzle tube having a first end and a second end. One of the nozzle body or nozzle tube includes a fuel plenum and a fuel passage extending therefrom, while the other of the nozzle body or nozzle tube includes a fuel injection hole slidably aligned with the fuel passage to form a fuel flow path therebetween at an interface between the body and the tube. The nozzle body and the nozzle tube are fixed against relative movement at the first ends of the nozzle body and nozzle tube, enabling the fuel flow path to close at the interface due to thermal growth after a flame enters the nozzle tube.

  6. Nuclear fuel assembly

    International Nuclear Information System (INIS)

    A nuclear fuel assembly comprises a cluster of elongated fuel, retained parallel and at the nodal points of a square network by a bottom supporting plate and by spacing grids. The supporting plate is connected to a top end plate via tie-rods which replace fuel pins at certain of the nodal points of the network. The diameter of the tie-rods is equal to that of the pins and both are slidably received in the grids

  7. Nuclear fuel assembly

    International Nuclear Information System (INIS)

    A fuel assembly of PWR comprises a fuel bundle portion supported by a plurality of support lattices and an upper and lower nozzles each secured to the upper and lower portions. Leaf springs are attached to the four sides of the upper nozzle for preventing rising of the fuel assembly by streams of cooling water by the contact with an upper reactor core plate. The leaf springs are attached to the upper nozzle so that four leaf springs are laminated. The uppermost leaf spring is bent slightly upwardly from the mounted portion and the other leaf springs are extended linearly from the mounted portion without being bent. The mounted portions of the leaf springs are stacked and secured to the upper nozzle by a bolt obliquely relative to the axial line of the fuel assembly. (I.N.)

  8. MOX fuel assembly

    International Nuclear Information System (INIS)

    The fuel assembly of the present invention comprises at least one water rod, first fuel rods filled with uranium/plutonium mixed oxide fuels, second fuel rods having axial length shorter than that of the first fuel rods and third fuel rods containing burnable poisons. If the third fuel rods are arranged on the same row and adjacent columns or on the same column and adjacent row relative to the positions where the second fuel rods are arranged or the position of the water rod replacing fuel rods, in other words, at a position extremely close to them, neutron spectrum is made softer and the neutron flux distribution is made higher. As a result, negative reactivity worth of the burnable poisons contained in the third fuel rods is enhanced, accordingly, a reactivity suppression effect comparable with that in conventional cases can be obtained by so much even if the number of the third fuel rods is reduced. The number of the MOX fuel rods is increased than a conventional case by so much as replacing the third fuel rods with the MOX fuel rods by the reduced amount thereby enabling to improve the efficiency using plutonium. (N.H.)

  9. Nuclear fuel assembly

    International Nuclear Information System (INIS)

    Purpose: To increase the fuel assembly rigidity while making balance in view of the dimension thereby improving the earthquake proofness. Constitution: In a nuclear fuel assembly having a control rod guide thimble tube, the gap between the thimble tube and fuel insert (inner diameter of the guiding thimble tube-outer diameter of the fuel insert) is made greater than 1.0 mm. Further, the wall thickness of the thimble tube is made to about 4 - 5 % of the outer diameter, while the flowing fluid pore cross section S in the thimble tube is set as: S = S0 x A0/A where S0: cross section of the present flowing fluid pore, A: effective cross section after improvement, = Π/4(d2 - D2) in which d is the thimble tube inner diameter and the D is the fuel insert outer diameter. A0: present effective cross section. (Seki, T.)

  10. Nuclear fuel assembly

    International Nuclear Information System (INIS)

    Purpose: To obtain a nuclear fuel assembly having a function of eliminating corrosion products exfoliating from the surface of a fuel can, thereby reduce the radioactive crud in primary sodium coolant during operation of a FBR type reactor. Constitution: Nickel plates or grids made of metal plate with a nickel coated on the surface thereof are inserted in the upper blanket of a nuclear fuel element and between nuclear fuel element corresponding to the gas plenum. The nickel becomes helpful at high temperature in adsorbing Mn-54 which accounts for a major portion of the corrosion products. (J.P.N.)

  11. Transfer of fuel assemblies

    International Nuclear Information System (INIS)

    Fuel assemblies of a nuclear reactor are transferred during fueling or refueling or the like by a crane. The work-engaging fixture of the crane picks up an assembly, removes it from this slot, transfers it to the deposit site and deposits it in its slot at the deposit site. The control for the crane includes a strain gauge connected to the crane line which raises and lowers the load. The strain gauge senses the load on the crane. The signal from the strain gauge is compared with setpoints; a high-level setpoint, a low-level setpoint and a slack-line setpoint. If the strain gauge signal exceeds the high-level setpoint, the line drive is disabled. This event may occur during raising of a fuel assembly which encounters resistance. The high-level setpoint may be overridden under proper precautions. The line drive is also disabled if the strain gauge signal is less than the low-level setpoint. This event occurs when a fuel assembly being deposited contacts the bottom of its slot or an obstruction in, or at the entry to the slot. To preclude lateral movement and possible damage to a fuel assembly suspended from the crane line, the traverse drive of the crane is disabled once the strain-gauge exceets the lov-level setpoint. The traverse drive can only be enabled after the strain-gauge signal is less than the slack-line set-point. This occurs when the lines has been set in slack-line setting. When the line is tensioned after slack-li ne setting, the traverse drive remains enabled only if the line has been disconnected from the fuel assembly

  12. Method of loading fuel assembly

    International Nuclear Information System (INIS)

    Purpose: To shorten the fuel assembly loading time by loading fuel assembly group as one body into the reactor core. Method: A fuel assembly is fed from an auxiliary reactor building via a pit crane into the reactor container, and is stood from lateral position to vertical position. Further, the fuel assemblies are moved laterallyiin a pool of the container, and every four assembly groups are formed by an aligning jig. These assembly groups are associated into one body and loaded into the container. Thus, the round trip time of the crane in the fuel assembly loading work can be shortened. (Yoshihara, H.)

  13. Fuel assembly supporting structure

    International Nuclear Information System (INIS)

    For use in forming the core of a pressurized-water reactor, a fuel assembly supporting structure for holding a bundle of interspaced fuel rods, is formed by interspaced end pieces having holes in which the end portions of control rod guide tubes are inserted, fuel rod spacer grids being positioned by these guide tubes between the end pieces. The end pieces are fastened to the end portions of the guide tubes, to integrate the supporting structure, and in the case of at least one of the end pieces, this is done by means which releases that end piece from the guide tubes when the end pieces receive an abnormal thrust force directed towards each other and which would otherwise place the guide tubes under a compressive stress that would cause them to buckle. The spacer grids normally hold the fuel rods interspaced by distances determined by nuclear physics, and buckling of the control rod guide tubes can distort the fuel rod spacer grids with consequent dearrangement of the fuel rod interspacing. A sudden loss of pressure in a pressurized-water reactor pressure vessel can result in the pressurized coolant in the vessel discharging from the vessel at such high velocity as to result in the abnormal thrust force on the end pieces of each fuel assembly, which could cause buckling of the control rod guide tubes when the end pieces are fixed to them in the normal rigid and unyielding manner

  14. Fuel assemblies chemical cleaning

    International Nuclear Information System (INIS)

    NPP Paks found a thermal-hydraulic anomaly in the reactor core during cycle 14 that was caused by corrosion product deposits on fuel assemblies (FAs) that increased the hydraulic resistance of the FAs. Consequently, the coolant flow through the FAs was insufficient resulting in a temperature asymmetry inside the reactor core. Based on this fact NPP Paks performed differential pressure measurements of all fuel assemblies in order to determine the hydraulic resistance and subsequently the limit values for the hydraulic acceptance of FAs to be used. Based on the hydraulic investigations a total number of 170 FAs was selected for cleaning. The necessity for cleaning the FAs was explained by the fact that the FAs were subjected to a short term usage in the reactor core only maximum of 1,5 years and had still a capacity for additional 2 fuel cycles. (authors)

  15. Method of assembling nuclear fuel assembly

    International Nuclear Information System (INIS)

    Thin films are formed to the surface of a fuel rod for preventing the occurrence of injuries at the surface of the fuel rod. That is, in a method of assembling a nuclear fuel assembly by inserting fuel rods into lattice cells of a support lattice, thin films of polyvinyl alcohol are formed to a predetermined thickness at the surface of each of the fuel rods and, after insertion of the fuel rods into the lattice cells, the nuclear fuel assemblies are dipped into water or steams to dissolve and remove the thin films. Since polyvinyl alcohol is noncombustible and not containing nuclear inhibitive material as the ingredient, they cause no undesired effects on plant facilities even if not completely removed from the fuel rods. The polyvinyl alcohol thin films have high strength and can sufficiently protect the fuel rod. Further, scraping damages caused by support members of the support lattice upon insertion can also be prevented. (T.M.)

  16. Nuclear fuel assembly spacer

    International Nuclear Information System (INIS)

    A spacer for use in a fuel assembly of a nuclear reactor having thin, full-height divider members, slender spring members and laterally oriented rigid stops and wherein the total amount of spacer material, the amount of high neutron cross section material, the projected area of the spacer structure and changes in cross section area of the spacer structure are minimized whereby neutron absorption by the spacer and coolant flow resistance through the spacer are minimized

  17. Nuclear reactor fuel assembly

    International Nuclear Information System (INIS)

    The description is given of a nuclear reactor fuel assembly comprising fuel elements arranged in a supporting frame composed of two end pieces, one at the top and the other at the bottom, on which are secured the ends of a number of vertical tubes, each end piece comprising a plane bottom on which two series of holes are made for holding the tubes and for the passage of the coolant. According to the invention, the bottom of each end piece is fixed to an internal plate fitted with the same series of holes for holding the tubes and for the fluid to pass through. These holes are of oblong section and are fitted with fixing elements cooperating with corresponding elements for securing these tubes by transversal movement of the inside plate

  18. Seismic behaviour of fuel assembly

    International Nuclear Information System (INIS)

    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

  19. 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.

  20. Nuclear fuel assembly

    International Nuclear Information System (INIS)

    In a nuclear fuel assembly comprising a nuclear fuel bundle in which a plurality of nuclear rods are bond by an upper tie plate, spacers and lower tie plate and a channel box containing them, the inner surface of the channel box and the surface of the lower tie plate opposing thereto are fabricated into step-like configuration respectively and the two fabricated surfaces are opposed to each other to constitute a step-like labyrinth flow channel. With such a configuration, when a fluid flows from higher pressure to lower pressure side, pressure loss is caused due to fluid friction in proportion with the length of the flow channel, due to the change of the flowing direction and, further, in accordance with deceleration or acceleration at each of the stepped portions. The total for each of the pressure loses constitutes the total pressure loss in the labyrinth. That is, if the pressure difference between the inside and the outside of the channel box is identical, the amount of leakage is reduced by so much as the increase of the total pressure loss, to thereby improve the stability of the reactor core and fuel economy. (T.M.)

  1. Fuel assembly supports

    International Nuclear Information System (INIS)

    Purpose: To prevent fuel assembly from lifting by forming through holes in the entrance nozzle and the connection pipe respectively opposed to each other and forming an expanded portion and inserting therein a stopper member at the position where the two holes are joined. Constitution: A through hole is formed in a connection tube slanted upwardly and inwardly from a high pressure plenum to the inside of the connection tube. While on the other hand, another through hole slanted with same angle is also formed to the reduced diameter portion of an entrance nozzle at the position corresponding to the above hole in the connection tube. Further, an expanded diameter section is formed to the inside of the connection tube and the outside of the reduced diameter section of the entrance nozzle, and a steel ball is mounted therein. (Kawakami, Y.)

  2. YALINA Booster conversion project

    International Nuclear Information System (INIS)

    The YALINA Booster subcritical assembly was constructed at the Joint Institute for Power and Nuclear Research (JIPNR)-SOSNY, Belarus to examine the physics of Accelerator Driven Systems (ADS). The assembly has fast and thermal zones to study the coupling between the two zones, the transuranics transmutation, and the ADS kinetics. It is driven by external neutron source located at the assembly center. The central fast zone (the booster zone) consists of high enriched uranium (HEU) fuel rods loaded in a lead matrix and it is surrounded by thermal zone. The thermal zone has low enriched uranium (LEU) fuel rods loaded in polyethylene moderator. Between the two zones, there is a thermal neutron absorber zone. (JIPNR)-SOSNY has an International Science and Technology Center project in collaboration with Argonne National Laboratory of USA to convert the HEU fuel of YALINA Booster to LEU fuel without penalizing its performance. The first step of this project is to characterize and define the performance of the YALINA Booster subcritical assembly with HEU fuel by performing detailed analytical and experimental studies. The second step is to convert the booster zone to use uranium fuel rods with 21% enrichment. The YALINA Booster configuration is modified to reach the original subcriticality level. The analytical analyses have developed accurate calculational models without geometrical approximations for performing Monte Carlo and Deterministic calculations. MCNP, MCNPX, MCB, MONK, ERANOS, and PARTISN computer codes with different nuclear data libraries based on ENDF/VI, JEF2.2, and JEF3.1 have been used for static and kinetic analyses. The geometrical details are included explicitly without approximation or homogenization. In the experimental program, the subcriticality has been measured as a function of the number of the fuel rods loaded in the subcritical assembly. Different methods have been used to measure the assembly subcriticality during the fuel loading process. In

  3. Swivel base for fuel assembly storage

    International Nuclear Information System (INIS)

    An invention is described the principal object of which is to provide a nuclear fuel assembly storage rack capable of supporting spent fuel assemblies without generating stresses in the fuel assemblies. The storage rack consists of a lower and upper support for supporting and retaining the spent fuel assemblies in their vertical positions. Relief from any stresses in the fuel assembly during storage is obtained by the provision of a swivel base in the lower support. (U.K.)

  4. BN-600 fuel elements and fuel assemblies operating experience

    International Nuclear Information System (INIS)

    Consideration is given to the data on fuel burnup of standard fuel assemblies of the BN-600 reactor first core charge and that for modified core; data on operation ability of fuel assemblies of the first charge type are given. Data on main results of primary post-irradiation examination of fuel assemblies and fuel elements and maximal values of fuel burnup, achieved in particular fuel assemblies of BN-600 reactor are presented. 4 figs.; 1 tab

  5. FUEL ASSEMBLY SHAKER TEST SIMULATION

    Energy Technology Data Exchange (ETDEWEB)

    Klymyshyn, Nicholas A.; Sanborn, Scott E.; Adkins, Harold E.; Hanson, Brady D.

    2013-05-30

    This report describes the modeling of a PWR fuel assembly under dynamic shock loading in support of the Sandia National Laboratories (SNL) shaker test campaign. The focus of the test campaign is on evaluating the response of used fuel to shock and vibration loads that a can occur during highway transport. Modeling began in 2012 using an LS-DYNA fuel assembly model that was first created for modeling impact scenarios. SNL’s proposed test scenario was simulated through analysis and the calculated results helped guide the instrumentation and other aspects of the testing. During FY 2013, the fuel assembly model was refined to better represent the test surrogate. Analysis of the proposed loads suggested the frequency band needed to be lowered to attempt to excite the lower natural frequencies of the fuel assembly. Despite SNL’s expansion of lower frequency components in their five shock realizations, pretest predictions suggested a very mild dynamic response to the test loading. After testing was completed, one specific shock case was modeled, using recorded accelerometer data to excite the model. Direct comparison of predicted strain in the cladding was made to the recorded strain gauge data. The magnitude of both sets of strain (calculated and recorded) are very low, compared to the expected yield strength of the Zircaloy-4 material. The model was accurate enough to predict that no yielding of the cladding was expected, but its precision at predicting micro strains is questionable. The SNL test data offers some opportunity for validation of the finite element model, but the specific loading conditions of the testing only excite the fuel assembly to respond in a limited manner. For example, the test accelerations were not strong enough to substantially drive the fuel assembly out of contact with the basket. Under this test scenario, the fuel assembly model does a reasonable job of approximating actual fuel assembly response, a claim that can be verified through

  6. Monte Carlo modeling and analyses of YALINA- booster subcritical assembly Part II : pulsed neutron source.

    Energy Technology Data Exchange (ETDEWEB)

    Talamo, A.; Gohar, M. Y. A.; Rabiti, C.; Nuclear Engineering Division

    2008-10-22

    One of the most reliable experimental methods for measuring the kinetic parameters of a subcritical assembly is the Sjoestrand method applied to the reaction rate generated from a pulsed neutron source. This study developed a new analytical methodology for characterizing the kinetic parameters of a subcritical assembly using the Sjoestrand method, which allows comparing the analytical and experimental time dependent reaction rates and the reactivity measurements. In this methodology, the reaction rate, detector response, is calculated due to a single neutron pulse using MCNP/MCNPX computer code or any other neutron transport code that explicitly simulates the fission delayed neutrons. The calculation simulates a single neutron pulse over a long time period until the delayed neutron contribution to the reaction is vanished. The obtained reaction rate is superimposed to itself, with respect to the time, to simulate the repeated pulse operation until the asymptotic level of the reaction rate, set by the delayed neutrons, is achieved. The superimposition of the pulse to itself was calculated by a simple C computer program. A parallel version of the C program is used due to the large amount of data being processed, e.g. by the Message Passing Interface (MPI). The new calculation methodology has shown an excellent agreement with the experimental results available from the YALINA-Booster facility of Belarus. The facility has been driven by a Deuterium-Deuterium or Deuterium-Tritium pulsed neutron source and the (n,p) reaction rate has been experimentally measured by a {sup 3}He detector. The MCNP calculation has utilized the weight window and delayed neutron biasing variance reduction techniques since the detector volume is small compared to the assembly volume. Finally, this methodology was used to calculate the IAEA benchmark of the YALINA-Booster experiment.

  7. Nuclear fuel assembly debris filter

    International Nuclear Information System (INIS)

    This patent describes a nuclear fuel assembly having fuel rods held in a spaced array by grid assemblies, guide tubes extending through the grid assemblies and attached at their upper and lower ends to an upper end fitting and a lower end fitting, the end fittings having openings therethrough for coolant flow, and a debris filter. The debris filter comprises: a plate attached to the bottom periphery of and spanning the lower end fitting; and the plate having substantially triangular-shaped flow holes therethrough that each measure approximately 0.181 inch from the base to the apex with the majority of the triangular- shaped flow holes arranged in groups of four to define square clusters that each measure approximately 0.405 inch on each side whereby the portions of the plate between the flow holes in each cluster are diagonally oriented relative to the sides of the plate

  8. Alternative systems for fuel gas boosters for small gas turbine engines

    Science.gov (United States)

    Faulkner, Henry B.

    1992-04-01

    The study was done to investigate alternative technologies for fuel gas boosters for gas turbine engines under 5 MW output. The goal was to identify concepts which would significantly reduce the overall life cycle cost of these boosters. In a broad review of alternative systems, centrifugal compressors were found to be most promising. Electrically driven centrifugals, either direct drive or gear driven, were found to be quite limited in speed. Therefore they require many stages for these applications, and no cost advantage was indicated. Considerable promise was indicated for centrifugals driven by bleed air from the engine compressor, using turbocompressor units which are conversions of existing turbochargers for internal combustion engines. A first cost advantage of 30 to 80 percent was indicated for applications with an annual market size of at least ten units. Considerable savings in installation and maintenance costs are expected in addition.

  9. System for assembling nuclear fuel elements

    International Nuclear Information System (INIS)

    An automatic system is described for assembling nuclear fuel elements, in particular those employing mixed oxide fuels. The system includes a sealing mechanism which allows movement during the assembling of the fuel element along the assembly stations without excessive release of contaminants. (U.K.)

  10. Fuel assembly for a boiling water reactor

    International Nuclear Information System (INIS)

    A boiling water reactor fuel assembly is described which has vertical fuel rods and guide tubes positioned below the fuel rods and receiving control rod fingers and acting as water pipes, the guide tubes each being formed of a plurality of parts including a part secured to a grid plate positioned in the fuel assembly container, and low parts which fit into holes formed in the bottom of the fuel assembly. There is a flexible connection between the upper and lower parts of the guide tubes to allow for a certain tolerance in the procedure of manufacturing the various parts to allow insertion of the fuel rod bundle into the fuel assembly container

  11. Monte Carlo modeling and analyses of YALINA-booster subcritical assembly part 1: analytical models and main neutronics parameters.

    Energy Technology Data Exchange (ETDEWEB)

    Talamo, A.; Gohar, M. Y. A.; Nuclear Engineering Division

    2008-09-11

    This study was carried out to model and analyze the YALINA-Booster facility, of the Joint Institute for Power and Nuclear Research of Belarus, with the long term objective of advancing the utilization of accelerator driven systems for the incineration of nuclear waste. The YALINA-Booster facility is a subcritical assembly, driven by an external neutron source, which has been constructed to study the neutron physics and to develop and refine methodologies to control the operation of accelerator driven systems. The external neutron source consists of Californium-252 spontaneous fission neutrons, 2.45 MeV neutrons from Deuterium-Deuterium reactions, or 14.1 MeV neutrons from Deuterium-Tritium reactions. In the latter two cases a deuteron beam is used to generate the neutrons. This study is a part of the collaborative activity between Argonne National Laboratory (ANL) of USA and the Joint Institute for Power and Nuclear Research of Belarus. In addition, the International Atomic Energy Agency (IAEA) has a coordinated research project benchmarking and comparing the results of different numerical codes with the experimental data available from the YALINA-Booster facility and ANL has a leading role coordinating the IAEA activity. The YALINA-Booster facility has been modeled according to the benchmark specifications defined for the IAEA activity without any geometrical homogenization using the Monte Carlo codes MONK and MCNP/MCNPX/MCB. The MONK model perfectly matches the MCNP one. The computational analyses have been extended through the MCB code, which is an extension of the MCNP code with burnup capability because of its additional feature for analyzing source driven multiplying assemblies. The main neutronics parameters of the YALINA-Booster facility were calculated using these computer codes with different nuclear data libraries based on ENDF/B-VI-0, -6, JEF-2.2, and JEF-3.1.

  12. Fuel assembly self-excited vibration and test methodology

    International Nuclear Information System (INIS)

    PWR fuel assemblies normally experience low amplitude, random vibration under normal reactor flow conditions. This normal fuel assembly vibration has almost no impact on grid-rod fretting wear. However, some fuel assembly designs experience a high resonant fuel assembly vibration under normal axial flow conditions. This anomalous fuel assembly vibration is defined as fuel assembly self-excitation vibration (FASE), because the assembly vibrates resonantly without any external periodic excitation force. Fuel assembly self-excitation vibration can cause severe grid-rod fretting if the assembly operates at the flow rate, which causes high fuel assembly vibration. This paper will describe the characteristics of fuel assembly self-excitation vibration and the test methodology to identify the fuel assembly vibration. Several fuel assembly designs are compared under standard test conditions. The causes for the fuel assembly self-excitation vibration are analyzed and discussed. The test acceptance criteria are defined for newly developed PWR fuel assemblies. (authors)

  13. Impact Analysis for Fuel Assemblies in Spent Fuel Storage Rack

    International Nuclear Information System (INIS)

    The design and structural integrity evaluation of a spent fuel storage rack (SFSR) utilized for storing and protecting the spent fuel assemblies generated during the operation of a reactor are very important in terms of nuclear safety and waste management. The objective of this study is to show the validity of the SFSR design as well as fuel assembly through a structural integrity evaluation based on a numerical analysis. In particular, a dynamic time history analysis considering the gaps between the fuel assemblies and the walls of the storage cell pipes in the SFSR was performed to check the structural integrity of the fuel assembly and storage cell pipe

  14. Nuclear fuel assembly identification using computer vision

    International Nuclear Information System (INIS)

    This report describes an improved method of remotely identifying irradiated nuclear fuel assemblies. The method uses existing in-cell TV cameras to input an image of the notch-coded top of the fuel assemblies into a computer vision system, which then produces the identifying number for that assembly. This system replaces systems that use either a mechanical mechanism to feel the notches or use human operators to locate notches visually. The system was developed for identifying fuel assemblies from the Fast Flux Test Facility (FFTF) and the Clinch River Breeder Reactor, but could be used for other reactor assembly identification, as appropriate

  15. Fuel fire tests of selected assemblies

    Science.gov (United States)

    Kydd, G.; Spindola, K.; Askew, G. K.

    1982-04-01

    A varing assortment of clothing assemblies was tested in the Fuel Fire Test Facility at the Naval Air Development Center. Included was a Nomex-Kevlar Cloque Coverall which had relatively good protection from fuel flames.

  16. Evaluation of reactivity monitoring techniques in experiments with pulsed neutron source in the Yalina-Booster subcritical assembly; Evaluacion de tecnicas de monitorizacion de la reactividad en experimentos con fuente de neutrones pulsada en el conjunto subcritico Yalina-Booster

    Energy Technology Data Exchange (ETDEWEB)

    Becares, V.; Villamarin, D.; Fernandez-Ordonez, M.; Gonzalez-Romero, E. M.

    2010-07-01

    As a part of EUROTRANS program, it has carried out an experimental campaign focused in the validation of reactivity monitoring techniques in the Yalina-Booster subcritical assembly. The aim of this paper is to present the analysis of part of the experiments results, in particular those carried out with a pulsed neutron source.

  17. Assembly mechanism for nuclear fuel bundles

    International Nuclear Information System (INIS)

    This invention relates to an assembly mechanism for nuclear power reactor fuel bundles using a novel, simple and inexpensive means. The mechanism is readily operable remotely, avoids separable parts and is applicable to fuel assemblies in which the upper tie plate is rigidly mounted on the tie rods which hold it in place. (UK)

  18. Poolside inspection facility for PWR fuel assemblies

    International Nuclear Information System (INIS)

    Pool side inspection programme for LWRs started in India with the inspection of BWR fuel assemblies at Tarapur and this involved sipping, visual inspection, UT and Eddy current testing. In view of the possibility of having WWER type of reactors in our country, a R and D program has been initiated for study of behavior of these types of fuel. The program would involve irradiation, pool side inspection and hot cell examination of specially designed fuel assemblies. Well characterized fuel assemblies irradiated in research reactor are transferred to the fuel pool with the help of fuel transfer system. The fuel assemblies are taken out of the transfer system, sipping test performed and de channeled using under water handling and cutting tools. The fuel pins are then taken out of assembly and loaded on to the stand for underwater UT and Eddy current testing. The details of the handling and inspection facilities provided in pool for inspection of the hexagonal fuel assemblies has been discussed in the text. Dismantling and inspection procedure used for control assembly pins have also been discussed. (author)

  19. Fuel sub-assemblies for nuclear reactors

    International Nuclear Information System (INIS)

    A fuel assembly for a liquid metal cooled fast breeder nuclear reactor comprises a bundle of spaced fuel pins within a tubular wrapper or sleeve. The wrapper is extended at one end by a tubular neutron shield of massive steel and the other end, has a spike extension whereby the sub-assembly can be located by plugging into a support structure. The invention provides that lateral displacement of individual fuel pin-containing wrappers to accommodate dimensional changes within the fuel assembly is effected by movement of each wrapper relative to its spike extension. (author)

  20. Thermal Analysis of a TREAT Fuel Assembly

    Energy Technology Data Exchange (ETDEWEB)

    Papadias, Dionissios [Argonne National Lab. (ANL), Argonne, IL (United States); Wright, Arthur E. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2014-07-09

    The objective of this study was to explore options as to reduce peak cladding temperatures despite an increase in peak fuel temperatures. A 3D thermal-hydraulic model for a single TREAT fuel assembly was benchmarked to reproduce results obtained with previous thermal models developed for a TREAT HEU fuel assembly. In exercising this model, and variants thereof depending on the scope of analysis, various options were explored to reduce the peak cladding temperatures.

  1. Modular fuel-cell stack assembly

    Science.gov (United States)

    Patel, Pinakin

    2010-07-13

    A fuel cell assembly having a plurality of fuel cells arranged in a stack. An end plate assembly abuts the fuel cell at an end of said stack. The end plate assembly has an inlet area adapted to receive an exhaust gas from the stack, an outlet area and a passage connecting the inlet area and outlet area and adapted to carry the exhaust gas received at the inlet area from the inlet area to the outlet area. A further end plate assembly abuts the fuel cell at a further opposing end of the stack. The further end plate assembly has a further inlet area adapted to receive a further exhaust gas from the stack, a further outlet area and a further passage connecting the further inlet area and further outlet area and adapted to carry the further exhaust gas received at the further inlet area from the further inlet area to the further outlet area.

  2. Method of straightening a bowed nuclear fuel assembly

    International Nuclear Information System (INIS)

    A method of removing bow in a nuclear fuel assembly is disclosed. The fuel assembly has top and bottom ends fittings and a plurality of longitudinally extending thimble tube members interconnecting top and bottom end fittings. At least two transverse fuel rod support grids are axially spaced along the thimble tube members. A plurality of fuel rods are transversely spaced and supported by the fuel rod support grids. In one embodiment, a weight of known magnitude is secured on the bottom end fitting and the fuel assembly is raised with the weight secured thereon so that the weight exerts a downward force on the fuel assembly for straightening the fuel assembly and eliminating compressive stresses within the fuel assembly. In another embodiment, the bottom end fitting is secured onto the upender used for transporting fuel assemblies into and out of the containment building and the fuel assembly is pulled for straightening the fuel assembly and eliminating compressive stresses within the fuel assembly. (Author)

  3. Seismic response of nuclear fuel assembly

    Directory of Open Access Journals (Sweden)

    Hlaváč Z.

    2014-06-01

    Full Text Available The paper deals with mathematical modelling and computer simulation of the seismic response of fuel assembly components. The seismic response is investigated by numerical integration method in time domain. The seismic excitation is given by two horizontal and one vertical synthetic accelerograms at the level of the pressure vessel seating. Dynamic response of the hexagonal type nuclear fuel assembly is caused by spatial motion of the support plates in the reactor core investigated on the reactor global model. The modal synthesis method with condensation is used for calculation of the fuel assembly component displacements and speeds on the level of the spacer grid cells.

  4. Irradiated MTR fuel assemblies sipping test

    International Nuclear Information System (INIS)

    This paper describes the procedure and methodology used to perform sipping test with the IEA-R1 fuel assemblies at the storage pool, and presents the results obtained for Cs-137 sipping water activity for each fuel assembly analyzed. Discussion is made correlating corrosion pits to the activity values measured. A Cs-137 leaking rate is determined which can be compared to the criteria established for canning spent fuel assemblies inside the pool of for shipment abroad. 3 refs., 13 figs., 1 tab

  5. Fuel assembly identification for nuclear power reactors

    International Nuclear Information System (INIS)

    The standard refers to fuel assemblies of light-water power reactors. It contains stipulations for uniform marking in order that the fuel assemblies may be identified. A figure consisting of 8 alpha-numerical characters is used for marking, the first three of which represent the operator who ordered the fuel assembly, while the four characters to follow symbolize a series number. The last character serves as a test mark to scrutinize reading mistakes. The alpha-numerical characters include the Arabic numerals 0-9 and, following them, the letters A-Y of the German alphabet, leaving out B, F, I, O, Q, Z (30 characters). (orig./HP)

  6. Fuel assembly identification for power reactors

    International Nuclear Information System (INIS)

    The standard refers to fuel assemblies of light-water power reactors. It contains stipulations for uniform marking in order that the fuel assemblies may be identified. A figure consisting of 8 alpha-numerical numbers is used for marking, the first three of which represent the operator who ordered the fuel assembly, while the four numbers to follow symbolize a series number. The last number serves as a test mark to scrutinize reading mistakes. The alpha-numerical numbers include the Arabic numerals 0-9 and, following them, the letters A-Y of the German alphabet, leaving out B, F, I, O, Q, Z (30 characters). (orig./HP)

  7. Sipping inspection method for fuel assembly

    International Nuclear Information System (INIS)

    Purpose: To shorten the periodical inspection time required for sipping inspection (failed fuel inspection) in PWR type reactors or the likes by mounting a plurality of a fuel assemblies in a sipping can to thereby inspect the plurality of fuel assemblies simultaneously. Method: Upon sipping inspection of fuels temporarily stored in a spent fuel storage pool, a plurality of fuel assemblies are mounted in a sipping can. Then, after placing a cover, N2 gases are caused to flow as carrier gas from a gas container by the operation to valves. Then, N2 gases are circulated by a jet pump and water in the can is sucked by a water pump. A radioactivity detector such as a scintillation counter is provied to a part of the loop for measuring the amount of the radioactivity in the portion where water is passed to check whether leakage is present or not. (Horiuchi, T.)

  8. Fuel assembly for BWR type reactor

    International Nuclear Information System (INIS)

    A fuel assembly of an BWR type reactor of present invention, in which a plurality of fuel rods are arranged in a regular square lattice like configuration include vibration-filled fuel rods in which granular nuclear fuel materials and granular non-nuclear fuel materials having a smaller neutron absorbing cross sectional area are mixed and filled. With such a constitution, the content of the mixed and filled non-nuclear fuel materials in the vibration filled fuel rods is at least 20% by a volume ratio in average in fuel assemblies. In addition, a burnable poison is optionally added and mixed to the granular mixture of the nuclear fuel material and the diluting granules. With such a constitution, the manufacturing cost can be reduced, and the combustion rate of the nuclear fission materials is increased to improve reactor core characteristics, thereby enabling to obtain sufficient Pu loading amount per assembly, and fuel assemblies excellent in flexibility in design and economic property can be obtained. (T.M.)

  9. Dimension detection device for fuel assembly

    International Nuclear Information System (INIS)

    The present invention provides a device of facilitating remote and automatic inspection for the outer diameter of spacers and the length of springs of fuel assemblies to be used in a nuclear power plant. Namely, the device of the present invention comprises a mechanism for vertically supporting and rotating the fuel assemblies, a sensor holding frame equipped with a displacement sensor for detecting dimension, a mechanism for vertically moving the holding frame, and a mechanism for horizontally moving the holding frame. The dimension of the fuel assemblies is detected based on the moving amount of the horizontally moving mechanism. Even if the fuel assemblies are twisted, tilted or deviated to front-to-back or right-to-left direction, data can be collected/amended based on the displacement of each mechanisms. According to the device of the present invention, since automatic and remote inspection is possible, an operator's radiation exposure can be reduced. (I.S.)

  10. Establishment of China Nuclear Fuel Assembly Database

    Institute of Scientific and Technical Information of China (English)

    CHENPeng; ZHANGYing-chao; LIUTing-jin; JINYong-li

    2003-01-01

    During researching, designing, manufacturing and post irradiation, a large amount of data on fuel assembly of China nuclear power plants has been accumulated. It is necessary to collect the data together,so that the researchers, designers, manufactures and managers could use the data conveniently. It was proposed to establish a China Nuclear Fuel Assembly Database through the Internet on workstations during the year of 2003 to 2006, so the data would be shared in China nuclear industry.

  11. A classification scheme for LWR fuel assemblies

    International Nuclear Information System (INIS)

    With over 100 light water nuclear reactors operating nationwide, representing designs by four primary vendors, and with reload fuel manufactured by these vendors and additional suppliers, a wide variety of fuel assembly types are in existence. At Oak Ridge National Laboratory, both the Systems Integration Program and the Characteristics Data Base project required a classification scheme for these fuels. This scheme can be applied to other areas and is expected to be of value to many Office of Civilian Radioactive Waste Management programs. To develop the classification scheme, extensive information on the fuel assemblies that have been and are being manufactured by the various nuclear fuel vendors was compiled, reviewed, and evaluated. It was determined that it is possible to characterize assemblies in a systematic manner, using a combination of physical factors. A two-stage scheme was developed consisting of 79 assembly types, which are grouped into 22 assembly classes. The assembly classes are determined by the general design of the reactor cores in which the assemblies are, or were, used. The general BWR and PWR classes are divided differently but both are based on reactor core configuration. 2 refs., 15 tabs

  12. Fluid flow test for KMRR fuel assemblies

    International Nuclear Information System (INIS)

    The objective of this test program is to obtain the experimental data of pressure drop and subchannel flow distribution for the KMRR (Korea Multipurpose Research Reactor) fuel assembly, and to investigate mechanical integrity of the fuel assembly and flow tube in the test flow condition. The experimental data produced through this study are applicable to the KMRR fuel design and thermal-hydraulic analysis of the reactor. Pressure drop correlations for each types of fuels were developed which can be applicable over Reynolds number of 6x9x102∼8.0x104. Local velocity in the subchannels of the fuel assemblies was measured with laser doppler velocimeter system, and the velocitily distribution was also calculated with a computer program developed through this study. The experimental data are used as input for the core thermal margin analysis and safety analysis in steady/accident conditions of the KMRR. (Author)

  13. Nuclear reactor fuel assembly with fuel rod removal means

    International Nuclear Information System (INIS)

    A fuel assembly is described for a nuclear reactor. The assembly has a bottom nozzle, at least one longitudinally extending control rod guide thimble attached to and projecting upwardly from the bottom nozzle and transverse grids spaced along the thimble. An organized array of elongated fuel rods are transversely spaced and supported by the grids and axially captured between the bottom nozzle and a top nozzle. The assembly comprises: (a) a transversely extending adapter plate formed by an arrangement of integral cross-laced ligaments defining a plurality of coolant flow openings; (b) means for mounting the adapter plate on an upper end portion of the thimble and spaced axially above and disposed transversely over the upper ends of all of the fuel rods present in the fuel assembly such that ones of the ligaments overlie corresponding ones of the fuel rods so as to prevent the fuel rods from moving upwardly through the coolant flow openings; and (c) removable plug means confined within the adapter plate and positioned over and spaced axially above selected ones of the fuel rods in providing access to at least one fuel rod for removal thereof upwardly through the axially spaced adapter plate without removing the top nozzle from the fuel assembly

  14. Fuel assembly support structure for reactor

    International Nuclear Information System (INIS)

    Purpose: To restrict a part or whole of injected molten fuel within a predetermined area by constructing interior surfaces of walls and bottom of a molten fuel container with high-melting materials, fixing at an upper opening a fuel assembly support cover, and forming a thru-hole in the bottom. Constitution: A plurality of cover-fitted support elements for fuel assemblies are mounted and fixed on the supports in the reactor container, with leg pipes inserted in insert holes. Fuel assemblies are fitted in insert holes of the support elements to make up the core. If the power increases due to an accident, causing failure of fuel cans installed in bundle in the assemblies, most of the molten fuel falls to the bottom of the container. As the bottom is graded down from center to periphery, the molten fuel settles much in the peripheral area, but the part is lined with high-melting material so that the part will not be melted. (Yoshihara, H.)

  15. Fuel assembly for a boiling water reactor

    International Nuclear Information System (INIS)

    The fuel assembly of a boiling water reactor contains a number of vertical fuel rods with their lower ends against a bottom tie plate. The rods are positioned by spacers, which are fixed to the canning. The upward motion is reduced by the top plate of a special design. (G.B.)

  16. Nuclear reactor fuel assembly spacer grids

    International Nuclear Information System (INIS)

    Designs of nuclear reactor fuel assembly spacer grids for supporting and spacing fuel elements are described which do not utilize resilient grid plate protrusions in the peripheral band but retain the advantages inherent in the combination resilient and rigid protrusion cells. (U.K.)

  17. Grid structure for nuclear reactor fuel assembly

    International Nuclear Information System (INIS)

    Described is a nuclear fuel element support system comprising an egg-crate-type grid made up of slotted vertical portions interconnected at right angles to each other, the vertical portions being interconnected by means of cross straps which are dimpled midway between their ends to engage fuel elements disposed within openings formed in the egg-crate assembly. The cross straps are disposed at an angle, other than a right angle, to the vertical portions of the assembly whereby their lengths are increased for a given span, and the total elastic deflection capability of the cell is increased. The assembly is particularly adapted for computer design and automated machine tool fabrication

  18. Grid for nuclear reactor fuel assemblies

    International Nuclear Information System (INIS)

    A grid of improved design for a nuclear reactor fuel assembly which includes a multiplicity of interleaved straps enclosed in a peripheral frame which forms a grid of egg-crate configuration is described. Each cell formed by the grid straps, except those containing control rod guide tubes, supports a fuel rod which is held in place by springs projecting laterally inwardly into each cell from the grid straps. The springs extend parallel to the fuel rods and are spaced at 900 intervals around the rod. Further, each of two adjacent springs contact a fuel rod at two points along its length and each of the other two adjacent springs contact the fuel rod at one point thus imparting strength and flexibility to the fuel assembly containing such grids

  19. Apparatus and method for assembling fuel elements

    International Nuclear Information System (INIS)

    A nuclear fuel element assembling method and apparatus is preferably operable under programmed control unit to receive fuel rods from storage, arrange them into axially aligned stacks of closely monitored length, and transfer the stacks of fuel rods to a loading device for insertion into longitudinal passages in the fuel elements. In order to handle large numbers of one or more classifications of fuel rods or other cylindrical parts, the assembling apparatus includes at least two feed troughs each formed by a pair of screw members with a movable table having a plurality of stacking troughs for alignment with the feed troughs and with a conveyor for delivering the stacks to the loading device, the fuel rods being moved along the stacking troughs upon a fluid cushion. 23 claims, 6 figures

  20. Fuel assembly for BWR type reactor

    International Nuclear Information System (INIS)

    A fuel assembly has a 9 x 9 square lattice arrangement having a water channel which occupies an area of 3 x 3 lattice pattern corresponding to 9 fuel rods. Fuel pellets comprise those of not more 7 kinds which have fission products at enrichment degrees different by a spun of not less than 10%. Fuel rods comprise from 4 to 12 first type fuel rods and remaining second type fuel rods. The first type fuel rod is loaded with fuel pellets of fissionable products having an enrichment degree axially different at the upper and the lower portions. The second type fuel rod is loaded with fuel pellets of fissionable products having the same enrichment degree in the vertical direction. With such a constitution, the enrichment degree of fissionable products of fuel pellets in the fuel assembly for a BWR type reactor having different reactor constitution and operation conditions can be used in common. Accordingly, the degree of freedom for the design of the distribution of the enrichment degree is increased. (I.N.)

  1. Fuel assembly for a nuclear reactor

    International Nuclear Information System (INIS)

    The subject of the patent is a spacer design applicable, primarily, to LWR, and especially, though not specifically PWR, fuel assemblies. The spacer consists of an egg-box type of assembly formed of interlocking pressed plates giving a square lattice whose openings accommodate fuel pins or regulating rods. The pressed plates are formed to provide pressed-out spring-like flanges which hold the fuel pins in position and guide the regulating rods. Additional pressed-out flanges ensure the correct configuration of the spacer structure. The spacer is designed to present as little resistance as possible to coolant flow. (JIW)

  2. Reconstitutable fuel assembly for a nuclear reactor

    International Nuclear Information System (INIS)

    A reconstitutable fuel assembly for a nuclear reactor which includes a mechanical, rather than metallurgical, arrangement for connecting control rod guide thimbles to the top and bottom nozzles of a fuel assembly. Multiple sleeves enclosing control rod guide thimbles interconnect the top nozzle to the fuel assembly upper grid. Each sleeve is secured to the top nozzle by retaining rings disposed on opposite sides of the nozzle. Similar sleeves enclose the lower end of control rod guide thimbles and interconnect the bottom nozzle with the lowermost grid on the assembly. An end plug fitted in the bottom end of each sleeve extends through the bottom nozzle and is secured thereto by a retaining ring. Should it be necessary to remove a fuel rod from the assembly, the retaining rings in either the top or bottom nozzles may be removed to release the nozzle from the control rod guide thimbles and thus expose either the top or bottom ends of the fuel rods to fuel rod removing mechanisms

  3. Nuclear fuel bundle disassembly and assembly tool

    International Nuclear Information System (INIS)

    A nuclear power reactor fuel bundle is described which has a plurality of tubular fuel rods disposed in parallel array between two transverse tie plates. It is secured against disassembly by one or more locking forks which engage slots in tie rods which position the transverse plates. Springs mounted on the fuel and tie rods are compressed when the bundle is assembled thereby maintaining a continual pressure against the locking forks. Force applied in opposition to the springs permits withdrawal of the locking forks so that one tie plate may be removed, giving access to the fuel rods. An assembly and disassembly tool facilitates removal of the locking forks when the bundle is to be disassembled and the placing of the forks during assembly of the bundle. (U.S.)

  4. WWER-440 fuel cycles possibilities using improved fuel assemblies design

    International Nuclear Information System (INIS)

    Practically five years cycle has been achieved in the last years at NPP Dukovany. There are two principal means how it could be achieved. First, it is necessary to use fuel assemblies with higher fuel enrichment and second, to use fuel loading with very low leakage. Both these conditions are fulfilled at NPP Dukovany at this time. It is known, that the fuel cycle economy can be improved by increasing the fuel residence time in the core up to six years. There are at least two ways how this goal could be achieved. The simplest way is to increase enrichment in fuel. There exists a limit, which is 5.0 w % of 235U. Taking into account some uncertainty, the calculation maximum is 4.95 w % of 235U. The second way is to change fuel assembly design. There are several possibilities, which seem to be suitable from the neutron - physical point of view. The first one is higher mass content of uranium in a fuel assembly. The next possibility is to enlarge pin pitch. The last possibility is to 'omit' FA shroud. This is practically unrealistic; anyway, some other structural parts must be introduced. The basic neutron physical characteristics of these cycles for up-rated power are presented showing that the possibilities of fuel assemblies with this improved design in enlargement of fuel cycles are very promising. In the end, on the basis of neutron physical characteristics and necessary economical input parameters, a preliminary evaluation of economic contribution of proposals of advanced fuel assemblies on fuel cycle economy is presented (Authors)

  5. Polymer electrolyte membrane assembly for fuel cells

    Science.gov (United States)

    Yen, Shiao-Ping S. (Inventor); Kindler, Andrew (Inventor); Yavrouian, Andre (Inventor); Halpert, Gerald (Inventor)

    2002-01-01

    An electrolyte membrane for use in a fuel cell can contain sulfonated polyphenylether sulfones. The membrane can contain a first sulfonated polyphenylether sulfone and a second sulfonated polyphenylether sulfone, wherein the first sulfonated polyphenylether and the second sulfonated polyphenylether sulfone have equivalent weights greater than about 560, and the first sulfonated polyphenylether and the second sulfonated polyphenylether sulfone also have different equivalent weights. Also, a membrane for use in a fuel cell can contain a sulfonated polyphenylether sulfone and an unsulfonated polyphenylether sulfone. Methods for manufacturing a membrane electrode assemblies for use in fuel cells can include roughening a membrane surface. Electrodes and methods for fabricating such electrodes for use in a chemical fuel cell can include sintering an electrode. Such membranes and electrodes can be assembled into chemical fuel cells.

  6. Fuel assembly for BWR type nuclear reactor

    International Nuclear Information System (INIS)

    In the fuel assembly of the present invention, fuel rods and one or a plurality of water rods or water channels are bundled by upper and lower tie plates and one or more of spacers, and the outer circumference of the bundle is covered with a channel box. In the present invention, a groove capable of flowing coolants is disposed on the surface of the water rod or the water channel. Specifically, the groove is disposed, continuously or intermittently, at portions corresponding to the first spacer and from the second to the fourth spacers. With such a constitution, coolants stagnating at the upper portion of the spacer due to gas/liquid counter flow limit (CCFL) are caused to flow down passing through the groove easily upon occurrence of LOCA. Accordingly, cooling of fuel rods at the center of the fuel assembly can be promoted, thereby suppressing the temperature elevation on the surface of the fuel rods. (I.S.)

  7. TRU transmutation type BWR fuel assembly

    International Nuclear Information System (INIS)

    The BWR fuel assembly is formed by bundling a plurality of fuel rods and a water channel disposed at the center of the assembly by a plurality of spacers. An upper tie plate and a lower tie plate are disposed to upper and lower portions of the fuel rods and the water channel respectively. An upper end plug of the water channel is attached detachably to a cylindrical main body of the water channel. A zircaloy tube incorporating TRU nuclides is contained and secured in the water channel. The zircaloy tube has such a structure as capable of incorporating and sealing oxides or metal materials containing TRU nuclides. Since the zircaloy tube containing TRU nuclides is contained not in fuel region but in the water rod, the loaded uranium amount of fuels is not reduced but the reactivity can be ensured. (I.N.)

  8. Appearance detection device for fuel assembly

    International Nuclear Information System (INIS)

    The present invention concerns a device for detecting the appearance of fuel assemblies for a power plant, in which the device photographs corners of fuel assemblies by a TV-camera to perform detection with higher reliability. Namely, heretofore, fuel assembly to substantially square pillar shape for a BWR and a PWR has been rotated and one or two faces have been detected from the front by the TV-camera. In the present invention, a TV-camera used exclusively for corners is additionally disposed on or near the diagonal line of the corners. With such a constitution, corners of the fuel assemblies can be photographed simultaneously with the conventional appearance test. As a result, since appearance test for the front and the corners can be conducted at the same time, extremely effective detection can be conducted in terms of detection of a rupture of grids and prevention of dead angle. The corners of assemblies which tend to undergo damages upon charge/discharge of fuels can be detected carefully. Accordingly, a highly reliable detection can be conducted. (I.S.)

  9. WWER-440 fuel cycles possibilities using modified fuel assemblies design

    International Nuclear Information System (INIS)

    A nearly equilibrium five-year cycle has been achieved at Dukovany NPP over the last years. This means that working fuel assemblies with an average enrichment of 4.25 w % (control assemblies) with an average enrichment of 3.82 w %) are normally loaded and reloaded for five years. Operation at uprated thermal power (105% of the original one, increase from 1375 MWt to 1444 MWt) is being prepared by use of working fuel assemblies with an average enrichment of 4.38 w % (control assemblies with an average enrichment of 4.25 w %). With the aim of fuel cycle economy improvement, the fuel residence time in the core has to be prolonged up to six years with one cycle duration time up to 18 months and preserving loadings with very low leakage. In order to achieve this goal, at least neutron-physical characteristics of fuel assemblies must be improved and such changes should be evaluated from other viewpoints. Some particular changes have already been analyzed earlier. Designs of new fuel assemblies with higher (and in the central part of a fuel assemblies the highest possible, i.e. 4.95 w %) enrichment with preserving low pin power non-uniformity are described in the presented paper. An fuel assemblies with an average enrichment of 4.66 w % (lower than originally evaluated) containing six fuel pins with 3.35 w % Gd2O3 content was selected in the end. Fuel pins have bigger pellet diameter, bigger pin pitch and thinner fuel assemblies shroud. A newly designed fuel assemblies was evaluated from the viewpoint of physics (pin power non-uniformity, criticality of fuel at transport and storage and determination of basic quantities for spent fuel storage purposes by ORIGEN code), thermo-hydraulics (comparison of subchannel output temperatures and the departure from nucleate boiling ratio - DNBR) and mechanical properties. The purpose of this study was to simulate an fuel assemblies subject to the loads during its six- year lifetime whereas normal working conditions were taken into

  10. Grid for nuclear fuel assembly

    International Nuclear Information System (INIS)

    A spacer grid for nuclear fuel rods is formed of generally identical metal straps arranged in crossed relation to define a multiplicity of cells adapted to receive elongated fuel elements or the like. The side walls of each cell have openings for intercell mixing of coolant and tabs from edges of the openings defining helical coolant deflectors in the cells. Tabs from adjacent side walls are fixedly secured together to provide rigidifying flanges for the grid. Spring fingers at the ends of the cells provide for holding fuel rods against fixed stops

  11. Modular fuel-cell stack assembly

    Science.gov (United States)

    Patel, Pinakin; Urko, Willam

    2008-01-29

    A modular multi-stack fuel-cell assembly in which the fuel-cell stacks are situated within a containment structure and in which a gas distributor is provided in the structure and distributes received fuel and oxidant gases to the stacks and receives exhausted fuel and oxidant gas from the stacks so as to realize a desired gas flow distribution and gas pressure differential through the stacks. The gas distributor is centrally and symmetrically arranged relative to the stacks so that it itself promotes realization of the desired gas flow distribution and pressure differential.

  12. Fuel assembly for a boiling water reactor

    International Nuclear Information System (INIS)

    A fuel assembly for a boiling water reactor comprises a plurality of fuel rods which constitute four partial bundles and are surrounded by a fuel channel system comprising one partial tube for each partial bundle. Each of the four partial bundles rests on a bottom tie plate and is positioned with respect to the others by means of a common top tie plate which is provided with a lifting loop which is sufficiently strong to be able to lift the four partial bundles simultaneously, a major part of the lifting force being transmitted to said bottom tie plates via a plurality of supporting fuel rods

  13. Blockages in LMFBR fuel assemblies: a review

    Energy Technology Data Exchange (ETDEWEB)

    Han, J T; Fontana, M H

    1977-01-01

    Experimental and analytical investigations performed in the United States, Germany, Great Britain, and Japan on the effects of partial flow blockages in liquid-metal fast breeder reactor fuel assemblies are reviewed and the results presented. Generalized models are developed from experimental data obtained for blockages of various sizes, shapes, and porosity, with and without pins, utilizing water and sodium as the coolant. Generally, the recirculating flow in the wake behind a blockage is a relatively effective heat transfer mechanism. Experiments where sodium boiling was made to occur behind the blockages indicate that boiling is stable for the configurations tested; these results are predicted by analytical models. Blockages at the inlet of fuel assemblies tend to have insignificant effects in the fuel assembly unless flow is reduced grossly and therefore would be detectable. Blockages in the heat generating zone have to be quite large to cause sodium boiling under normal reactor operating conditions.

  14. Fuel assembly for BWR type reactor

    International Nuclear Information System (INIS)

    When fuel rods are suddenly oscillated by earthquakes, and a void ratio is abruptly reduced, it is forecast that feed back of negative reactivity due to generation of voids is delayed to cause power increase in a short period of time. Then, in a fuel assembly comprising a large number of fuel rods bundled by an upper tie plate, a lower tie plate and a plurality of spacers and contained in a channel box, stirring means for coolants flowing the periphery of fuel rods are disposed in a lower sub-cool boiling region. Coolants flown into the fuel assembly are directed to fuel rods by the coolant stirring means to mix the coolants, whereby the temperature difference between the periphery of the surface of the fuel rods and bulk coolants is reduced, to decrease a sub-cool void amount. Then, even if the fuel rods are oscillated, the reduction of a sub-cool void ratio is small, which scarcely gives influences of fuel rod oscillation on the power of the reactor core. (N.H.)

  15. Fuel assembly for BWR type reactor

    International Nuclear Information System (INIS)

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

  16. Core and transition fuel assembly for a nuclear reactor

    International Nuclear Information System (INIS)

    A core and a transition fuel assembly for a nuclear reactor are described. They have a first fuel assembly including structure for laterally spacing parallel and coextending fuel rods positioned at preselected core elevations and also a second fuel assembly including lateral spacing structure at preselected core elevations at least one of which is different than the elevations of the spacing structure of the first fuel assembly. The transition fuel assembly is positioned between the first and second assemblies and includes lateral spacing structure positioned at each core elevation where the first and second fuel assemblies have a spacing structure. The transition fuel assembly ensures that contact among the fuel assemblies of the core is through the spacing structures. 9 claims

  17. Asymmetric fuel assembly for BWR type reactor

    International Nuclear Information System (INIS)

    A coolant turning introduction member is properly extended at coolant flow channels on the side of control rod of an inner frame for supporting the insertion of a water channel. With such a constitution, the thermal margin of the fuel rods can be made uniform over the entire region of the channel box by supplying coolants uniformly for an asymmetrical fuel assembly which can effectively suppress local peaking coefficient thereby enabling to improve performances at limit power. In addition, in the asymmetrical fuel assembly, a flow vane disposed to the outer frame plate of a spacer is increased in the size at coolant flow channels on the side of the control rod. Then, sufficient amount of coolants can surely be supplied to fuel rods at coolant flow channels on the side of the control rod. (N.H.)

  18. Fluid flow test for KMRR fuel assemblies

    International Nuclear Information System (INIS)

    Hydraulic and velocity measurment tests were carried out for the KMRR fuel assembly. Two types of the KMRR fuel assembly are consist of longitudinally finned rods. Experimental data of the pressure drops and friction factors for the KMRR fuel assemlby were produced. The measurement technique for the turbulent flow structure in subchannels using the LDV was obtained. The measurement of the experimental constant of the thermal hydraulic analysis code was investigated. The results in this study are used as the basic data for the development of an analysis code. The measurement technique acquired in this study can be applied to the KMRR thermal hydraulic commissioning test and development of the domestic KMRR fuel fabrication. (Author)

  19. Nuclear fuel assembly identification using computer vision

    International Nuclear Information System (INIS)

    A new method of identifying fuel assemblies has been developed. The method uses existing in-cell TV cameras to read the notch-coded handling sockets of Fast Flux Test Facility (FFTF) assemblies. A computer looks at the TV image, locates the notches, decodes the notch pattern, and produces the identification number. A TV camera is the only in-cell equipment required, thus avoiding complex mechanisms in the hot cell. Assemblies can be identified in any location where the handling socket is visible from the camera. Other advantages include low cost, rapid identification, low maintenance, and ease of use

  20. Fabrication of PWR fuel assembly and CANDU fuel bundle

    International Nuclear Information System (INIS)

    For the project of localization of nuclear fuel fabrication, the R and D to establish the fabrication technology of CANDU fuel bundle as well as PWR fuel assembly was carried out. The suitable boss height and the prober Beryllium coating thickness to get good brazing condition of appendage were studied in the fabrication process of CANDU fuel rod. Basic Studies on CANLUB coating method also were performed. Problems in each fabrication process step and process flow between steps were reviewed and modified. The welding conditions for top and bottom nozzles, guide tube, seal and thimble screw pin were established in the fabrication processes of PWR fuel assembly. Additionally, some researches for a part of PWR grid brazing problems are also carried out

  1. Nuclear reactor fuel assembly spacer grid

    International Nuclear Information System (INIS)

    A nuclear reactor fuel assembly spacer grid having grid straps provided with spring clips bent to widthwise encircle the grid straps and having their two ends welded together. Spring portions compressibly contact the fuel rods. The spring clips may have pairs of separated flat portions, straddling the control rod guide thimble in adjacent thimble cells so as not to interfere with the guide thimbles. The spring clips are made of a material having good radiation stress relaxation properties. (author)

  2. Corrosion of fuel assembly materials

    International Nuclear Information System (INIS)

    Corrosion of zircaloy-4 is reviewed in relation with previsions of improvement in PWRs performance: higher fuel burnup; increase coolant temperature, implying nucleate boiling on the hot clad surfaces; increase duration of the cycle due to load-follow operation. Actual knowledge on corrosion rates, based partly on laboratory tests, is insufficient to insure that external clad corrosion will not constitute a limitation to these improvements. Therefore, additional testing within representative conditions is felt necessary

  3. Nuclear reactor fuel assembly spacer grid

    International Nuclear Information System (INIS)

    A spacer grid for a nuclear fuel assembly is described. It consists of a lattice of grid plates forming multiple cells that are penetrated by fuel elements. Resilient protrusions and rigid protrusions projecting into the cells from the plates bear against the fuel element to effect proper support and spacing. Pairs of intersecting grid plates, in a longitudinally spaced relationship, cooperate with other plates to form a lattice wherein each cell contains adjacent panels having resilient protrusions arranged opposite adjacent panels having rigid protrusions. The peripheral band bounding the lattice is provided solely with rigid protrusions projecting into the peripheral cells. 8 claims

  4. RBMK fuel assemblies: Current status and perspectives

    International Nuclear Information System (INIS)

    The safety enhancement measures implemented since 1986 have led to substantial burnup reduction in the RBMK fuel assemblies and consequently to economical losses. With the purpose to compensate the losses, computer analysis and experiments were performed during the last decade. The works were aimed at the RBMK fuel charge perfection to reduce void reactivity effect and to increase fuel burnup. The paper presents principle results of the studies which are currently under implementation or are supposed to be implemented in the nearest future. (author)

  5. Appearance detection device for fuel assembly

    International Nuclear Information System (INIS)

    The prevent invention provides an appearance detection device which improves accuracy of images on a display and facilitates editing and selection of images upon detection of appearance of a reactor fuel assembly. Namely, the device of the present invention comprises (1) television cameras movable along fuel assemblies of a reactor, (2) a detection means for detecting the positions of the television cameras, (3) a convertor for converting analog image signals of the television cameras to digital image signals, (4) a memory means for sampling a predetermined portion of the images of the television camera and storing it together with the position signal obtained by the detection means and (5) a computer for selecting a plurality of images and positions from the above-mentioned means and joining them to one or a plurality of static images of the fuel assembly. At least two television cameras are disposed oppositely with each other. Then, position signals of the television cameras are designated by the stored sampling signals, and the fuel assembly at the position can be displayed quickly. It is scrolled, compressed or enlarged and formed into images. (I.S.)

  6. Transport container for unirradiated fuel assemblies

    International Nuclear Information System (INIS)

    This invention relates to a space-saving construction of a transport container for unirradiated fuel assemblies with a high security against the occurrence of a critical state under emergency conditions. The container has an internal and external part. The interspace is filled with a highly neutron absorbing material consisting of boron glass or ceramic particles coated with a cured resin film. 2 figs

  7. Measurement Protocols for Optimized Fuel Assembly Tags

    International Nuclear Information System (INIS)

    This report describes the measurement protocols for optimized tags that can be applied to standard fuel assemblies used in light water reactors. This report describes work performed by the authors at Pacific Northwest National Laboratory for NA-22 as part of research to identify specific signatures that can be developed to support counter-proliferation technologies.

  8. Calibration of spent fuel measurement assembly

    International Nuclear Information System (INIS)

    The LVR-15 research reactor (Czech Republic) had been converted from the highly enriched IRT-2M to the low enriched IRT-4M fuel. For the possibility of the independent pre-transport evaluation of IRT-2M burnup, a spectrometric system was developed. This spectrometric system consists of the fuel holder, the collimator and the portable Canberra Big MAC HPGe (High Purity Germanium) detector. In order to have well reproducible and reliable experimental data for modeling of the measurement system, calibration with the 110mAg isotope with known activity was performed. This isotope was chosen for having energies similar to isotopes measured in fuel assemblies. The 110mAg isotope was prepared by irradiating of the silver foil in LVR-15 research reactor; its activity was evaluated in the LVR-15's spectrometric laboratory. From the measured data, an efficiency curve of the spectrometric system has been determined. The experimental data were compared to the calculation results with the MCNPX model of the spectrometric system. - Highlights: • Calibration of research reactor spent fuel measurement assembly. • On-site prepared 110mAg isotope used for the measurement. • Calculated self-shielding factor for the IRT-2M fuel. • Applicable to other research reactor fuel geometries

  9. Method of straightening a bowed nuclear fuel assembly

    International Nuclear Information System (INIS)

    This patent describes a method of removing the bow in a nuclear fuel assembly, the fuel assembly having top and bottom end fittings, a plurality of longitudinally extending thimble tube members interconnecting top and bottom end fittings, at least two transverse fuel rod support grids axially spaced along the thimble tube members, and a plurality of fuel rods transversely spaced and supported by the fuel rod support grids, the method comprising the steps of securing the bottom end fitting to a predetermined location under water within the containment building of a nuclear fuel reactor and pulling vertically upward along the longitudinal axis of the nuclear fuel assembly with a force on the top end fitting so that a force of between three thousand and four thousand pounds is exerted on the nuclear fuel assembly for substantially straightening the fuel assembly and eliminating most of the compressive stresses within the fuel assembly

  10. Assembly mechanism for nuclear fuel bundles

    International Nuclear Information System (INIS)

    A description is given of a nuclear power reactor fuel bundle having tie rods fastened to a lower tie plate and passing through openings in the upper tie plate with the assembled bundle secured by rotatable locking sleeves which engage slots provided in the upper tie plate. Pressure exerted by helical springs mounted around each of the fuel rods urge the upper tie plate against the locking sleeves. The bundle may be disassembled after depressing the upper tie plate and rotating the locking sleeves to the unlocked position

  11. Assembly mechanism for nuclear fuel bundles

    International Nuclear Information System (INIS)

    The invention relates to a nuclear power reactor fuel bundle of the type wherein several rods are mounted in parallel array between two tie plates which secure the fuel rods in place and are maintained in assembled position by means of a number of tie rods secured to both of the end plates. Improved apparatus is provided for attaching the tie rods to the upper tie plate by the use of locking lugs fixed to rotatable sleeves which engage the upper tie plate. (auth)

  12. Reactor and fuel assembly design for improved fuel utilization in liquid moderated thermal reactors

    International Nuclear Information System (INIS)

    An improved reactor and fuel assembly design is disclosed wherein a light water reactor is initially run with undermoderated fuel assemblies to take advantage of increased conversion ratio, and after a suitable period of operation, the neutron spectrum for the undermoderated assemblies is shifted to lower energies to increase reactivity by withdrawing a number of fuel rods from the assemblies. The increased reactivity allows for continued operation of the modified assembly, and the fuel rods which are removed are used to construct similar assemblies which are also capable of continued operation. The improved reactor and fuel assembly design results in improved fuel utilization and neutron economy and reduced control requirements for the reactor

  13. Testing bench for spent fuel assemblies

    International Nuclear Information System (INIS)

    In the framework of a program for realizing pressurized water reactors, the D. Tech. SECS-SELECI of the French Atomic Energy Commission has transformed and adapted the shielded cell CLEMENTINE at SACLAY so that nondestructive and destructive tests could be carried out on complete 900 MW power reactor assemblies. Various operations have been carried out on both pins and assemblies since 1978. The work on the cell equipment has led to the development of a metrological test bench for examining irradiated fuels. This equipment includes a support for the assembly, a vertical girder and a displaceable tool-carrying trolley. This trolley, which moves along the Z-axis, is provided with tools for the metrological examinations associated with the displacement of the XY table, the assembly being remote controlled from a working zone situated in front of the cell. Visual examination of the four faces of the assembly is performed by displacing mirrors, which reflect the image of the object out of the cell onto a TV camera. Vertical measurements are made using optical sighting and comparing the lengths of objects with a graduated standard scale rigidly attached to the bench. Measurements made in a horizontal plane along a given Z-axis take the displacement of the sighting marks fixed to the mechanism into consideration. The displacement of this mechanism is a function of the number of pulses imparted to the system. A laser device is used to obtain the required pin spacing at various different heights in the assemblies

  14. FUEL ASSEMBLY SHAKER AND TRUCK TEST SIMULATION

    Energy Technology Data Exchange (ETDEWEB)

    Klymyshyn, Nicholas A.; Jensen, Philip J.; Sanborn, Scott E.; Hanson, Brady D.

    2014-09-25

    This study continues the modeling support of the SNL shaker table task from 2013 and includes analysis of the SNL 2014 truck test campaign. Detailed finite element models of the fuel assembly surrogate used by SNL during testing form the basis of the modeling effort. Additional analysis was performed to characterize and filter the accelerometer data collected during the SNL testing. The detailed fuel assembly finite element model was modified to improve the performance and accuracy of the original surrogate fuel assembly model in an attempt to achieve a closer agreement with the low strains measured during testing. The revised model was used to recalculate the shaker table load response from the 2013 test campaign. As it happened, the results remained comparable to the values calculated with the original fuel assembly model. From this it is concluded that the original model was suitable for the task and the improvements to the model were not able to bring the calculated strain values down to the extremely low level recorded during testing. The model needs more precision to calculate strains that are so close to zero. The truck test load case had an even lower magnitude than the shaker table case. Strain gage data from the test was compared directly to locations on the model. Truck test strains were lower than the shaker table case, but the model achieved a better relative agreement of 100-200 microstrains (or 0.0001-0.0002 mm/mm). The truck test data included a number of accelerometers at various locations on the truck bed, surrogate basket, and surrogate fuel assembly. This set of accelerometers allowed an evaluation of the dynamics of the conveyance system used in testing. It was discovered that the dynamic load transference through the conveyance has a strong frequency-range dependency. This suggests that different conveyance configurations could behave differently and transmit different magnitudes of loads to the fuel even when travelling down the same road at

  15. Method for the detection of defective nuclear fuel assemblies

    International Nuclear Information System (INIS)

    There is applied an ultrasonic transmitter on a tape carrier by means of which the ultrasonic transmitter can be guided underwater between the fuel assemblies. If a fuel assembly is defective, i.e. filled with water, the reflection coefficient at the front interface between cladding and inner space of the fuel assembly will decrease. Essential parts of the ultrasonic signal will move through the liquid and will not be reflected until the backward liquid/metal interface of the fuel assembly. This impulse echo is different from that of the gas-filled fuel assembly. (DG)

  16. Advanced membrane electrode assemblies for fuel cells

    Science.gov (United States)

    Kim, Yu Seung; Pivovar, Bryan S

    2014-02-25

    A method of preparing advanced membrane electrode assemblies (MEA) for use in fuel cells. A base polymer is selected for a base membrane. An electrode composition is selected to optimize properties exhibited by the membrane electrode assembly based on the selection of the base polymer. A property-tuning coating layer composition is selected based on compatibility with the base polymer and the electrode composition. A solvent is selected based on the interaction of the solvent with the base polymer and the property-tuning coating layer composition. The MEA is assembled by preparing the base membrane and then applying the property-tuning coating layer to form a composite membrane. Finally, a catalyst is applied to the composite membrane.

  17. Exponential experiments on PWR spent fuel assemblies

    International Nuclear Information System (INIS)

    An Exponential experiment system which is composed of a neutron detector, a signal analysis system and a neutron source, Cf-252 has been installed in order to experimentally determine the neutron effective multiplication factor for PWR spent fuel assembly. The axial background neutron flux is measured as a preliminary performance test. From the results, the spacer grid position is determined to be consistent with the design specifications within a 2.3% relative error. The induced fission neutron for four of the assemblies is also measured by scanning the neutron source, Cf-252 or the neutron detector. The exponential decay constants have been evaluated by the application of Poisson regression to the net induced fission neutron counts. It was revealed that the average exponential decay constants for the C15, J14, G23 and J44 assemblies were determined to be 0.130, 0.127, 0.125 and 0.121, respectively. (author)

  18. Assembly mechanism for nuclear fuel bundles

    International Nuclear Information System (INIS)

    A method of securing a fuel bundle to permit easy remote disassembly is described. Fuel rods are held loosely between end plates, each end of the rods fitting into holes in the end plates. At the upper end of each fuel rod there is a spring pressing against the end plate. Tie rods are used to hold the end plates together securely. The lower end of each tie rod is screwed into the lower end plate; the upper end of each tie rod is attached to the upper end plate by means of a locking assembly described in the patent. In order to remove the upper tie plate during the disassembly process, it is necessary only to depress the tie plate against the pressure of the springs surrounding the fuel rods and then to rotate each locking sleeve on the tie rods from its locked to its unlocked position. It is then possible to remove the tie plate without disassembling the locking assembly. (LL)

  19. Experimental studies on seismic behavior of PWR fuel assembly rows

    International Nuclear Information System (INIS)

    To qualify fuel assembly seismic resistance and to promote the development of new spacer grid designs, FRAMATOME (fuel division) and CEA have launched a large scale experimental program which will lead to improved models for safety analyses. The tests performed on reduced scale fuel assemblies were devoted to analyzing the assembly behavior during seismic motion. The aim of this paper is to present the main results of tests performed on rows of 5 and 13 assemblies

  20. Methodology to Access Fuel Assembly Dimension Stability on Design Stage

    International Nuclear Information System (INIS)

    The fuel assembly dimension stability (growth and distortion) is important fuel performance characteristic. Excessive growth leads to extreme axial force which can provoke significant fuel assembly distortion and, ultimately, fuel assembly component structural failure. Fuel assembly distortion may adversely affect fuel handling, RCCA insertability, plant operation, etc. Thus, fuel assembly dimension stability should be assessed during design stage to ensure that predicted growth and distortion are acceptable such no restriction on plant safety and plant operation are required. Traditionally, fuel assembly dimensional stability has not been explicitly addressed during design stage. The strong design, operating experience and lead test assemblies operation data have been required to conclude that a particular design has adequate dimensional stability. Sometimes, the new design implementation was not successful due to limited design and operating experience. A systematic methodology has been developed to explicitly address fuel assembly dimensional stability and reduce risk associated with implementing a new fuel assembly design. A detailed mechanical fuel assembly model has been developed. The model includes all fuel assembly components which contribute to fuel assembly dimensional stability. Innovative skeleton structure and explicit spacer grid models have been proposed to simulate fuel assembly component and their interaction. The required material properties have been updated to be consistent with currently available fuel material inventory. The fuel assembly component testing program has been updated to provide information required to develop the detailed fuel assembly model. The computer code SAVAN has been used to facilitate fuel assembly dimension stability assessment. SAVAN is a two-dimensional finite element analysis code developed by ENUSA. Westinghouse, ENUSA and KNF have agreed to upgrade the original SAVAN code to the new methodology. The upgraded SAVAN

  1. Fuel assembly inspection stand at NPP cooling pool

    International Nuclear Information System (INIS)

    The stand for the RBMK fuel assembly inspection under conditions of the fuel cooling pool is described. Results of experimental testing the techniques and stand equipment at the Ignalinskaya NPP are presented. The stand provides for visual control using the TSU-24M television system; eddy current examination of peripheral fuel elements; identification of failed fuel elements; measurement of diameters of peripheral fuel elements and gaps between them; registration of fuel element assembly carriage coordinates and turn angle; measurement of fuel assembly cross dimensions, bending and twist using ultrasonic transducers; drive control; carriage positioning; drive control by setting transducers in operational position

  2. Removal and replacement of fuel rods in nuclear fuel assembly

    International Nuclear Information System (INIS)

    Apparatus for replacing components of a nuclear fuel assembly stored in a pit under about 10 m. of water. The fuel assembly is secured in a container which is rotatable from the upright position to an inverted position in which the bottom nozzle is upward. The bottom nozzle plate is disconnected from the control-rod thimbles by means of a cutter for severing the welds. To guide and provide lateral support for the cutter a fixture including bushings is provided, each encircling a screw fastener and sealing the region around a screw fastener to trap the chips from the severed weld. Chips adhering to the cutter are removed by a suction tube of an eductor. (author)

  3. Fuel injection assembly for use in turbine engines and method of assembling same

    Energy Technology Data Exchange (ETDEWEB)

    Berry, Jonathan Dwight; Johnson, Thomas Edward; York, William David; Uhm, Jong Ho

    2015-12-15

    A fuel injection assembly for use in a turbine engine is provided. The fuel injection assembly includes an end cover, an endcap assembly, a fluid supply chamber, and a plurality of tube assemblies positioned at the endcap assembly. Each of the tube assemblies includes housing having a fuel plenum and a cooling fluid plenum. The cooling fluid plenum is positioned downstream from the fuel plenum and separated from the fuel plenum by an intermediate wall. The plurality of tube assemblies also include a plurality of tubes that extends through the housing. Each of the plurality of tubes is coupled in flow communication with the fluid supply chamber and a combustion chamber positioned downstream from the tube assembly. The plurality of tube assemblies further includes an aft plate at a downstream end of the cooling fluid plenum. The plate includes at least one aperture.

  4. Safety of WWER-440 fuel assembly transport

    International Nuclear Information System (INIS)

    The effective multiplication factor was calculated for a fuel assembly system in special trasport cases when flooded with water. The cases are wooden boxes lined with cadmium sheets 0.5 mm in thickness and with textile. The outer shell consists of 8 mm steel sheet. The calculation was carried out for an indefinite lattice of cases with fuel assemblies, the cases touching each other. It was assumed that the lattice was completely flooded. The wood composition was taken to be C 50%, O 44%, H 6%. The outer steel shell was replaced by pure iron. Soaking of wood was neglected as it is a long-term process. The calculation was carried out using programs THETA and CELLPAR-II, effective cross sections for the superthermal region were taken from ABBN and for the thermal region from the THETA program library. The calculated value of the effective multiplication factor shows that in no case can a chain reaction take place in the assembly lattice containing cases immersed in water and contacting each other. This calculated example represents a configuration with the highest effective multiplication factor value. The transport cases are thus satisfactory from the safety aspect. (J.F.)

  5. Interface ring for gas turbine fuel nozzle assemblies

    Science.gov (United States)

    Fox, Timothy A.; Schilp, Reinhard

    2016-03-22

    A gas turbine combustor assembly including a combustor liner and a plurality of fuel nozzle assemblies arranged in an annular array extending within the combustor liner. The fuel nozzle assemblies each include fuel nozzle body integral with a swirler assembly, and the swirler assemblies each include a bellmouth structure to turn air radially inwardly for passage into the swirler assemblies. A radially outer removed portion of each of the bellmouth structures defines a periphery diameter spaced from an inner surface of the combustor liner, and an interface ring is provided extending between the combustor liner and the removed portions of the bellmouth structures at the periphery diameter.

  6. Spent fuel assembly source term parameters

    International Nuclear Information System (INIS)

    Containment of cask contents by a transport cask is a function of the cask body, one or more closure lids, and various bolting hardware, and seals associated with the cavity closure and other containment penetrations. In addition, characteristics of cask contents that impede the ability of radionuclides to move from an origin to the external environment also provide containment. In essence, multiple release barriers exist in series in transport casks, and the magnitude of the releasable activity in the cask is considerably lower than the total activity of its contents. A source term approach accounts for the magnitude of the releasable activity available in the cask by assessing the degree of barrier resistance to release provided by material characteristics and inherent barriers that impede the release of radioactive contents. Standardized methodologies for defining the spent-fuel transport packages with specified regulations have recently been developed. An essential part of applying the source term methodology involves characterizing the response of the spent fuel under regulatory conditions of transport. Thermal and structural models of the cask and fuel are analyzed and used to predict fuel rod failure probabilities. Input to these analyses and failure evaluations cover a wide range of geometrical and material properties. An important issue in the development of these models is the sensitivity of the radioactive source term generated during transport to individual parameters such as temperature and fluence level. This paper provides a summary of sensitivity analyses concentrating on the structural response and failure predictions of the spent fuel assemblies

  7. Fuel assembly with a flute for water distribution

    International Nuclear Information System (INIS)

    The fuel assembly is arranged so that groups of fuel rods are enclosed into walls. The top end of the assembly has a peripherical distribution channel which recieves water for emergency cooling and distributes it evenly over the fuel rods. (G.B.)

  8. Optical matrix for nuclear fuel assemblies

    International Nuclear Information System (INIS)

    In order to detect the presence of fuel rods, it was selected a reflection optical transducer, which provides a measurable electrical signal when the beam at a certain distance is interrupted then there is a reflection causing a excitation to the sensor that provides a change of state at the output of transducer. This step is coupled through an operational amplifier which drives the opto coupler circuit isolating this step of the interface and a personal computer. This work presents the description of components, designs, signal coupler and opto isolater circuit, interface circuit and tutorial assemble program. (Author)

  9. The repair of irradiated fuel assemblies of RBMK-1500

    International Nuclear Information System (INIS)

    In 1988 the irradiated fuel assemblies RBMK-1500 examination stand was put into operation at unit 2 of Ignalina NPP. The examination stand was intended to research irradiated fuel assemblies. Some destructive and non-destructive examinations of irradiated fuel assemblies have been developed together with Research Institute of Atomic Reactors. Since 1991 the examination stand has been using for visual examination of irradiated fuel assemblies before loading into the reactor. Visual examination revealed some irradiated fuel assemblies with damaged heat exchange intensifying (HEI) grids. Such defects do not allow loading fuel assemblies into the reactor. In 1996 the examination stand was completed with the module allowed to repair damaged heat exchange intensifying grids. Special fuel rod safety margins were calculated for such fuel assemblies. In 1997 five irradiated fuel assemblies RBMK-1500 were repaired and loaded into the reactor of unit 2. At the moment all repaired fuel assemblies are under control in accordance with the experiment. There have not been any failures. (author)

  10. Spacer grid for nuclear fuel assembly

    International Nuclear Information System (INIS)

    A nuclear fuel assembly grid (12) having four substantially solid perimeter plates (28) forming a rigid quadangle (20) surrounding the fuel elements (36) and having stop surfaces (32) formed on the internal surfaces thereof for contacting each adjacent fuel element. A plurality of interlaced strips (24) are attached to and extend from each plate to the oppositely facing plate, forming a lattice of regularly spaced openings (16) through which the fuel elements traverse the grid. These strips are of two types, the first consisting of two perpendicular center strips (44,44') that divide the grid into four symmetric quadrants, each center strip having a spring tab (48') projecting into each opening (16') contiguous to the center strip. The second type of interlaced strip consists of the remainder of the strips (52), half of which are oriented parallel to one center strip and the other half are oriented parallel to other center strip. The second type, or interior, strips have a generally undulating bent stop surface (50) such that one bend (56) projects into each adjacent contiguous opening on the side of the interior strip facing the respective parallel center strip. Each interior strip also has a spring tab (48'') projecting into each adjacent contiguous opening on the side of the interior strip opposite the respective parallel center strip

  11. In-mast sipping for VVER-440 fuel assemblies

    International Nuclear Information System (INIS)

    The in-core sipping facility of the Jaslovske Bohunice nuclear power station installed by Siemens has proved to work satisfactorily since 1986. The process of in-mast sipping reduces the number of fuel assembly handling steps and, as a consequence, also the time spent on fuel assembly inspection and exchange in the reactor. While fuel assemblies are being handled, their cladding tubes are inspected for leakages. Inspection has indicated the operation of VVER reactors to be reliable. (orig.)

  12. Spring and stop assembly for nuclear fuel bundle

    International Nuclear Information System (INIS)

    A removable spring and stop assembly is described for use with a nuclear fuel bundle in a nuclear reactor core. The assembly includes a bolt threaded through a top section of a stop member by which the assembly (and a flow channel) is secured to the fuel bundle, the adjacent end threads of the bolt. The stop member is upset or deformed by which the bolt is captured in the assembly. (U.S.)

  13. Fuel assembly unlatching and handling gripper

    International Nuclear Information System (INIS)

    A refueling machine is provided with a latching/unlatching rod which is provided with a hexagonally configured head portion for mated engagement with a hexagonally configured socket defined within a latching/unlatching screw of a fuel assembly whereby the fuel assembly may be securely mechanically connected to the lower core support plate of the reactor internals. The latching/unlatching rod is fixedly connected to a housing which is co-axially disposed within a torque tube, secured to the lower end of a spur gear rotatably engaged with a drive spur gear through means of an idler gear whereby torque is transmitted to the torque tube. The torque tube has a square-shaped configuration in cross-section, and the housing has similarly configured flanged portions for cooperation therewith whereby rotary torque is transmitted to the housing and the latching/unlatching rod. The housing latching/unlatching rod, and torque tube are all co-axially disposed within the refuelling machine gripper tube and outer stationary mast, and a dual winch drive system is provided for independently controlling the vertical movements of the gripper tube and latching/unlatching rod respectively. (author)

  14. Nuclear fuel assembly debris resistant bottom nozzle

    International Nuclear Information System (INIS)

    A debris resistant bottom nozzle useful in a fuel assembly for a nuclear reactor is described, the bottom nozzle comprising: (a) support means adapted to rest on a lower core plate of a nuclear reactor; and (b) a plate fixed on the support means and being of a substantial solid configuration with a plurality of spaced cut-out regions therein adapted to align directly above inlet holes in the lower core plate; and (c) a plurality of open separate criss-cross structures, each of the criss-cross structures fixed to the plate and extending across one of the cut-out regions therein, the criss-cross structures defining individual openings small enough in cross-sectional size to filter out debris of damage-inducing size larger than 0.190 inch in width otherwise collects in unoccupied spaces of a lowermost grid of the fuel assembly, but large enough in size to let pass debris of nondamage-inducing size which otherwise passes through the unoccupied spaced of the lowermost grid

  15. Seismic behavior of a fuel assembly in the reactor core

    International Nuclear Information System (INIS)

    A general approach to 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 deflected shapes 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

  16. Central position detection method for fuel assembly and device therefor

    International Nuclear Information System (INIS)

    The present invention provides a method for detecting a central position of a fuel assembly by an image processing technique without influenced by a deviation of the central position of the fuel assembly depending on the accuracy for the stoppage of an underwater vehicle and rattling of fuels in a fuel basket. Namely, a characteristic amount comparing method and a linear detecting method are utilized by image processing techniques. Images are taken by a camera disposed at a predetermined position, and common characteristically shaped portions of each of the top portions of fuel assemblies are detected based on the photographed images. The central position at the top of the fuel assembly is detected based on the characteristic. In a case of a BWR fuel assembly, a channel fastener screw portion and a handle at the top of the fuel constitute the characteristic portions. The longitudinal component of the handle is detected by the linear method, and the aperture like circular portion of the channel fastener screw portion is detected by the characteristic amount comparing method. In a case of a PWR type fuel assembly, two positioning pin holes at a fuel top corner portion are detected using the characteristic amount comparing method. The central position of the fuel assembly is detected based on each of the results. (I.S.)

  17. Method and device for cleaning spent fuel assembly

    International Nuclear Information System (INIS)

    A spent fuel assembly is immersed in a liquid metal in a pot disposed below a cleaning vessel which is under the floor of an argon gas cell, and the liquid metal in the pot is heated by a heater disposed at the periphery of the cleaning vessel, and the spent fuel assembly is preheated by the heated liquid metal. Then, in a state where the spent fuel is pulled up from the pot in the cleaning vessel, heating gases are blown to the fuel assembly from above, high temperature argon gases are blown to wash out the liquid metals deposited on the spent fuel assembly. In this way, the spent fuel assembly can be heated to a predetermined preheating temperature in a short period of time. Since the amount of the liquid metal to be recovered by a vapor trap is reduced, the capacity of a storage tank exclusively used for vapor trap can be reduced. (T.M.)

  18. Zirconium fuel cladding corrosion prediction in fuel assembly operation conditions

    Energy Technology Data Exchange (ETDEWEB)

    Kritsky, V.G.; Berezina, I.G., E-mail: kritsky@givnipiet.spb.ru, E-mail: alemaskina@givnipiet.ru [Leading Inst. ' VNIPIET' , Saint Petersburg (Russian Federation)

    2010-07-01

    At present, the work to extend fuel cycles is carried out at NPP with VVER reactors. With the increase of fuel assembly burn-up to 70-100 MWd/kg U and linear power, the local coolant «nucleate boiling» is inevitable which in combination with coolant «acidification» alongside with the existing water chemistry norms will increase zirconium alloy corrosion. The rate of Zr alloy corrosion under reactor irradiation depends on temperature and heat flux through fuel cladding, coolant chemistry (concentrations of H{sub 2}O{sub 2}, OH{sup -}, O{sub 2}, hydrogen, ammonia, strong alkalis - LiOH, KOH, pH, ets.), steam content, alloy composition and some other parameters. A generalized model for calculating Zr alloys corrosion, which take into account the above-mentioned factors, was developed: K = k{sub 1}e {sup -}ΣvQ{sub 1}/R(T+ΔT) + k{sub 2} 1/1 - α + β Φ{sup n} where K{sub 1}, K{sub 2} are the coefficients depending on the water chemistry conditions and composition of Zr alloys; α is the value of steam content; Φ is a neutron flux; n is the coefficient depending on the fuel assembly type; β is the coefficient considering the impact of impurities suppressing the radiolysis, Q{sub 1} is energy contributions of alloying components and water impurities to oxide formation, v{sub i} - stehiometry coefficient. This model allows to predict a fuel cladding corrosion taking into account the alloys composition, water chemistry and fuel burn-up. The model was verified with the help of autoclave and reactor tests for commercial and modified Zr alloys. The activation energy of oxidation process is calculating on the base of ideal mixed oxide formation model. The success of such approach makes possible to propose a generalized model for calculating the corrosion of different Zr alloys in all types of water chemistry environments of old and new LWRs. (author)

  19. Assembly mechanism for nuclear fuel bundles

    International Nuclear Information System (INIS)

    In a nuclear power reactor fuel bundle having tie rods fastened to a lower tie plate and passing through openings in the upper tie plate, the assembled bundle is secured by locking lugs fixed to rotatable locking sleeves which engage the upper tie plate. Pressure exerted by helical springs mounted around each of the tie rods urge retaining lugs fixed to a retaining sleeve associated with respective tie rods into a position with respect to the locking sleeve to prevent accidental disengagement of the upper plate from the locking lugs. The bundle may be disassembled by depressing the retaining sleeves and rotating the locking lugs to the disengaged position, and then removing the upper tie plate

  20. Method of welding nuclear reactor fuel assemblies

    International Nuclear Information System (INIS)

    Disclosed is a method of welding tabs projecting outwardly from grip straps in a fuel assembly grid to a control rod guide thimble positioned in a cell in the grid including providing a weld guide having openings therein which receive dimples on the strap when the weld guide is placed in a cell adjacent to the cell containing the control rod guide thimble. The weld guide includes an opening which falls into alignment with a tab so that when a welding gun electrode is placed through the opening and into contact with a tab, the other electrode is automatically centered on its tab thus permitting accurate spot welding of the parts. To make a second spot weld on the same tab but at a point outwardly from the first spot weld, a second weld guide having an opening therein displaced a greater distance from a reference point on the weld guide, is placed in the same cell and the welding process repeated

  1. Reversible BWR fuel assembly and method of using same

    International Nuclear Information System (INIS)

    A nuclear fuel assembly is described comprising: (a) a flow channel; (b) a lower nozzle assembly structurally attached to the flow channel to form therewith an external envelope; (c) an invertible fuel bundle adapted to be inserted into the envelope, the fuel bundle comprising elongated fuel rods held in a spaced lateral array between top and bottom tie plates. Each of the top and bottom tie plates is substantially identical and has means for supporting the fuel bundle within the envelope in either of two mutually inverted vertical orientations whereby the orientation of the fuel bundle in a flow channel may be reversed during burn-up operation

  2. Separator assembly for use in spent nuclear fuel shipping cask

    Science.gov (United States)

    Bucholz, James A.

    1983-01-01

    A separator assembly for use in a spent nuclear fuel shipping cask has a honeycomb-type wall structure defining parallel cavities for holding nuclear fuel assemblies. Tubes formed of an effective neutron-absorbing material are embedded in the wall structure around each of the cavities and provide neutron flux traps when filled with water.

  3. Transient assembly of active materials fueled by a chemical reaction

    Science.gov (United States)

    Boekhoven, Job; Hendriksen, Wouter E.; Koper, Ger J. M.; Eelkema, Rienk; van Esch, Jan H.

    2015-09-01

    Fuel-driven self-assembly of actin filaments and microtubules is a key component of cellular organization. Continuous energy supply maintains these transient biomolecular assemblies far from thermodynamic equilibrium, unlike typical synthetic systems that spontaneously assemble at thermodynamic equilibrium. Here, we report the transient self-assembly of synthetic molecules into active materials, driven by the consumption of a chemical fuel. In these materials, reaction rates and fuel levels, instead of equilibrium composition, determine properties such as lifetime, stiffness, and self-regeneration capability. Fibers exhibit strongly nonlinear behavior including stochastic collapse and simultaneous growth and shrinkage, reminiscent of microtubule dynamics.

  4. Precharacterization Report for Instrumented Fuel Assembly (IFA)-527

    Energy Technology Data Exchange (ETDEWEB)

    Cunningham, M. E.; Bradley, E. R.; Daniel, J. L.; Davis, N. C.; Lanning, D. D.; Williford, R. E.

    1981-07-01

    This report is a resource document covering the rationale, design, fabrication, and preirradiation characterization of instrumented fuel assembly (IFA)-527. This assembly is being irradiated in the Halden Boiling Water Reactor (HBWR) in Norway as part of the Experimental Support and Development of Single-Rod Fuel Codes Program conducted by Pacific Northwest laboratory (PNL) and sponsored by the Fuel Behavior Research Branch of the U.S. Nuclear Regulatory Commission (NRC). Data from this assembly will be used to better understand light water reactor (LWR) fuel behavior under normal operating conditions.

  5. Dynamic modelling of PWR fuel assembly for seismic behaviour

    International Nuclear Information System (INIS)

    Vibration and snap back tests have shown that the behaviour of PWR fuel assemblies was non linear : the fuel assembly eigenfrequencies decrease with the excitation level or with the motion amplitude, which was supposed to be due to the slippage of the fuel rods through the grids. Up to now the fuel assembly models were linear and composed by one beam alone representing both the guide thimbles and the fuel rods or by two beams (one for the guide thimbles and one for the fuel rods). The stiffness of such models' were adjusted to fit with the measured eigenfrequency corresponding to a given amplitude. The aim of this paper is to identify the influence of the slippage between grids and fuel rods on the dynamic behaviour of the fuel assembly. For that purpose a non linear fuel assembly model is proposed representing explicitly the slippage phenomenon and is applied to the reduced scale fuel assemblies which have been tested in the framework of a collaboration between FRAMATOME and CEA-DMT. Comparisons between calculations and experiments will be presented and the limitation of this model will be also discussed

  6. The calculation of the YALINA BOOSTER zero power sub critical assembly driven by external neutron sources: Brazillian contribution

    Science.gov (United States)

    Carluccio, Thiago; Rossi, Pedro Carlos Russo; Maiorino, José Rubens

    2011-08-01

    The YALINA-Booster is an experimental zero power Accelerator Driven Reactor (ADS), which consists of a sub-critical assemby driven by external neutron sources. It has a fast spectrum booster zone in the center, surrounded by a thermal one. The sub-critical core is driven by external neutron sources. Several experiments have been proposed in the framework of IAEA Coordinated Reserch Project (CRP) on ADS. This work shows results obtained by IPEN modelling and simulating experiments proposed at CRP, using the MCNP code. The comparison among our results, the experimental one and the results obtained by other participants is being done by CRP coordinators. This coolaborative work has an important role in the qualification and improvement of calculational methodologies.

  7. RBMK-1500 fuel assemblies repair experience at Ignalina NPP

    International Nuclear Information System (INIS)

    The Ignalina nuclear power plant (INPP) is located in the north-east of Lithuania, closer to the borders with Belarus and Latvia. There are 2 units at INPP, each of which is equipped with RBMK-1500 reactor. The RBMK-1500 is a graphite moderated, channel-type, boiling water reactor. Its design thermal power is 4800 MW. However, for safety reasons, these reactors are currently running at reduced power of maximum 4200 MW. The RBMK-1500 reactor is the most advanced version of RBMK design and the RBMK-1500 fuel assembly has advanced version too. The fuel assembly contains two fuel bundles. Each bundle has 18 fuel rods arranged within two concentric rings in a central carried rod. The lower bundle of the fuel assembly is provided with an end grid and ten spacing grids. The top bundle has additionally 18 specifically design spacers, which act as turbulence enhances to improve the heat transfer characteristics. The hoop has 12 inclined grooves from which a steam and water mixture gets additional turbulence. In 1988 the irradiated fuel assemblies RBMK-1500 examination stand was put into operation at unit 2 of Ignalina NPP. The examination stand was intended to research irradiated fuel assemblies. Some destructive and non-destructive examinations of irradiated fuel assemblies have been developed together with Research Institute of Atomic Reactors. Since 1991 the examination stand has been using for visual examination of irradiated fuel assemblies before loading into the reactor. Visual examination apparented some irradiated fuel assemblies with damaged heat exchange intensifying (HEI) grids. Such defects do not allow to load a fuel assemblies into the reactor. In 1996 the examination stand was completed with the module allowed to repair damaged heat exchange intensifying grids. Special fuel rod safety margins were calculated for such fuel assemblies. Till now it have been repaired 15 and loaded into reactor 9 fuel assemblies with damaged hoop of HEI grid skeleton at unit 2

  8. Spacing grid for a nuclear fuel sub-assembly

    International Nuclear Information System (INIS)

    The description is given of a fuel pin spacing grid for a nuclear fuel sub-assembly. The grid includes several strips shaped to form a hexagonal honeycomb cell assembly. The cells are of one piece construction, each cell being formed from an individual strip. Every other side of the cell has an opening, the other sides being continuous. Each continuous side includes a shaped part acting as guide for a fuel pin

  9. Methodology and results of operational calculations of fuel temperature in fuel elements of the BN-600 reactor fuel assemblies

    International Nuclear Information System (INIS)

    The article presents methodology of peak fuel temperature determination and computational investigations of fuel temperature condition in fuel elements of fuel assemblies of various types during the BN-600 reactor operation. The effect of sodium uranate in the gap between fuel and cladding of the fuel element on the heat transfer processes is considered

  10. Implement of MOX fuel assemblies in the design of the fuel reload for a BWR

    International Nuclear Information System (INIS)

    At the present time the use of mixed oxides as nuclear fuel is a technology that has been implemented in mixed reloads of fuel for light water reactors. Due to the plutonium production in power reactors, is necessary to realize a study that presents the plutonium use like nuclear fuel. In this work a study is presented that has been carried out on the design of a fuel assembly with MOX to be proposed in the supply of a fuel reload. The fissile relationship of uranium to plutonium is presented for the design of the MOX assembly starting from plutonium recovered in the reprocessing of spent fuel and the comparison of the behavior of the infinite multiplication factor is presented and of the local power peak factor, parameters of great importance in the fuel assemblies design. The study object is a fuel assembly 10 x 10 GNF2 type for a boiling water reactor. The design of the fuel reload pattern giving fuel assemblies with MOX, so the comparison of the behavior of the stop margin for a fuel reload with UO2 and a mixed reload, implementing 12 and 16 fuel assemblies with MOX are presented. The results show that the implement of fuel assemblies with MOX in a BWR is possible, but this type of fuels creates new problems that are necessary to study with more detail. In the development of this work the calculus tools were the codes: INTREPIN-3, CASMO-4, CMSLINK and SIMULATE-3. (Author)

  11. Method of handling and/or storing a nuclear fuel assembly consisting of an elongated frame that contains fuel rods and fuel assembly designed specially for this method

    International Nuclear Information System (INIS)

    In order to assure subcriticality during handling, transport and/ore storage of nuclear reactor fuel assemblies and additional body containing a neutron absorbing material and touching beside the fuel rods is fixed to the frame of the fuel assembly. This body has a handle with an adapted coupling element mounted on a holding device for handling and/or storage of the fuel assembly. (orig./RW)

  12. The influence of fuel assembly characteristics on reactor safety

    International Nuclear Information System (INIS)

    To improve fuel utilization and nuclear plant economy, most nuclear plants of China adopt increased fuel enrichment and long cycle analysis. Core power distribution will be worse with these advanced items. Radial and axial peak increase too. This is a challenge to reactor safety. Since the fuel assembly is the most important part of a reactor core, fuel assembly characteristics affect reactor safety a lot. A few aspects of influence on reactor safety are discussed in this paper as a reference for fuel assembly design. A better fuel assembly design can increase heat exchange ability, especially in cold wall cells. The grids nearby core outlet can efficiently mix the flow of hot channel and average channel to decrease DNBR. In safety analysis, we always suppose the center of center assembly is the hot channel, but sometimes based on actual power distribution the hot channel occurs at side cell or corner cell. So the distribution of grids pressure drop coefficients can affect the minimum DNBR. A better fuel assembly design can help to spread core power distribution, decrease radial and axial peak efficiently. To spread core power distribution, different neutronic poisons are added into fuel pellet by different ways, and then the relative effects on reactor safety are different. At the same time, better fuel assembly design should leave enough margins for reactor safety to handle high burnup condition and so on. Fuel pellet and clad capabilities are getting worse versus increasing fuel burnup. This is a challenge to reactor safety, so more attentions should be paid to fuel burnup characteristics. (author)

  13. Sipping test of fuel assemblies in LVR-15 reactor

    International Nuclear Information System (INIS)

    The LVR-15 reactor is a light water research type which is situated at NRI in Rez near Prague. The poster describes the procedure and methodology used for sipping test of the fuel assemblies. These tests are designed to evaluate the leakage of fuel and fission products from the tested fuel assembly. From 1995 to 2003 there have been performed about 200 tests. Examples of results of sipping water activity measurements are presented. The values of activities of 137Cs and 134Cs are used for decision if the fuel assembly can be used in reactor core, transported to storage pool or if it is necessary to put the fuel assembly into the special protective can. The used limits of activities are discussed. (author)

  14. Optical fiber scope for inspecting fuel assembly

    International Nuclear Information System (INIS)

    Since a fiber scope has only one objective section, it has to observe a plurality of places successively. Then, if the time for the observation is long, the objective section is deteriorated by radiation rays, which causes a problem of interrupting the observation and increasing operator's radiation dose. In view of the above, one or two light guides are combined with an image guide to form one objective section, and a plurality of them are formed in parallel and gathered as a comb-like shape. A prism is put into a window of the objective section and resins are filled or a glass cover is attached, to make the objective section smooth and flat. Compared with the case of using only one objective section, it is no more necessary for successive observation, and objection can be conducted at one time. For example, if a fiber scope having nine objective sections is used for observing 8 x 8 arrangement fuel assembly, the observation time is shortened to 1/9. Since the prism, the glass cover, and the resins are used for making the window flat, cruds deposited between the optical fiber and a reflection mirror are easily removed, to obtain clear images. (N.H.)

  15. Effects of radial void distribution within fuel assembly on assembly neutronic characteristics

    International Nuclear Information System (INIS)

    The effect of radial subchannel-wise void distribution in a fuel assembly on assembly neutronic characteristics has been investigated using the assembly calculation code SRAC95 and the subchannel analysis code THERMIT2. With the iterative calculation of assembly calculation and the subchannel analysis (Method 1), subchannel-wise void fraction distribution, pin-power distribution and the infinite multiplication factor of the assembly are calculated. The results are compared with the result of the assembly calculation using uniform void distribution as input (Method 2). The calculation is performed for two assembly configurations in the present study: one is a fuel assembly that does not include a water rod (Case 1) and the other is the assembly that includes a water rod (Case 2). The differences in the infinite multiplication factor and pin-power peaking factor between the two methods are small in both cases. In typical BWR fuel assemblies that are investigated in the present study, the method that does not consider the radial subchannel-wise void fraction distribution within a fuel assembly (Method 2) is accurate enough for practical applications. (author)

  16. Burnup monitoring of VVER-440 spent fuel assemblies

    International Nuclear Information System (INIS)

    This paper reports on the results of the experiments performed on spent VVER-440 fuel assemblies at the Paks Nuclear Power Plant (NPP), Hungary. The fuel assemblies submerged in the service pit were examined by high-resolution gamma spectrometry (HRGS). The assemblies were moved to the front of a collimator tube built in the concrete wall of the pit in the reactor block at the NPP, and lifted down and up under water for scanning by the refueling machine. The HPGe detector was placed behind the collimator in an outside staircase. The measurements involved scanning of the assemblies along their length of all the 6 sides, at 5-12 measurement positions side by side. Axial and azimuthal burnup profiles were taken in this way. Assembly groups for measurements were selected according to their burnup (10–50 GWd/tU) and special positions (e. g. control assembly, neighbour of control assembly). Burnup differences were well observable between assembly sides looking towards the center of the core and opposite directions. Also, burnup profiles were different for control assemblies and normal (working) fuel assemblies. The ratio of the measured activities of Cs-134 and Cs-137 was evaluated by relative efficiency (intrinsic) calibration. Measurement uncertainty is around 3 %. Taking into account irradiation history and cooling time (i. e.the time elapsed since the discharge of the assembly out of the core), the activity ratio Cs-134/Cs-137 shows good correlation with the declared burnup.

  17. Fuel assembly for BWR-type reactor

    International Nuclear Information System (INIS)

    74 fuel rods and 2 large diameter water rods are disposed in 9 x 9 square lattice. Both upper and lower ends thereof are bundled by tie plates to constitute a fuel bundle, and the fuel bundle is surrounded by a channel box. Among eight short fuel rods, four short fuel rods are disposed to four corners on the second layer from the outermost circumference of the fuel bundle, and four short fuel rods are disposed to the center of each of the sides at the outermost circumference of the fuel bundle. Eight long fuel rods are disposed in adjacent with the short fuel rods at the outermost circumference of the fuel bundle. Eight long fuel rods are disposed to the second layer from the outermost circumference of the fuel bundle and in adjacent with the former eight long fuel rods. The long fuel rods contain burnable poisons in the fuel pellets filled in the most of upper portion than the upper end of the effective length of the short fuel rod disposed to the outermost circumference of the fuel bundle. (I.N.)

  18. Methodology of thermalhydraulic tests of fuel assemblies for WWER-1000

    International Nuclear Information System (INIS)

    At present 11 units with WWER-1000 are in operation in Ukraine. The NPPs are provided with nuclear fuel from Russia. The fuel assemblies are fabricated and delivered to Ukrainian NPPs from Russia. However the contemporary tendencies of nuclear energy development in the world assume a diversification of nuclear fuel vendors. Therefore the creation of the own nuclear fuel cycle of Ukraine is in mind in the strategy of nuclear energy development of Ukraine. As a part of the fuel assemblies fabrication process complex of the thermalhydraulic tests should be carried out to confirm design characteristics of the fuel assemblies before they are loaded in the reactor facility. The experimental basis and scientific infrastructure for the thermalhydraulic tests arrangement and realization of the programs and procedures for the core equipment examination are under consideration. (author)

  19. Framatome experience in fuel assembly repair and reconstitution

    International Nuclear Information System (INIS)

    Since 1985, FRAMATOME has build up extensive experience in the poolside replacement of fuel rods for repair or R and D purposes and the reconstitution of fuel assemblies (i.e. replacement of a damaged structure to enable reuse of the fuel rod bundle). This experience feedback enables FRAMATOME to improve in steps the technical process and the equipment used for the above operations in order to enhance their performance in terms of setup, flexibility, operating time and safety. In parallel, the fuel assembly and fuel rod designs have been modified to meet the same goals. The paper will describe: - the overall experience of FRAMATOME with UO2 fuel as well as MOX fuel; the usual technical process used for fuel replacement and the corresponding equipment set; - the usual technical process for fuel assembly reconstitution and the corresponding equipment set. This process is rather unique since it takes profit of the specific FRAMATOME fuel assembly design with removable top and bottom nozzles, so that fuel rods insertion by pulling through in the new structure is similar to what is done in the manufacturing plant; - the usual inspections done on the fuel rods and/or the fuel assembly; - the design of the new reconstitution equipment (STAR) compared with the previous one as well as their comparative performance. The final section will be a description of the alternative reconstitution process and equipment used by FRAMATOME in reactors in which the process cannot be used for several reasons such as compatibility or administrative authorization. This process involves the pushing of fuel rods into the new structure, requiring further precautions. (author)

  20. Manufacturing method of fuel assembly and channel box for the fuel assembly

    International Nuclear Information System (INIS)

    An MOX fuel assembly to be used for a BWR type reactor comprises a channel box, a great number of fuel rod bundles and a water rod. BP members incorporated with a burnable neutron absorbing poison (BP) are buried in the vicinity of corners of four sides of the channel box in the longitudinal direction. The channel box is formed by fitting the BP members in concaves formed in the longitudinal direction of zircaloy plates, laminating other zircaloy plates and welding the seams. Then, hot rolling, cold rolling and annealing are conducted to form them into a single plate. Integrated two single plates after bending treatment are abutted and welded, and heat-treatment is applied to complete the channel box. With such a constitution, since the BP member is not brought into contact with reactor water directly, crevice corrosion or galvanic corrosion can be prevented. (I.N.)

  1. Bottom grid mounted debris trap for a fuel assembly

    International Nuclear Information System (INIS)

    This patent describes a fuel assembly for a nuclear reactor including nuclear fuel rods, each fuel rod in the fuel assembly having a cladding tube and a lower end plug attached to the tube, at least a bottom grid supporting each and everyone of the fuel rods in an organized array and disposed in spaced relationship above the lower end plugs of the fuel rods. A bottom nozzle is disposed in spaced relationship below the bottom grid and is disposed below the lower end plugs of the fuel rods. A coolant flows upwardly through the bottom nozzle and to the bottom grid. A trap is included for catching debris carried by the flowing coolant to substantially prevent the same from reaching the bottom grid. The debris trap comprises: a fuel rod nonsupport structure disposed completely across the entire expanse of the fuel assembly and axially between the bottom nozzle and the bottom grid and generally aligned with the lower end plugs of the fuel rods. The structure forms hollow cells each being open at opposite ends and defining a central cavity which receives one of the fuel rod lower end plugs in nonsupporting and noncontacting relationship while providing for passage of coolant flow therethrough from the bottom nozzle to the bottom grid. Each of the fuel rod lower end plugs extends into a respective hollow cell of the structure

  2. The Booster

    CERN Multimedia

    1972-01-01

    Where the beams from the Booster's four rings begin to recombine, before transfer to the PS. On the left are dipoles for vertical steering, and on the right is the tank containing two septum magnets which form the first combining element.

  3. Stress analysis of screws in the fuel channel fastener assembly

    International Nuclear Information System (INIS)

    The function of fuel channel fastener assembly is to keep enough clearance between fuel channels, allowing the insertion of control rod and fixing the channel on the fuel bundle. The assembly device is not safety related component, however, in case of the screw breaking, it may cause loose parts, which might adversely affect the normal operation of inserting and pulling fuel assemblies, and/or the movement of the control rods. In this paper, the possible loading conditions applied to the fuel channel fastener assembly are considered to analyze the stress state in screw. In order to assess the improper positioning of fuel channel, explicit finite element procedures is employed to simulate the complex contact/impact behaviors occurring between the fastener assembly and the neighboring fuel channel or the fuel rack, in which the effects of dynamic impact on the screw and initial contact speed are the main concern. The analysis results reveal that the reduced neck close to the screw head has the highest stress. If the external loads drive the stress up to the yielding limit, crack initiation will occur on the screw neck and thereby, under the tensile loadings and reactor core environment, initiating intergranular stress corrosion cracking (IGSCC) on the screw

  4. Fuel assembly assessment from CVD image analysis: A feasibility study

    International Nuclear Information System (INIS)

    The Swedish Nuclear Inspectorate commissioned a feasibility study of automatic assessment of fuel assemblies from images obtained with the digital Cerenkov viewing device currently in development. The goal is to assist the IAEA inspectors in evaluating the fuel since they typically have only a few seconds to inspect an assembly. We report results here in two main areas: Investigation of basic image processing and recognition techniques needed to enhance the images and find the assembly in the image; Study of the properties of the distributions of light from the assemblies to determine whether they provide unique signatures for different burn-up and cooling times for real fuel or indicate presence of non-fuel. 8 refs, 27 figs

  5. Transport of fresh MOX fuel assemblies for MONJU initial core

    International Nuclear Information System (INIS)

    Transport of fresh MOX fuel assemblies for prototype FBR MONJU initial core started in July 1992 and ended in March 1994. As much as 205 fresh MOX fuel assemblies were transported in nine transport missions. The packaging for fuel assemblies, which has shielding and shock absorbing material inside, meets IAEA regulatory requirements for type B(U) packaging. Moreover, this packaging design features such advanced technologies as high-performance neutron shielding material and automatic hold-down mechanism for fuel assemblies. Every effort was paid to execute safe transport in conjunction with the cooperation of every competent organization. No transport accident occurred during the transport and all the transport missions were successfully completed on schedule. (author)

  6. The development of flow test technology for PWR fuel assembly

    International Nuclear Information System (INIS)

    KAERI has an extensive program to develope PWR fuel assembly. In relation to the program, development of flow test technology is needed to evaluate the thermal hydraulic compactibility and mechanical integrity of domestically fabricated nuclear fuels. A high-pressure and high-temperature flow test facility was designed to test domestically fabricated fuel assembly. The test section of the facility has capacity of a 6x6 full length PWR fuel assembly. A flow test rig was designed and installed at Cold Test Loop to carry out model experiments with 5x5 rod assembly under atmosphere pressure to get information about the characteristics of pressure loss of spacer grids and velocity distribution in the subchannels. LDV measuring technology was established using TSI's Laser Dopper Velocimeter 9100-3 System

  7. The Model of Temperature Dynamics of Pulsed Fuel Assembly

    CERN Document Server

    Bondarchenko, E A; Popov, A K

    2002-01-01

    Heat exchange process differential equations are considered for a subcritical fuel assembly with an injector. The equations are obtained by means of the use of the Hermit polynomial. The model is created for modelling of temperature transitional processes. The parameters and dynamics are estimated for hypothetical fuel assembly consisting of real mountings: the powerful proton accelerator and the reactor IBR-2 core at its subcritica l state.

  8. Parallel processing of neutron transport in fuel assembly calculation

    International Nuclear Information System (INIS)

    Group constants, which are used for reactor analyses by nodal method, are generated by fuel assembly calculations based on the neutron transport theory, since one or a quarter of the fuel assembly corresponds to a unit mesh in the current nodal calculation. The group constant calculation for a fuel assembly is performed through spectrum calculations, a two-dimensional fuel assembly calculation, and depletion calculations. The purpose of this study is to develop a parallel algorithm to be used in a parallel processor for the fuel assembly calculation and the depletion calculations of the group constant generation. A serial program, which solves the neutron integral transport equation using the transmission probability method and the linear depletion equation, was prepared and verified by a benchmark calculation. Small changes from the serial program was enough to parallelize the depletion calculation which has inherent parallel characteristics. In the fuel assembly calculation, however, efficient parallelization is not simple and easy because of the many coupling parameters in the calculation and data communications among CPU's. In this study, the group distribution method is introduced for the parallel processing of the fuel assembly calculation to minimize the data communications. The parallel processing was performed on Quadputer with 4 CPU's operating in NURAD Lab. at KAIST. Efficiencies of 54.3 % and 78.0 % were obtained in the fuel assembly calculation and depletion calculation, respectively, which lead to the overall speedup of about 2.5. As a result, it is concluded that the computing time consumed for the group constant generation can be easily reduced by parallel processing on the parallel computer with small size CPU's

  9. Removal and replacement of fuel rods in nuclear fuel assembly

    International Nuclear Information System (INIS)

    To remove the bottom nozzle of a nuclear fuel assembly, the nozzle plate must be disconnected from the control-rod thimbles. For nozzles whose control-rod thimbles are connected to the nozzle plate by screw fasteners having lock pins welded to the nozzle plate, a cutter for severing the welds is provided. The cutter is rotated by a motor at the work position through a long floating shaft. A long feed shaft operated by a thumb nut at the work position feeds the floating shaft and cutter downwardly through the weld. The bushings extend from a bushing plate, each encircling a screw fastener. Each bushing has a yieldable sleeve for sealing the region around a screw fastener to trap the chips from the severed weld. The cutter is indexed from weld to weld by indexing plates. To remove chips adhering to the cutter, the suction tube of a suction-pump-operated eductor is inserted in the auxiliary hole and the cutter is inserted in the bushing and chips are removed by suction. By inserting the suction tube into the bushings which seal the regions around the screw fasteners and enabling the eductor, the captured chips may be removed. Once the welds are severed the screw-fasteners may be unscrewed and removed by the eductor. The bottom nozzle may then be removed

  10. Fuel assembly bottom nozzle with integral debris trap

    International Nuclear Information System (INIS)

    A fuel assembly is described for a nuclear reactor including nuclear fuel rods, at least one grid supporting the fuel rods in an organized array, and at least one guide thimble supporting the grid, an improved bottom nozzle disposed adjacent and below the grid, supporting the guide thimble and adapted to allow flow of liquid coolant into the fuel assembly, the improved bottom nozzle comprising: (a) means spaced below the grid and a lower end of the fuel rods and supporting the guide thimble and allowing flow of coolant into the fuel assembly; (b) means mounted about the supporting means and extending toward but spaced from the grid and lower end of the fuel rods so as to define an open region between the supporting means and the grid and lower end of the fuel rods; and (c) a trap disposed within the open region and on the supporting means, the trap being adapted for passage of the guide thimble through to the supporting means and flow of the coolant for capturing and retaining debris carried by the flowing coolant within the trap to substantially prevent entry of debris into the fuel assembly

  11. Fuel services progress in visual examination and measurements on fuel assemblies and associated core components

    International Nuclear Information System (INIS)

    AREVA NP Fuel Services have many years of experience in visual examination and measurements on fuel assemblies and associated core components by using state of the art cameras and measuring technologies. The used techniques allow the surface and dimensional characterization of materials and shapes by visual examination. Based on these techniques measurements without contact to the measuring object, under water and with adequate accuracy are possible. New enhanced and sophisticated technologies for fuel services are such as: - Endoscopy at fuel assemblies; - Photogrammetry for measuring the deformation of a fuel assembly or fuel channel; - Shielded color cameras for use under water and close inspection of a fuel assembly. I. Endoscopy at fuel assemblies: Post irradiation programs requires, if manifestations are given, that spacer cells shall be inspected to check soundness of spacer springs. The inspection enforces the extraction of one or several fuel rods at reactor site in order to allow the accessibility of the spacer cells. AREVA NP Fuel Services developed a proportioned endoscope, which permits examination of fuel assembly spacers. The endoscope is designed to be inserted top down into a spacer cell. A 'short endoscope' enables the inspection of the upper spacers; a longer one, which is under development, will enable the inspection of all the spacers of a fuel assembly. II. Photogrammetry for measuring the deformation of a fuel assembly or fuel channel: Manual picture analysis methods for measuring parts or the whole fuel assembly are used at AREVA NP Fuel Services for years. Now, research is done to get a computer assist photogrammetry system for analyzing the pictures. The system consist of a waterproofed HD digital camera which is connected with a computer for remote control of the camera as well picture analyzing and a long handle tool to run the camera into the pool. On the computer an especial software for analyzing pictures by photogrammetry is

  12. Scratch preventing method of assembling nuclear fuel bundles, and the assembly

    International Nuclear Information System (INIS)

    This patent describes a method of assembling a bundle of nuclear fuel elements for service in a nuclear reactor. It comprises a group of fuel rod elements each arranged in a space apart, parallel array and thus secured by each element traversing through a series of spacing units positioned at intervals along the length of the grouped fuel rod elements and having openings for receiving the fuel rod elements traversing therethrough, consisting essentially of the steps of: providing a scratch resisting, temporary protective barrier consisting of a water soluble coating of sodium silicate covering the outer surface of the fuel rod elements, then assembling the fuel bundle by passing each of the fuel rod elements through the openings of a series of spacing units positioned at intervals to fit together an adjoined composite fuel bundle assembly of a spaced apart parallel array of the fuel rod elements secured with spacing units, and removing the scratch resisting, temporary protective barrier consisting of water soluble coating of sodium silicate from the assembled fuel bundle with hot water

  13. Immunity booster

    International Nuclear Information System (INIS)

    The immunity booster is, according to its patent description, microbiologically pure water with an D/(D+H) isotopic concentration of 100 ppm, with physical-chemical characteristics similar to those of distilled water. It is obtained by sterilization of a mixture of deuterium depleted water, with a 25 ppm isotopic concentration, with distilled water in a volume ratio of 4:6. Unlike natural immunity boosters (bacterial agents as Bacillus Chalmette-Guerin, Corynebacterium parvum; lipopolysaccharides; human immunoglobulin) or synthetical products (levamysol; isoprinosyne with immunostimulating action), which cause hypersensitivity and shocks, thrill, fever, sickness and the immunity complex disease, the water of 100 ppm D/(D + H) isotopic concentration is a toxicity free product. The testing for immune reaction of the immunity booster led to the following results: - an increase of cell action capacity in the first immunity shielding stage (macrophages), as evidenced by stimulation of a number of essential characterizing parameters, as well as of the phagocytosis capacity, bactericide capacity, and opsonic capacity of serum; - an increase of the number of leucocyte particularly of the granulocyte in peripheral blood, produced especially when medullar toxic agents like caryolysine are used; - it hinders the effect of lowering the number of erythrocytes in peripheral blood produced by experimentally induced chronic inflammation; - an increase of nonspecific immunity defence capacity against specific bacterial aggression of both Gram-positive bacteria (Streptococcus pneumoniae558) and of the Gram-negative ones (Klebsiella pneumoniae 507); - an increase of immunity - stimulating activity (proinflamatory), like that of levamisole as evidenced by the test of stimulation of experimentally induced inflammation by means of carrageenan. The following advantages of the immunity booster are stressed: - it is toxicity free and side effect free; - can be orally administrated as food

  14. Inert matrix fuel assembly as an option for the Laguna Verde NPP fuel reloads

    International Nuclear Information System (INIS)

    The availability of large amounts of reactor and weapons grade plutonium in the world shows the necessity of anticipating situations for the use and disposition of it. Because Light Water Reactor (LWRs) prevail on the stage of electric energy generation by nuclear power, it is important to take into account the potential of these reactors to reduce the plutonium inventory. Several studies performed in Pressurized Water Reactors (PWRs) show that reactor and weapons grade plutonium can effectively be burned in these reactors, in assemblies with fertile-free fuel, and maintaining reactivity control and other safety issues at least comparable to those related to the standard fuel normally used. The Instituto Nacional de Investigaciones Nucleares, currently carries out research on diverse alternatives to use Inert Matrix Fuel (IMF) as an option to fuel reloads for the two BWR/5 Units at the Laguna Verde Nuclear Power Plant. This work presents first the neutronic analysis of a fuel assembly conceptual design, which contains a combination of plutonium oxide (in an inert matrix) fuel rods, uranium oxide fuel rods, and uranium oxide with gadolinia fuel rods. Then, simulations for three different fuel assembly reload options were performed for Unit 1. Results of reactor operation from the different reload options are presented. The results obtained with reload fuel using inert matrix fuel assemblies observe a decrease in the length of operation cycle in the plant. However, the mass of uranium used is minor to require for make all fuel assemblies. (author)

  15. Post DNB heat transfer experiments for PWR fuel assemblies

    International Nuclear Information System (INIS)

    Nuclear Power Engineering Corporation (NUPEC) and Mitsubishi performed heat transfer experiments on post DNB (departure from nucleate boiling) for the pressurized water reactor (PWR) fuel assemblies under the sponsorship of the Japanese Ministry of Economy, Trade and Industry (METI) as one of a series of fuel assembly verification tests. Based on the obtained experimental data, a new evaluation model for the fuel rod heat transfer behavior after DNB was developed. A large safety margin, which had remained in the present thermal-hydraulic design that did not allow DNB, was confirmed by applying the developed model to the PWR plant safety analysis. (author)

  16. Tomographic imaging of severely disrupted fuel assemblies tested in TREAT

    International Nuclear Information System (INIS)

    A series of CT codes is under development in the Reactor Analysis and Safety Division of Argonne National Laboratory for use as a post-test examination tool to analyze segments of the final fuel-bundle configuration of TREAT tests. This paper presents the results of CT analysis for fuel assemblies using neutron radiography. Fuel relocation following overpower transients in the TREAT reactor is examined for sections of the assemblies, and results are compared to metallographic sections. Further improvements are expected to increase the use and reliability of CT analysis as a standard post-test examination tool

  17. Guideline for design requirement on KALIMER driver fuel assembly duct

    International Nuclear Information System (INIS)

    This document describes design requirements which are needs for designing the driver fuel assembly duct of the KALIMER as design guidance. The driver fuel assembly duct of the KALIMER consists of fuel rods, mounting rail, nosepiece, duct with pad, handling socket with pad. Fuel rod consists of top end plug, bottom end plug with solid ferritic-martensitic steel rod and key way, fuel slug, cladding, and wire wrap. In the assembly, the rods are in a triangular pitch array, and the rod bundle attaches to the nosepiece with mounting rails. The bottom end of the assembly duct is formed by a long nosepiece which provides the lower restraint function and the coolant inlet. It contains functional requirements, performance and operational requirements, interfacing systems requirements, core restraint and interface requirements, design limits and strength requirements, structural requirements, environmental requirements, reliability and safety requirements, standard and codes, QA programs, and other requirements. The design requirements are intended to be used for the design of the driver fuel assembly duct of the KALIMER. (author). 16 refs., 4 figs

  18. ABB. CASE's GUARDIANTM Debris Resistant Fuel Assembly Design

    International Nuclear Information System (INIS)

    ABB CE's experience, that 72% of all recent fuel-rod failures are caused by debris fretting, is typical. In response to this problem, ABB Combustion Engineering began supplying in the late 1980s fuel assemblies with a variety of debris resistant features, including both long-end caps and small flow holes. Now ABB CAE has developed an advanced debris resistant design concept, GUARDIANTM, which has the advantage of capturing and retaining more debris than other designs, while displacing less plenum or active fuel volume than the long end-cap design. GUARDIANTM design features have now been implemented into four different assembly designs. ABB CASE's GUARDIANTM fuel assembly is an advanced debris-resistant design which has both superior filtering performance and uniquely, excellent debris retention, Retention effectively removes the debris from circulation in the coolant so that it is not able to threaten the fuel again. GUARDIANTM features have been incorporated into four ABB. CAE fuel assembly designs. These assemblies are all fully compatible with the NSLS, and full-batch operation with GUARDIANTM began in 1992. The number of plants of both CAE and non-CAE design which accept GUARDIANTM for debris protection is expected to grow significantly during the next few years

  19. Preliminary thermal hydraulic analysis of hyper fuel assembly using Matra

    International Nuclear Information System (INIS)

    Sub-channel analysis of HYPER fuel assembly was performed using MATRA which is a subchannel analysis code developed by KAERI based on COBRA-IV-I. The MATRA code was considered for comparison between codes and assessing the capability of overcoming the limitation of the SLTHEN code used in the previous works. Two types of single fuel assembly, i.e., average assembly and hot assembly were considered for the present work. The predicted peak cladding temperatures of the average and hot assemblies were 536,2 C and 653,8 C, respectively with the reference design parameters. The comparison of results obtained by two codes shows that there is a good agreement for the predicted thermal hydraulic behaviour. It is judged that MATRA as well as SLTHEN is a very useful tool for thermal hydraulic design of the HYPER core and MATRA can be used to make up for the limitation of SLTHEN. (author)

  20. Effect of Heterogeneity of JSFR Fuel Assemblies to Power Distribution

    International Nuclear Information System (INIS)

    The Japanese sodium-cooled fast reactor JSFR is an oxide fueled system rated at 1,500 MWe. The core is composed of large fuel assemblies with an inner duct for each assembly. Thus, the assembly heterogeneity is rather strong. The purpose of the present paper is to make clear the effect of the heterogeneity to assembly and core characteristics, especially to power distribution. The inner duct is located at one corner of a hexagonal assembly, and the effect of the location has been investigated. We have compared the power distribution when the inner duct is always located near the core center and/or far from the core center. The power at the core center increased and decreased by ~10%, respectively compared to the case when the inner duct is randomly located. Thus, the location has important effect to power distribution. (author)

  1. Nuclear Fuel Assembly Assessment Project and Image Categorization

    International Nuclear Information System (INIS)

    A project has been underway to add digital imaging and processing to the inspection of nuclear fuel by the International Atomic Energy Agency. The ultimate goals are to provide the inspector not only with the advantages of Ccd imaging, such as high sensitivity and digital image enhancements, but also with an intelligent agent that can analyze the images and provide useful information about the fuel assemblies in real time. The project is still in the early stages and several interesting sub-projects have been inspired. Here we give first a review of the work on the fuel assembly image analysis and then give a brief status report on one of these sub-projects that concerns automatic categorization of fuel assembly images. The technique could be of benefit to the general challenge of image categorization

  2. Nuclear Fuel Assembly Assessment Project and Image Categorization

    Energy Technology Data Exchange (ETDEWEB)

    Lindsey, C.S.; Lindblad, T.; Waldemark, K. [Royal Inst. of Tech., Stockholm (Sweden); Hildingsson, Lars [Swedish Nuclear Power Inspectorate, Stockholm (Sweden)

    1998-07-01

    A project has been underway to add digital imaging and processing to the inspection of nuclear fuel by the International Atomic Energy Agency. The ultimate goals are to provide the inspector not only with the advantages of Ccd imaging, such as high sensitivity and digital image enhancements, but also with an intelligent agent that can analyze the images and provide useful information about the fuel assemblies in real time. The project is still in the early stages and several interesting sub-projects have been inspired. Here we give first a review of the work on the fuel assembly image analysis and then give a brief status report on one of these sub-projects that concerns automatic categorization of fuel assembly images. The technique could be of benefit to the general challenge of image categorization

  3. Statistical methods in the mechanical design of fuel assemblies

    International Nuclear Information System (INIS)

    The mechanical design of a fuel assembly is still being mainly performed in a de terministic way. This conservative approach is however not suitable to provide a realistic quantification of the design margins with respect to licensing criter ia for more and more demanding operating conditions (power upgrades, burnup increase,..). This quantification can be provided by statistical methods utilizing all available information (e.g. from manufacturing, experience feedback etc.) of the topic under consideration. During optimization e.g. of the holddown system certain objectives in the mechanical design of a fuel assembly (FA) can contradict each other, such as sufficient holddown forces enough to prevent fuel assembly lift-off and reducing the holddown forces to minimize axial loads on the fuel assembly structure to ensure no negative effect on the control rod movement.By u sing a statistical method the fuel assembly design can be optimized much better with respect to these objectives than it would be possible based on a deterministic approach. This leads to a more realistic assessment and safer way of operating fuel assemblies. Statistical models are defined on the one hand by the quanti le that has to be maintained concerning the design limit requirements (e.g. one FA quantile) and on the other hand by the confidence level which has to be met. Using the above example of the holddown force, a feasible quantile can be define d based on the requirement that less than one fuel assembly (quantile > 192/19 3 [%] = 99.5 %) in the core violates the holddown force limit w ith a confidence of 95%. (orig.)

  4. Steam vent tube for BWR fuel assembly

    International Nuclear Information System (INIS)

    This patent describes an improvement in a fuel bundle for a boiling water reactor having: vertically aligned spaced apart fuel rods for forming a fuel rod group within the fuel bundle for generation of a fission reaction in the presence of water moderator, a lower tie plate for admitting water moderator through the lower tie plate to the interstitial volume between the fuel rods and supporting the vertically aligned and spaced apart fuel rods, an upper tie plate for permitting water and steam to be discharged from the top of the fuel bundle and maintaining the vertically aligned and spaced apart fuel rods in upstanding spaced apart side-by-side relation, a surrounding fuel channel for confining moderator flow along a path over the fuel rods and from the lower tie plate to the upper tie plate. The improvement comprises: a least one steam vent tube overlying at least one of the part length rods; means supporting the stem vent tube in the volume overlying the part length rod, the steam vent tube being supported in the volume of the fuel bundle between the end of the part length rod and the upper tie plate; the steam vent tube defining an opening disposed to the end of the part length rod for the receipt of steam moderator within the void overlying the part length rod; the steam vent tube further defining an opening disposed to the upper tie plant and away from the end of the part length rod for the discharge of steam moderator from the fuel bundle

  5. Development of quality assurance methods for low enriched fuel assemblies

    International Nuclear Information System (INIS)

    As the Reduced Enrichment for Research and Test Reactors (RERTR) fuel development program has furthered the technology of low enriched uranium fuels, much effort has been expended to specify requirements, perform appropriate inspections, and to qualify experimental fuel plates and assemblies for irradiation. A great deal of consideration has been given to generate examinations and criteria that are both applicable to the unique fuel types being developed and consistent with industry practices for inspecting plate-type reactor fuel. Recent developments in quality assurance (QA) methodologies have given a heightened confidence in satisfactory fuel plate performance. At the same time, recommendations are given to further develop a system suitable for the testing and acceptance of production fuel elements containing low enriched uranium fuels. (author)

  6. The development of flow test technology for PWR fuel assemblies

    International Nuclear Information System (INIS)

    The objective of this project is to design and construct a high temperature and pressure flow test facility and to develop flow test technology for the evaluation of PWR fuel performance. For the nuclear fuel safety aspect it is of importance to evaluate the thermalhydraulic compatibility and mechanical integrity of a newly designed fuel through the design verification test. The PWR-Hot Test Loop facility is under construction to be used to perform a pressure drop test, a lift force test and a fretting corrosion test of a fullsize PWR fuel assembly at reactor operating conditions. This facility was designed to be used to produce the hydraulic parameters of the existing PWR fuel assemblies(14x14FA, 16x16FA, 17x17FA) and to verify a design of advanced fuel assemblies (KAFA-I and KAFA-II) developed by KAERI. The PWR-Cold Test Loop facility with the 5x5 Rod Bundles in the test section was designed and installed to carry out the flow distribution study by means of Laser Doppler Velocimeter. The LDV techniques have been developed and used to measure the flow velocity and turbulent intensity for evaluating mixing effects of a newly designed spacer grid with and without mixing vanes, cross flow between the fuel assemblies and a turbulent model. (Author)

  7. Thermomechanical evaluation of the fuel assemblies fabricated in the ININ

    International Nuclear Information System (INIS)

    The pilot plant of fuel production of the National Institute of Nuclear Research (ININ) provided to the Laguna Verde Nuclear Power Plant (CNLV) four fuel assemblies type GE9B. The fuel irradiation was carried out in the unit 1 of the CNLV during four operation cycles, highlighting the fact that in their third cycle the four assemblies were placed in the center of the reactor core. In the Nuclear Systems Department (DSN) of the ININ it has been carried out studies to evaluate their neutron performance and to be able to determine the exposure levels of this fuels. Its also outlines the necessity to carry out a study of the thermomechanical behavior of the fuel rods that compose the assemblies, through computational codes that simulate their performance so much thermal as mechanical. For such purpose has been developing in the DSN the FETMA code, together with the codes that compose the system Fuel Management System (FMS), which evaluates the thermomechanical performance of fuel elements. In this work were used the FETMA and FEMAXI codes (developed by JAERI) to study the thermomechanical performance of the fuel elements manufactured in the ININ. (Author)

  8. Irradiation of the experimental fuel assemblies with uranium-plutonium fuel in the BN-600 reactor

    International Nuclear Information System (INIS)

    Design features of experimental fuel assemblies (EFA) with uranium-plutonium mixed oxide fuel specific aspects of their arrangement within the BN-600 reactor core, conditions and basic results of EFA with the fuel mentioned in the BN-600 reactor are described

  9. Results of VVER-440 fuel assembly head benchmark

    International Nuclear Information System (INIS)

    In the WWER-440/213 type reactors, the core outlet temperature field is monitored with in-core thermocouples, which are installed above 210 fuel assemblies. These measured temperatures are used in determination of the fuel assembly powers and they have important role in the reactor power limitation. For these reasons, correct interpretation of the thermocouple signals is an important question. In order to interpret the signals in correct way, knowledge of the coolant mixing in the assembly heads is necessary. Computational Fluid Dynamics codes and experiments can help to understand better these mixing processes and they can provide information which can support the more adequate interpretation of the thermocouple signals. This benchmark deals with the 3D Computational Fluid Dynamics modeling of the coolant mixing in the heads of the profiled fuel assemblies with 12,2 mm rod pitch. Two assemblies of the twenty third cycle of the Paks NPPs Unit 3 are investigated. One of them has symmetrical pin power profile and another possesses inclined profile. In this benchmark, the same fuel assemblies are investigated by the participants thus the results calculated with different codes and models can be compared with each other. Aims of benchmark was comparison of participants results with each other and with in-core measurement data of the Paks NPP in order to test the different Computational Fluid Dynamics codes and applied Computational Fluid Dynamics models. This paper contains OKB 'GIDROPRESSs' results of Computational Fluid Dynamics calculations this benchmark. Results are:-In-core thermocouple signals above the selected assemblies;-Deviations between the in- ore thermocouple signals and the outlet average coolant temperatures of the assemblies;-Axial velocity and temperature profiles along three diameters at the level of the thermocouple;- Axial velocity and temperature distributions in the cross section at the level of the thermocouple;-Axial velocity and temperature

  10. Surface harmonics method for burnup calculations of VVER-1000 fuel assemblies with uranium and MOX fuel

    International Nuclear Information System (INIS)

    Development of the SUHAM-U code for burnup calculations of VVER-1000 fuel assemblies with uranium and MOX fuel is described. Developed SUHAM-U code has capacity to calculate burnup in each fuel or poison zone of each cell of VVER-1000 fuel assembly. In so doing Surface Harmonics Method is used for calculation of the detail neutron spectra in fuel assembly at separated burnup values. Verification of SUHAM-U code by burnup calculations of VVER-1000 fuel assemblies with uranium and MOX fuel has been carried out. Comparisons were carried out with calculations by UNK and RECOL codes. UNK code uses the first collisions probabilities method for solution of the neutron transport equation and RECOL code uses Monte-Carlo method with point-wise continues energy presentation of cross-sections. The main conclusion of all comparisons is the SUHAM-U code calculates the fuel burnup of VVER-1000 fuel assemblies with uranium and MOX fuel with enough high accuracy. Time expenditures are adduced. (authors)

  11. Surface harmonics method for burnup calculations of VVER-1000 fuel assemblies with uranium and MOX fuel

    Energy Technology Data Exchange (ETDEWEB)

    Boyarinov, V. F.; Davidenko, V. D.; Polismakov, A. A.; Tsibulsky, V. F. [Russian Research Center Kurchatov Inst., Nuclear Reactor Inst., 123182, Moscow (Russian Federation)

    2006-07-01

    Development of the SUHAM-U code for burnup calculations of VVER-1000 fuel assemblies with uranium and MOX fuel is described. Developed SUHAM-U code has capacity to calculate burnup in each fuel or poison zone of each cell of VVER-1000 fuel assembly. In so doing Surface Harmonics Method is used for calculation of the detail neutron spectra in fuel assembly at separated burnup values. Verification of SUHAM-U code by burnup calculations of VVER-1000 fuel assemblies with uranium and MOX fuel has been carried out. Comparisons were carried out with calculations by UNK and RECOL codes. UNK code uses the first collisions probabilities method for solution of the neutron transport equation and RECOL code uses Monte-Carlo method with point-wise continues energy presentation of cross-sections. The main conclusion of all comparisons is the SUHAM-U code calculates the fuel burnup of VVER-1000 fuel assemblies with uranium and MOX fuel with enough high accuracy. Time expenditures are adduced. (authors)

  12. Methods of RECORD, an LWR fuel assembly burnup code

    International Nuclear Information System (INIS)

    The RECORD computer code is a detailed rector physics code for performing efficient LWR fuel assembly calculations, taking into account most of the features found in BWR and PWR fuel designs. The code calculates neutron spectrum, reaction rates and reactivity as a function of fuel burnup, and it generates the few-group data required for use in full scale core simulation and fuel management calculations. The report describes the methods of the RECORD computer code and the basis for fundamental models selected, and gives a review of code qualifications against measured data. (Auth. /RF)

  13. Falling the fuel assembly in core mesh of reactor

    International Nuclear Information System (INIS)

    Accident reflecting drop of a fuel assembly (FA) in core mesh during the overload operations in the INP AS RUz research reactor is observed. Calculations and analysis of the accident situation were carried out for the reactor cores formed from fully high enriched IRT-3M type fuel (36% enrichment on '235U), the first mixed core consisting from 16 IRT-3M and 4 IRT-4M with low enriched fuel (19.7% enrichment on 235U), and the core fully formed from low enriched fuel. (authors)

  14. Hydrogen storage and integrated fuel cell assembly

    Science.gov (United States)

    Gross, Karl J.

    2010-08-24

    Hydrogen is stored in materials that absorb and desorb hydrogen with temperature dependent rates. A housing is provided that allows for the storage of one or more types of hydrogen-storage materials in close thermal proximity to a fuel cell stack. This arrangement, which includes alternating fuel cell stack and hydrogen-storage units, allows for close thermal matching of the hydrogen storage material and the fuel cell stack. Also, the present invention allows for tailoring of the hydrogen delivery by mixing different materials in one unit. Thermal insulation alternatively allows for a highly efficient unit. Individual power modules including one fuel cell stack surrounded by a pair of hydrogen-storage units allows for distribution of power throughout a vehicle or other electric power consuming devices.

  15. Method of testing fuel assemblies for nuclear reactors

    International Nuclear Information System (INIS)

    The stresses occurring in the fuel assemblies are simulated by power excursions. For this purpose the fuel assembly is placed in the neutron field of a test reactor and for a short time can be exposed to the much higher neutron field of a pulsed reactor. One possibility of design provides for the test and the pulsed reactor lying one above the other, separated by a neutron absorber and penetrated by a common irradiation channel. The fuel assembly then is to be moved from the position in the test reactor to the position in the pulsed reactor. The other possibility is to make the irradiation duct pass along the gap between both reactors and, by means of a tube-shaped absorber, open one or the other irradiation field. (DG)

  16. Atlas SOHO Booster and Centaur Erection

    Science.gov (United States)

    1995-01-01

    The launch vehicle for the Solar Heliospheric Observatory (SOHO) mission is a two stage Atlas-IIAS (Atlas/Centaur). The Atlas, consists of a solid rocket booster stage powered by four Thiokol Castor IVA solid rocket boosters (SRB) and a core vehicle stage (booster and sustainer) powered by Rocketdyne MA-5A liquid propellant engines (RP-1 fuel and liquid oxygen). The multiple firing Centaur is powered by two Pratt and Whitney (RL10A-4) liquid hydrogen and liquid oxygen engines with extendible nozzles. This video shows the erection of the Atlas booster and transportation (to 36-B launching pad) and erection of the Centaur.

  17. CFD Analysis for a Fuel Assembly of GRR-1

    International Nuclear Information System (INIS)

    The thermal-hydraulic analysis was conducted on the research reactor core for improvement on the primary cooling system of GRR(Greece Research Reactor)-1. In order to design a primary cooling system, key data were provided by the thermal-hydraulic analysis. The COOLOD code was employed to carry out the thermal-hydraulic analysis, but it was for one-dimensional calculation and single channel analysis. It can't reproduce the three-dimensional flow in complex geometries. Although pressure drop through the fuel assembly was one of the most important values to design the primary cooling system, there was no data of it from an experiment or an estimation. It should be certain that the flow distribution between coolant channels was even, since all coolant channels of a plate type fuel assembly were completely separated from each other. However, those can be obtained by conducting an experiment, a quite long time and financial resources contribute to preventing an experiment. Regarding these, the CFD (Computational Fluid Dynamics) method was a very useful alternative to reach a solution to these problems. The CFD method provide reliable and useful predictions instead of experiments due to its applicability to complex shapes which were as real as possible. This is a summary report of CFD analysis for a plate type fuel assembly of GRR-1. In this study, flow distribution between each coolant channel of the fuel assembly was predicted. In order to estimate the pressure drop through the fuel assembly, many calculations were done for various flow rate conditions. A correlation between pressure drop to flow rate was yielded from those calculation results. Temperature distribution was estimated on the fuel plates of assembly at normal operation, and was compared with the prediction results obtained by the COOLOD code. Finally, it was predicted whether or not the uncovered core can be maintained under the core melting point only by air cooling of natural circulation, when the loss

  18. DISSOLUTION OF IRRADIATED MURR FUEL ASSEMBLIES

    Energy Technology Data Exchange (ETDEWEB)

    Kyser, E.

    2010-06-17

    A literature survey on the dissolution of spent nuclear fuel from the University of Missouri Research Reactor (MURR) has been performed. This survey encompassed both internal and external literature sources for the dissolution of aluminum-clad uranium alloy fuels. The most limiting aspect of dissolution in the current facility configuration involves issues related to the control of the flammability of the off-gas from this process. The primary conclusion of this work is that based on past dissolution of this fuel in H-Canyon, four bundles of this fuel (initial charge) may be safely dissolved in a nitric acid flowsheet catalyzed with 0.002 M mercuric nitrate using a 40 scfm purge to control off-gas flammability. The initial charge may be followed by a second charge of up to five bundles to the same dissolver batch depending on volume and concentration constraints. The safety of this flowsheet relies on composite lower flammability limits (LFL) estimated from prior literature, pilot-scale work on the dissolution of site fuels, and the proposed processing flowsheet. Equipment modifications or improved LFL data offer the potential for improved processing rates. The fuel charging sequence, as well as the acid and catalyst concentrations, will control the dissolution rate during the initial portion of the cycle. These parameters directly impact the hydrogen and off-gas generation and, along with the purge flowrate determine the number of bundles that may be charged. The calculation approach within provides Engineering a means to determine optimal charging patterns. Downstream processing of this material should be similar to that of recent processing of site fuels requiring only minor adjustments of the existing flowsheet parameters.

  19. Measuring device for effective multiplication factors of a fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Ueda, Makoto

    1988-11-14

    Purpose: To measure the effective multiplication factor of a fuel assembly without using an external neutron source. Constitution: Neutron absorbers disposed on the surface of a fuel assembly incorporating a spontaneous neutron source so as to put the surface of the assembly therebetween is moved. As the neutron absorber, a cadmium plate is most suitable, but boron, gadolinium or disprosium may also be used. Neutron counting rate phi upon setting the distance between the neutron absorbers and the surface of the fuel assembly to greater than 2 cm and neutron counting rate phi' upon setting it to less than 2 cm are measured by neutron detectors. The effective multiplication factor of the fuel assembly is calculated based on the results of both of the measurements according to the following equation: K = (A(phi/phi')-1)/(AB(phi/phi'-1)) According to this method, exchange for the external neutron source is no more required, and the maintenance is easier and working efficiency is higher as compared with the prior art. Further, since phi/phi' can be determined in one identical detector in a short period of time, the measuring error can be reduced. (Horiuchi, T.).

  20. A spent fuel assemblies monitoring device by nondestructive analysis 'PYTHON'

    International Nuclear Information System (INIS)

    The monitoring of spent fuel assemblies (16 x 16 UOX) in KWG-reactor pool with the use of non-destructive methods (total Gamma and neutron counting) allow the control of average burn-up and the extremity burn-up. The measurements allow a safety-criticality control before loading the fuel assemblies into the transport casks. A device called PYTHON has been tested and qualified in France. This paper presents a description of the industrial PYTHON device and the results of the measurements. (orig.)

  1. Nuclear imaging of the fuel assembly in ignition experiments

    International Nuclear Information System (INIS)

    First results from the analysis of neutron image data collected on implosions of cryogenically layered deuterium-tritium capsules during the 2011-2012 National Ignition Campaign are reported. The data span a variety of experimental designs aimed at increasing the stagnation pressure of the central hotspot and areal density of the surrounding fuel assembly. Images of neutrons produced by deuterium–tritium fusion reactions in the hotspot are presented, as well as images of neutrons that scatter in the surrounding dense fuel assembly. The image data are compared with 1D and 2D model predictions, and consistency checked using other diagnostic data. The results indicate that the size of the fusing hotspot is consistent with the model predictions, as well as other imaging data, while the overall size of the fuel assembly, inferred from the scattered neutron images, is systematically smaller than models' prediction. Preliminary studies indicate these differences are consistent with a significant fraction (20%–25%) of the initial deuterium-tritium fuel mass outside the compact fuel assembly, due either to low mode mass asymmetry or high mode 3D mix effects at the ablator-ice interface

  2. Advanced PWR fuel assembly development programs in Korea

    International Nuclear Information System (INIS)

    Both KNFC and Westinghouse have continued to focus on developing products that will meet the challenge of increasing fuel duty requirements in Korea. These higher duty conditions include higher energy core designs through improved plant capacity factors, power uprate, extended fuel burnup, peaking factor increases, and more severe coolant chemistry (including high lithium concentration). Recent advanced fuel development activities in Korea include implementation of the 17x17 Robust Fuel Assembly (RFA), which is currently in operation with excellent performance in the United States and Europe, as well as the 16x16 PLUS7TM fuel assembly for use in KSNP plants. KNFC and Westinghouse are jointly developing advanced fuel that will meet future fuel duty challenges of 17x17 and 16x16 Westinghouse type plants. This paper focuses on advanced fuel assembly development programs that are underway and how these designs demonstrate improved margins under high duty plant operating conditions. In designing for these high duty conditions key design considerations for the various operational modes (i.e. power uprating, high burnup, long cycles, etc.) must be identified. These design considerations will include the traditional factors such as safety margin (DNB and LOCA), fuel rod design margin (e.g. corrosion, internal pressure, etc.) and mechanical design margins, among others. In addressing these design considerations, the fundamental approach is to provide additional design margin through materials, mechanical, and thermal performance enhancements, to assure flawless fuel performance. The foundation of all fuel designs is the product development process used to meet the demands of modern high duty operation including power uprating, high burnup, longer cycles, and high-lithium coolant chemistries. These advanced fuel assembly designs incorporate features that provide improved mechanical design margin, as well as thermal performance margin (DNB). Enhanced grid designs result in a

  3. Process for assembling a nuclear fuel element

    International Nuclear Information System (INIS)

    Before insertion into the spacers, the fuel rocks are coated with a self-hardening layer of water-soluble polyvinyl and/or polyether polymer to prevent scratches on the cladding tubes. After insertion, the protective conting is removed by means of water. (orig.)

  4. Lateral Stiffness Analysis of Fuel Assembly as Contact Condition for PGSFR

    International Nuclear Information System (INIS)

    To evaluate the fuel assembly bowing in the core, the lateral stiffness analysis is needed. In the fuel assembly, there are two load pads. One is the top load pad (TLP) and the other is above the core load pad (ACLP). These load pads supply the impact surface among the fuel assemblies. In this paper, the lateral stiffness analysis of the fuel assembly as the core contact condition will be executed using the finite element method. The lateral stiffness of a fuel assembly is established by the FE method. These analysis results will be utilized in a fuel assembly bowing analysis in the core

  5. Nuclear fuel assembly with large coolant conducting tube

    International Nuclear Information System (INIS)

    This patent describes a fuel assembly for a nuclear reactor comprising elongated fuel rods each containing a column of nuclear fuel; support means providing support positions for retaining the fuel rods in spaced array including a lower tie plate engaging the lower ends of the fuel rods; a nose piece extending from the lower tie plate and forming a coolant receiving chamber; a large diameter elongated coolant conducting tube extending upward through the assembly and occupying the space of the fuel rods, the coolant conducting tube having an opening at its lower end for receiving coolant and an opening at its upper end for discharging coolant; at least one space axially positioned intermediate between the upper and lower ends of the fuel rods for laterally supporting the fuel rods and the coolant conducting tube; and a mounting member for the large diameter coolant conducting tube. The lower end of the mounting member engaging the lower tie plate and the upper end of the mounting member is secured to the lower end of the coolant conducting tube. The mounting member has a relatively small diameter and is relatively flexible compared to the large diameter coolant conducting tube. In the event of lateral displacement of the upper end of the large diameter coolant conducting tube, excessive lateral forces on the spacer are avoided

  6. Nuclear fuel assembly with large coolant conducting tube

    Energy Technology Data Exchange (ETDEWEB)

    Nelson, H.L.; Dunlap, T.G.; Johnson, E.B.; Matzner, B.

    1987-06-23

    This patent describes a fuel assembly for a nuclear reactor comprising elongated fuel rods each containing a column of nuclear fuel; support means providing support positions for retaining the fuel rods in spaced array including a lower tie plate engaging the lower ends of the fuel rods; a nose piece extending from the lower tie plate and forming a coolant receiving chamber; a large diameter elongated coolant conducting tube extending upward through the assembly and occupying the space of the fuel rods, the coolant conducting tube having an opening at its lower end for receiving coolant and an opening at its upper end for discharging coolant; at least one space axially positioned intermediate between the upper and lower ends of the fuel rods for laterally supporting the fuel rods and the coolant conducting tube; and a mounting member for the large diameter coolant conducting tube. The lower end of the mounting member engaging the lower tie plate and the upper end of the mounting member is secured to the lower end of the coolant conducting tube. The mounting member has a relatively small diameter and is relatively flexible compared to the large diameter coolant conducting tube. In the event of lateral displacement of the upper end of the large diameter coolant conducting tube, excessive lateral forces on the spacer are avoided.

  7. SUPERCOLLIDER: Boosters

    International Nuclear Information System (INIS)

    Full text: The conventional construction contract for the Low Energy Booster (LEB) at the Superconducting Supercollider (SSC) being built in Ellis County, Texas, was awarded in December to Cajun Contractors of Dallas, Texas. The construction includes 1870 feet of cut-and-cover tunnel with a shielding berm, an equipment access corridor, and provision for emergency exits. It also includes ten surface buildings, as well as the usual infrastructure (concrete pads, roads, and utilities). In addition, 575 feet of cut and- cover tunnel for the next machine in the chain, the medium Energy Booster (MEB), is included. Beneficial occupancy of all structures is projected for mid-April next year. The ring magnets for the LEB include 96 conventional (non-superconducting) dipoles, each 2 metres long, and a total of 90 quadrupoles of the same cross-section as the dipoles but in eight different lengths between 0.55 and 0.71 meters. Prototypes of the dipoles and quadrupoles are being constructed by Stanford (SLAC) and Lawrence Berkeley Laboratory respectively. Production magnets are expected to be fabricated in collaboration with the Budker Institute for Nuclear Physics (BINP), Novosibirsk, starting this summer. BINP is also collaborating on the design and the fabrication of 12 dipole and 20 quadrupole magnets for the beam transfer and abort lines from the LEB to the Medium Energy Booster (MEB). The ring magnets for the MEB include 364 conventional dipoles (340 6.45 metres long and 24 5.75 metres long) and 230 quadrupoles (206 2.40 metres long and 12 each 2.8 and 0.54 metres long). A full-size development model of the dipole magnets is under construction at Fermilab. The Moscow Radiotechnical Institute is being considered as a collaborative source for the production of the quadrupole magnets, with production to start in summer of 1994 after approval of prototypes. The increasing Russian involvement in the SSC project was formalized in January when US Department of Energy Secretary

  8. Reassembling Procedure of the Fuel Assemblies for the Nuclear Power Ship ''Mutsu''

    International Nuclear Information System (INIS)

    Japan's first voyage utilized by nuclear power was made by the nuclear powered ship ''Mutsu'' in 1990. After a research voyage in 1992, decommissioning work of the nuclear reactor for ''Mutsu'' was started to change it from the nuclear power ship to an ordinary power ship. Thirty-four irradiated fuel assemblies of ''Mutsu'' were removed from the reactor and transported to the Reactor Fuel Examination Facility (RFEF) in Nuclear Science Research Institute (NSRI) of Japan Atomic Energy Agency (JAEA). ''Mutsu'' fuel assemblies were loaded into a hot cell of RFEF using the roof gate as the top loading procedure. After the reliability confirmation tests, fuel assemblies were reassembled for reprocessing. To perform the reliability confirmation tests and reassembling, new devices were developed and installed in the hot cells, ''Fuel assembly transportation device'' for transporting the fuel assemblies between the hot cells, ''Upper nozzle cutting device'' for removing the upper nozzle from the fuel assembly, ''Fuel rod drawing device'' for drawing a fuel rod from the fuel assembly and so on. Thirty-four fuel assemblies were reassembled as six PWR type fuel assemblies in order to adjust the acceptable specifications of the reprocessing plant in JAEA: the shape of fuel assembly is the same as the PWR type commercial reactor fuel and the average enrichment of uranium in the assembly is under 4.0%. This paper reports the reassembling techniques of the ''Mutsu'' irradiated fuel assemblies for reprocessing. (author)

  9. Manipulator for fuel assemblies in a spent fuel pool, especially for a LMFBR

    International Nuclear Information System (INIS)

    The spent fuel manipulator has - a travelling crane moving longitudinally: - a carriage moving on the travelling crane in a direction perpendicular to its motion so that the carriage is positioned over each assembly, - a telescopic rod carried by the carriage and terminating in a vertically mobile grapple, - a tubular shielded hood on the carriage extending downwards to house the rod, grapple and fuel assembly and maintaining a biologically acceptable level of radiation above the surface of the pool

  10. Criticality assessment of fuel assemblies with missing fuel rods - an intractable problem?

    International Nuclear Information System (INIS)

    In current certificates of package approval the arrangement of water and guide tubes within the array of fuel rods of a fuel assembly is specified in detail. Fuel assemblies with deviating water and guide tube arrangements or missing rods are not allowed to be loaded into the casks. The reason behind is that the reactivity of a standard fuel assembly increases if some rods are removed. For a certain number and arrangement of missing rods a maximum of reactivity is reached. Due to the missing fissile material the reactivity will decrease again if further rods are then removed. For the comprehensive assessment of the maximum of reactivity all possible configurations of fuel rods and missing rods have to be investigated. The paper describes the problem at hand in detail giving estimates for the complexity of the analysis

  11. Spacer grid for a PWR fuel assembly

    International Nuclear Information System (INIS)

    The spacer grid defines a block of square section cells each accommodating one fuel rod and is made up of interlocking flat strips welded together and made of zirconium alloy. A spring of nickel alloy is secured between each peripheral strip. The strip defining the wall of each of those cells opposite strip carries rigid bosses pressed out of the strip. The rods in those cells are gripped between bosses and spring sections. 7 figs

  12. Spacer grid for reducing bowing in a nuclear fuel assembly

    International Nuclear Information System (INIS)

    A bi-metallic spacer grid having a zircaloy perimeter strip consisting of oppositely facing, thin walled metal plates for closely surrounding the array of fuel rods. A rigid, stainless steel cross member extends between internal surfaces of the oppositely facing perimeter plates. In the preferred embodiment, the perimeter plates have cantilevered portions extending above and below the main body of the perimeter strip. The cross members interact with the enlarged portion by urging them outward relative to the perimeter strip as the fuel assembly heats up during operation. The outwardly projecting interface surfaces of each assembly mechanically interact with the interface surfaces of adjacent assemblies providing a mechanical restraint which limits bowing of the assembly. The effectiveness of the spacer grids in limiting bowing is therefore not dependent upon controlling the mechanisms responsible for causing bow. When the reactor is in a cold condition such as during refueling , the exterior dimensions of the spacer grids are the same as those of the other zircaloy grids, which assures adequate clearance for insertion and withdrawal of individual fuel assemblies

  13. ROSA-III system description for fuel assembly, 4

    International Nuclear Information System (INIS)

    The ROSA (Rig of Safety Assessment)-III System with fuel assembly No.4 and its instrumentations are described. The informations are necessary to understand and analyze the experimental data obtained from loss-of-coolant experiments (LOCEs) conducted in the ROSA-III facility. (author)

  14. Experimental study of new generation WWER-1000 fuel assemblies at JSC NCCP

    International Nuclear Information System (INIS)

    An experimental program for the study of fuel assembly thermomechanical stability has been established together with RF SSC IPPE and Russian Scientific Center Kurchatov Institute. Assembly fragments and small dummy models of fuel assembly skeletons and fuel rod bundles have been used for the tests. The test results are used for the design selection, verification of the design codes and substantiation of operating capacity of fuel assemblies with a rigid skeleton. The mechanical characteristics of units make it possible to perform fuel assembly strength and rigidity calculations, including the cases of abnormal operation. The mechanical characteristics of the skeleton and fuel rod bundle dummy models make it possible to check for the adequacy of the fuel assembly design model. The mechanical characteristics obtained during fuel rods bundle push through experiments make it possible to substantiate the fuel assembly serviceability under the conditions of fuel rods bundle and skeleton interaction

  15. Post-irradiation examination of Fugen reactor fuel assembly at reactor fuel examination facility

    International Nuclear Information System (INIS)

    Post-irradiation examination of the first assembly of a monitoring program for Heavy Water Reactor ''Fugen'' of PNC (Power Reactor and Nuclear Fuel Development Corporation) has been executed since Oct. 1983 at the Reactor Fuel Examination Facility, JAERI Tokai (Japan Atomic Energy Research Institute, Tokai Research Establishment). The fuel assembly is a cylindrical cluster, with 4,400mm length, composed of 28 rods in 3 concentric circles, 12 spring-grid spacers and the upper and lower tie plates. The fuel is plutonium-uranium mixed oxide (0.8 w/o), and the material of cladding tube is Zry-2. The average burnup of the fuel assembly is about 13,600 MWd/t. This paper describes the methods and some results on the post irradiation examination items as follows: 1. Radioactive measurement of water in transportation cask; 2. Visual inspection of the fuel assembly in dry cell, before and after removing the crud, by ultrasonic vibration method; 3. Chemical analyses and radioactive measurement of the crud materials; 4. Dimensional measurement of assembly length and rod-rod gaps, before and after removing the crud; 5. Disassembly and dimensional measurement of rod-rod gaps in the inner circles; 6. Several nondestructive testing techniques of fuel rods. (author)

  16. A new SCWR fuel assembly with two-row fuel rods between the hexagonal moderator channels

    International Nuclear Information System (INIS)

    Highlights: • We propose a SCWR fuel assembly with two-row fuel rods between the hexagonal moderator channels. • The new concept can resolve the contradiction between uniform and sufficient moderation. • Structural size and thermal–hydraulic performance are taken account of in the fuel assembly. • Larger infinite multiplication factor and smaller local power peaking factor could be obtained. • Two two-row hexagonal fuel assembly concepts are proposed for the engineering application. - Abstract: A new hexagonal fuel assembly (FA) design which has two rows of fuel rods between the hexagonal moderator channels is proposed for the thermal supercritical water cooled reactor (SCWR). The new concept is well considered for the performance of uniform moderation and sufficient moderation, and with respect to structural size and thermal–hydraulic performance. The neutron physical performance of the two-row hexagonal FA with acceptable configuration is discussed. The results show clearly that a better balance between uniform moderation and sufficient moderation can be obtained in the two-row hexagonal fuel assembly

  17. Vibration Pre-characterization of Partial Fuel Test Assembly

    International Nuclear Information System (INIS)

    To check applicability to a conventional reactor core and compatibility with a present fuel design requires hydraulic vibration testing for the annular fuel design in the form of a fuel bundle. Objective of the hydraulic vibration testing (or flow induced vibration testing) is to understand vibration behavior of an oscillating structure submerged in fluid flow and find out relationship between vibration responses of a structure and flow characteristics. Along the same line, a partial fuel test assembly (PFTA) was made in 4x4 arrays with 12 dummy annular fuel rods and 5 combination-type spacer grids of cantilever and vortex dimple spring as shown in Fig. 1. To be more focus for effects of the inner channel fluid mass on the dynamics of an annular fuel test assembly, mass-equivalent simulated lead pellets were eliminated in dummy fuel rods. A series of vibration testing using UMAP (underwater modal test equipment) in ambient and under still water were performed to identify dynamic characteristics of PFTA and evaluate the effects of test parameters and conditions. Objective of the test is to evaluate UMAP performances and prepare backup data for future response analysis of hydraulic vibration testing

  18. Nuclear fuel assembly top nozzle with improved arrangement of hold-down leaf spring assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Lawson, C.N.

    1991-06-26

    A fuel assembly has a top nozzle which includes a lower adapter plate and a plurality of guide structures which are attached to an extend along the periphery of the lower plate and upwardly therefrom. The top nozzle also includes an upper hold-down plate supported by a plurality of leaf spring assemblies. The upper plate is mounted to the guide structures for vertical slidable movement relative thereto. The leaf spring assemblies are provided in a non-peripheral arrangement relative to the periphery of the lower plate in which the assemblies are provided in a non-peripheral arrangement relative to the periphery of the lower plate in which the assemblies cross the interior of the lower plate in a diagonal fashion between adjacent ones of the guide structures. (author).

  19. Behaviour of spent fuel assemblies during extended storage

    International Nuclear Information System (INIS)

    This report is the final report of the IAEA Co-ordinated Research Programme on Behaviour of Spent Fuel Assemblies During Extended Storage (BEFAST, Phase I, 1981-86). It contains the results on wet and dry spent fuel storage technologies obtained from 11 institutes (10 countries: Austria, Canada, Czechoslovakia, Finland, German Democratic Republic, Hungary, Japan, Sweden, USA and USSR) participating in the BEFAST CRP during the time period 1981-86. Names of participating institutes and chief investigators are given. The interim spent fuel storage has been recognized as an important independent step in the nuclear fuel cycle. Due to the delay in commercial reprocessing of spent fuel in some cases it should be stored up to 30-50 years or more before reprocessing or final disposal. This programme was evaluated by all its participants and observers as very important and helpful for the nuclear community and it was decided to continue it further (1986-91) as BEFAST, Phase II

  20. Method and apparatus for dismantling and disposing of fuel assemblies

    International Nuclear Information System (INIS)

    This invention relates to apparatus and a method for dismantling, shearing, and compacting a fuel assembly frame skeleton. It uses an apparatus capable of hanging or being supported in the transfer canal of the spent fuel pit of the fuel handling building. This apparatus includes a bottom nozzle shear which is held under water to shear off the bottom nozzle and convey it to a scrap transfer bin. Then the remaining portion is brought to a skeleton compactor and shear, also held under water. The compacted skeleton is sheared into a number of smaller portions. After compacting and shearing, the individual portions are fed to the scrap transfer bin. The compacted and sheared skeleton assembly may be placed into a container that is adapted to hold four skeletons for off-site removal

  1. Determination of mixing factors for VVER-440 fuel assembly head

    International Nuclear Information System (INIS)

    CFD models have been developed for the heads of the old, the present and new type VVER-440 fuel assemblies using the experiences of former validation process. With these models, temperature distributions were investigated in typical assemblies and in-core thermocouple signals were calculated. The analyses show that coolant mixing is intensive but not-perfect in the assembly heads. Difference between the thermocouple signal and cross-sectional average temperature at measurement level depends on the assembly type. Using the models, weight factors of the rod bundle regions for the in-core thermocouple have been determined. With these factors, the thermocouple signals were estimated and results were statistically tested using the registered data of the Hungarian nuclear power plant. This test shows that deviations between measured and calculated temperatures can be significantly decreased and consequently monitoring uncertainties can be reduced with using the weight factors. (author)

  2. Determination of BWR Spent Nuclear Fuel Assembly Effective Thermal Conductivity

    International Nuclear Information System (INIS)

    The purpose of this calculation is to provide an effective thermal conductivity for use in predicting peak cladding temperatures in boiling water reactor (BWR) fuel assemblies with 7x7,8x8, and 9x9 rod arrays. The first objective of this calculation is to describe the development and application of a finite element representation that predicts peak spent nuclear fuel temperatures for BWR assemblies. The second objective is to use the discrete representation to develop a basis for determining an effective thermal conductivity (described later) for a BWR assembly with srneared/homogeneous properties and to investigate the thermal behavior of a spent fuel assembly. The scope of this calculation is limited to a steady-state two-dimensional representation of the waste package interior region. This calculation is subject to procedure AP-3.124, Calculations (Ref. 27) and guided by the applicable technical work plan (Ref. 14). While these evaluations were originally developed for the thermal analysis of conceptual waste package designs emplaced in the potential repository at Yucca Mountain, the methodology applies to storage and transportation thermal analyses as well. Note that the waste package sketch in Attachment V depicts a preliminary design, and should not be interpreted otherwise

  3. Refueling machine mounted fuel assembly inspection T.V. cameras

    International Nuclear Information System (INIS)

    This patent describes a refueling machine comprising a trolley, movable within a horizontal plane above fuel assemblies in a reactor core of a nuclear reactor facility, an outer, stationary mast fixedly mounted to the trolley and extending vertically downwardly therefrom, and an inter mast coaxially mounted within the outer mast and telescopically movable therein. A gripper assembly is fixedly secured to the lower end of the inner mast for attachment to the fuel assemblies for movement of the fuel assemblies into the outer mast and in and out of the reactor core. A basket-type framework surrounds the lower end of the stationary mast and vertically mounted television cameras are fixedly attached to the basket-type framework with their lenses oriented vertically downwardly. Light sources are fixedly attached to the basket-type framework below the television cameras and support cables are secured to the basket-type framework for moving the basket-type framework vertically relative to the stationary mast

  4. Determination of BWR Spent Nuclear Fuel Assembly Effective Thermal Conductivity

    Energy Technology Data Exchange (ETDEWEB)

    Matthew D. Hinds

    2001-10-17

    The purpose of this calculation is to provide an effective thermal conductivity for use in predicting peak cladding temperatures in boiling water reactor (BWR) fuel assemblies with 7x7,8x8, and 9x9 rod arrays. The first objective of this calculation is to describe the development and application of a finite element representation that predicts peak spent nuclear fuel temperatures for BWR assemblies. The second objective is to use the discrete representation to develop a basis for determining an effective thermal conductivity (described later) for a BWR assembly with srneared/homogeneous properties and to investigate the thermal behavior of a spent fuel assembly. The scope of this calculation is limited to a steady-state two-dimensional representation of the waste package interior region. This calculation is subject to procedure AP-3.124, Calculations (Ref. 27) and guided by the applicable technical work plan (Ref. 14). While these evaluations were originally developed for the thermal analysis of conceptual waste package designs emplaced in the potential repository at Yucca Mountain, the methodology applies to storage and transportation thermal analyses as well. Note that the waste package sketch in Attachment V depicts a preliminary design, and should not be interpreted otherwise.

  5. Calculation of the BN-600 fuel assemblies mode in a gas medium

    International Nuclear Information System (INIS)

    Potentiality of calculated modeling of temperature conditions of warming up elements of spent fuel assemblies of the BN-600 reactor during their transportation within gaseous medium is shown. The calculated modeling of spent fuel assemblies warming up in gaseous medium, their residual heat release values being different, permits substantiating and optimizing safe conditions of post-reactor handling of the fuel assemblies

  6. Design of Neptunium-bearing Fuel Assembly for Transmutation Research in CEFR

    International Nuclear Information System (INIS)

    In order to have a better understanding of irradiation performance of the fuel containing neptunium, an experimental assembly is designed for future irradiation in CEFR. There is only one fuel pin in the assembly with neptunium content of 5%. Temperature monitors and neutron fluence detectors are attached. The report presents the basic structure of the fuel pin and the assembly. (author)

  7. Orificing of water cross inlet in BWR fuel assembly

    International Nuclear Information System (INIS)

    A nuclear reactor fuel assembly is described comprising a bundle of elongated fuel rods disposed in side-by-side relationship so as to form an array of spaced fuel rods, a tubular flow channel member surrounding the fuel rods so as to direct flow of coolant/moderator fluid along the fuel rods, respective upper and lower tie plates at opposite ends of the fuel rods, and a hollow water cross having confronting side walls and a closed lower end wall at an inlet end. The water cross extends centrally through and disposed within the flow channel member so as to provide within the flow channel member separate compartments and to divide the bundle of fuel rods into mini-bundles being disposed in the respective compartments, the water cross including inlet cross flow means formed in the side walls near a lower end of the water cross above the closed end wall and near lower end portions of each of the mini-bundles of fuel rods, which inlet cross flow means provides both selected flow communication into the interior of the water cross and flow communication between the respective mini-bundles for minimizing maldistribution and equalizing flow

  8. Effect of Heterogeneity of JSFR Fuel Assemblies to Power Distribution

    International Nuclear Information System (INIS)

    Conclusion: 1) Strong heterogeneity of JSFR assemblies was successfully calculated by BACH. 2) Verification test of BACH: • Infinite assembly model; • Color set model; • Good agreement with Monte-Carlo results. 3) Core calculations 3 models for inner duct was used; inward model, outward model and homogeneous model. • keff difference between the inward and out ward model → 0.3%Δk; • ~20% effect on flux and power distributions. Therefore, we have to pay careful attention for the location of inner duct in fuel loading of JSFR

  9. The Welding Process of the Small In-pile Testing Fuel Assembly

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    The small in-pile testing fuel assembly is designed for high performance fuel assembly study. It has two parts of which are four fuel element with double layer cladding and a detect system for measurement of testing pressure and temperature. The fuel element is composed of UO2 pellets, the stainless steel cladding and end caps. The detect system is direct contact with the fuel element by electron beam welding. In the fabrication of the assembly, some special welding technologies are

  10. Vibration Characteristics of a Plate Type Fuel Assembly

    Energy Technology Data Exchange (ETDEWEB)

    Yim, J.S.; Kim, H.J.; Tahk, Y.W.; Oh, J.Y.; Lee, B.H. [Nuclear Fuel Design for Research Reactor, KAERI, 305-353 Daeduck Daero 1045, Yuseong-Ku, Taejon (Korea, Republic of)

    2011-07-01

    A flat fuel plate and a box type fuel assembly for a research reactor were modeled to be finite element meshes of the ANSYS to predict dynamic characteristics, such as natural frequencies and mode shapes. These characteristics provide the basic information about their vibrations. If the model is properly prepared, it can be used for further calculations of the dynamic behaviors under the SSE or even in the static stress calculation. With the FE Model, the natural frequencies and the mode shapes of a fuel plate and a FA were obtained in air and in water environments. The effects of fluid surrounding the fuel plate and the FA as well as the combs on the natural vibration of the FA are discussed. (author)

  11. Neutron resonance transmission analysis of reactor spent fuel assemblies

    International Nuclear Information System (INIS)

    A method called Neutron Resonance Transmission Analysis (NRTA) is under study which would use a pulsed neutron beam for nondestructive isotopic assay of a complete spent fuel assembly. Neutrons removed from the collimated beam by absorption or scattering in the resonances of the various isotopes in the spent fuel appear as dips in the neutron transmission. The method is completely insensitive to matrix materials such as oxide, fuel cladding, and other structural members. Measurements on spent fuel buttons using the NBS linac as a pulsed neutron source demonstrate a high accuracy capability for the isotopes 234235236238U, 239240241242Pu, 241Am, 243Am, and several fission products. The NRTA method offers high speed and modest operational cost, and it can be implemented with commercially available medical or radiographic γ-ray generators adapted for neutron production. (Auth.)

  12. Analysis of the sub-channel of SCWR two-row fuel assembly

    International Nuclear Information System (INIS)

    Based on the COBRA-Ⅳ code, a new sub-channel code system developed for the supercritical water cooled reactor (SCWR) fuel assembly is analyzed. In order to optimize the SCWR fuel assembly design, a sub-channel analysis of two rows SCWR fuel assembly is performed, including steady-state and transient calculation. For the steady-state calculation, several channel's parameters are selected to evaluate the thermal-hydraulic performance of the fuel assemblies. Based on the steady-state results, two transient calculations (change of fuel rod power and change of coolant flow) are carried out to estimate the dynamic behavior of the fuel assemblies. The results achieved so far indicate a good applicability of the sub-channel code for the SCWR fuel assembly analysis, which is good for the future optimization of SCWR fuel assembly design. (authors)

  13. The Technology Trend of Japanese Patent for the Nuclear Fuel Assembly Inspection

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Jai Wan; Choi, Young Soo; Lee, Nam Ho; Jeong, Kyung Min; Suh, Yong Chil; Kim, Chang Hoi; Shin, Jung Cheol

    2008-06-15

    Japanese technology patents for the nuclear fuel assembly inspection unit, from the year 1993 to the year 2006, were investigated. The fuel rods which contain fissile material are grouped together in a closely-spaced array within the fuel assembly. Various kinds of reactor including the PWR reactor are being operated in Japan. There are many kinds of nuclear fuel assemblies in Japan, and the shape and the size of these nuclear fuel assemblies are various also. As the structure of these various fuel assemblies is a regular square as the same as the Korean one, the inspection method described in Japanese technology patent can be applied to the inspection of the nuclear fuel assembly of the Korea. This report focuses on advances in VIT(visual inspection test) of nuclear fuel assembly using the state-of-the-art CCD camera system.

  14. Micro-Pocket Fission Detectors (MPFD) For Fuel Assembly Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Troy Unruh; Michael Reichenberger; Phillip Ugorowski

    2013-09-01

    Neutron sensors capable of real-time measurement of thermal flux, fast flux, and temperature in a single miniaturized probe are needed in irradiation tests required to demonstrate the performance of candidate new fuels, and cladding materials. In-core ceramic-based miniature neutron detectors or “Micro-Pocket Fission Detectors” (MPFDs) have been studied at Kansas State University (KSU). The first MPFD prototypes were tested in various neutron fields at the KSU TRIGA research reactor with successful results. Currently, a United States Department of Energy-sponsored joint KSU/Idaho National Laboratory (INL) effort is underway to develop a high-temperature, high-pressure version of the MPFD using radiation-resistant, high temperature materials, which would be capable of withstanding irradiation test conditions in high performance material and test reactors (MTRs). Ultimately, this more compact, more accurate, and longer lifetime flux sensor for critical mock-ups, existing and advanced reactor designs, high performance MTRs, and transient test reactors has the potential to lead to higher accuracy and resolution data from irradiation testing, more detailed core flux measurements and enhanced fuel assembly processing. Prior evaluations by KSU indicate that these sensors could also be used to monitor burn-up of nuclear fuel. If integrated into nuclear fuel assemblies, MPFDs offer several advantages to current spent fuel management systems.

  15. Rapid fuel drawer scanner for fast critical assembly safeguards

    International Nuclear Information System (INIS)

    An integrated scanning system incorporating highly efficient collimated neutron and high purity germanium gamma detectors with an on-line microprocessor has been developed to perform rapid inventorying of uranium and plutonium fuel drawers from fast critical assemblies. On-line least-squares fit procedures provide quantitative comparisons at a rate exceeding two drawers per minute. For plutonium-containing fuel, the neutron scan data can be related to the included 240Pu isotopic mass; individual 239Pu, 241Pu, and 241Am isotopic contents are obtained from simultaneous scans of the appropriate isolated gamma lines

  16. Device for transferring fast nuclear reactor fuel assemblies

    International Nuclear Information System (INIS)

    The description is given of a device for transferring fuel assemblies between a storage position near the reactor vessel and a position where the irradiated assemblies are evacuated and the provision of new assemblies for the reactor. This device can be dismantled and is movable as a whole for its successive use on several reactors and includes: - a platform mounted so as to rotate on a support made to rest on the structure of the reactor, the platform having at least one opening then being horizontal and mobile about a vertical axis to bring the opening successively in position with vertical wells giving access to the storage and evacuation positions of the assemblies provided in the reactor structure, - at least one hopper that can contain one assembly in a vertical position, located on the upper surface of the platform around the opening provided in it and fitted with a winch for the vertical moving of the assemblies inside the wells and the hopper, when these follow each other by rotation of the platform, - at least one connecting device carried on the platform for connecting the hopper and wells when these are in line

  17. Effect of 17 X 17 fuel assembly geometry on DNB

    International Nuclear Information System (INIS)

    A series of tests was run in which the DNB heat flux was determined in axially uniform heated rod bundles of both 17 x 17 (.374'' rod) and 14 x 14 (.422'' rod) reactor fuel assembly geometry. The purposes of this test series were (1) to assess the DNB performance of .374'' fuel rod assembly design, (2) to verify the present design DNB methods for .374'' rod geometry and (3) to obtain a body of data which will later be used to develop a new correlation. The comparison of the uniform to previously reported non-uniform axial heat flux DNB test results using the .422'' rod geometry showed that use of the non-uniform heat flux F-factor used with existing correlations brings the two sets into agreement. 10 references

  18. Numerical benchmarks for MTR fuel assemblies with burnable poison

    International Nuclear Information System (INIS)

    This work presents a preliminary version of a set of burn-up dependent numerical benchmarks of MTR fuel assemblies using burnable poisons. The numerical benchmark calculations were carried out using two different types of calculation methodologies: Monte Carlo methodology using MCNP-ORIGEN coupled codes and deterministic methodology using CONDOR collision probabilities code. The main purpose of this work is to provide a numerical benchmark for several geometries, for example number and diameter of the Cadmium wires. The numerical benchmark provides meat and Cadmium numerical density information and the geometry and material data of the calculated systems. These benchmarks provide information for the validation of MTR FA cell codes. This paper is the preliminary work of a 3 dimensional numerical benchmark for research reactors using MTR fuel assemblies with burnable poisons. A short description of the MCNP and ORIGEN coupling method and the CONDOR code are given in the present paper. (author)

  19. Russian fuel assemblies implementation experience at SUNPP-2

    International Nuclear Information System (INIS)

    The paper contains general information on switching SUNPP Unit 2 to usage of alternative design fuel assemblies. There are 3 Units in operation at SUNPP with total electrical capacity of 3000 MWt, and each of them has its own specific characteristics. The fuel of alternative design (TVS-A) manufactured by TVEL Corporation with gadolinium burnable absorber integrated into fuel elements has been operating at Unit 2 since 2005. Implementation of modified fuel is considered as design changes, and therefore it should be approved, and according to the Law of Ukraine 'On use of nuclear Energy and Radiation Safety' all necessary approval should undergo the State expertise on nuclear and radiation safety. Transient fuel cycles (containing both TVS and TVS-A) was performed by SUNPP's physicists (by means of KASKAD computer code) by use recommendations of RRC 'Kurchatov Institute' experts and was approved in according to the existing procedure. The paper also contains detailed information on the specific characteristics of the SUNPP Unit 2 reactor core, main stages in course of implementation of new fuel types, characteristics of 'mixed' fuel cycles, comparison of calculated and measured power distribution, comparison of calculated and experimental data (Authors)

  20. Process and device for fabricating nuclear fuel assembly grids

    International Nuclear Information System (INIS)

    The method for fabricating PWR fuel assembly grids consists to place the grid of which the constituent parts are held firmly in place within a frame into a sealed chamber full of inert gas. This chamber can rotate about an axis. The welding on one face at a time is carried out with a laser beam orthogonal to the axis orientation of the device. The laser source is outside of the chamber and the beam penetrates via a transparent view port

  1. Dynamics of nuclear fuel assemblies in vertical flow channels

    International Nuclear Information System (INIS)

    DYNMOD is a computer program designed to predict the dynamic behaviour of nuclear fuel assemblies in axial flow. The calculations performed by DYNMOD and the input data required by the program are described in this report. Examples of DYNMOD usage and a brief assessment of the accuracy of the dynamic model are also presented. It is intended that the report will be used as a reference manual by users of DYNMOD

  2. FIRST STEP blanket structure and fuel assembly design

    International Nuclear Information System (INIS)

    FIRST STEP (Fusion, Inertial, Reduced Requirement Systems Test for Special Nuclear Material, Tritium, and Energy Production) is an Inertial Confinement Fusion (ICF) plant designed to produce tritium, SNM, and energy using near-term technology. It is an integrated facility that will serve as a test bed for fusion power plant technology. The design of the blanket structure and blanket fuel assembly for wetted-wall FIRST STEP reactors is presented here

  3. A CFD Simulation Process for Fast Reactor Fuel Assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Kurt D. Hamman; Ray A. Berry

    2010-09-01

    A CFD modeling and simulation process for large-scale problems using an arbitrary fast reactor fuel assembly design was evaluated. Three-dimensional flow distributions of sodium for several fast reactor fuel assembly pin spacing configurations were simulated on high performance computers using commercial CFD software. This research focused on 19-pin fuel assembly “benchmark” geometry, similar in design to the Advanced Burner Test Reactor, where each pin is separated by helical wire-wrap spacers. Several two-equation turbulence models including the k–e and SST (Menter) k–? were evaluated. Considerable effort was taken to resolve the momentum boundary layer, so as to eliminate the need for wall functions and reduce computational uncertainty. High performance computers were required to generate the hybrid meshes needed to predict secondary flows created by the wire-wrap spacers; computational meshes ranging from 65 to 85 million elements were common. A general validation methodology was followed, including mesh refinement and comparison of numerical results with empirical correlations. Predictions for velocity, temperature, and pressure distribution are shown. The uncertainty of numerical models, importance of high fidelity experimental data, and the challenges associated with simulating and validating large production-type problems are presented.

  4. A CFD simulation process for fast reactor fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Hamman, Kurt D., E-mail: Kurt.Hamman@inl.go [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Berry, Ray A. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States)

    2010-09-15

    A CFD modeling and simulation process for large-scale problems using an arbitrary fast reactor fuel assembly design was evaluated. Three-dimensional flow distributions of sodium for several fast reactor fuel assembly pin spacing configurations were simulated on high performance computers using commercial CFD software. This research focused on 19-pin fuel assembly 'benchmark' geometry, similar in design to the Advanced Burner Test Reactor, where each pin is separated by helical wire-wrap spacers. Several two-equation turbulence models including the k-{epsilon} and SST (Menter) k-{omega} were evaluated. Considerable effort was taken to resolve the momentum boundary layer, so as to eliminate the need for wall functions and reduce computational uncertainty. High performance computers were required to generate the hybrid meshes needed to predict secondary flows created by the wire-wrap spacers; computational meshes ranging from 65 to 85 million elements were common. A general validation methodology was followed, including mesh refinement and comparison of numerical results with empirical correlations. Predictions for velocity, temperature, and pressure distribution are shown. The uncertainty of numerical models, importance of high fidelity experimental data, and the challenges associated with simulating and validating large production-type problems are presented.

  5. Fuel assembly simulations using LRGR-CFD and CGCFD

    International Nuclear Information System (INIS)

    In addition to the traditional fuel assembly simulation approaches using system codes, subchannel codes or porous medium approaches, as well as detailed CFD simulations to analyze single sub channels, a Low Resolution Geometry Resolving (LRGR) CFD approach and a Coarse-Grid-CFD (CGCFD) approach is taken. Both methods are based on a low resolution mesh that allows the capture of large and medium scale flow features such as recirculation zones, which cannot be reproduced by the system codes, subchannel codes and porous media approaches. The LRGR approach allows for instance fine-tuning the porous parameters which are important input for a porous medium approach. However, it should be noted that the prediction of detailed flow features such as secondary flows is not feasible. Using this approach, the consequences of flow blockages for detection possibilities and cladding temperatures can be discussed. Within the Coarse-Grid CFD approach a subgrid model (SGM) accounts for sub grid volumetric forces which are derived from validated CFD simulations. The volumetric forces take account of the non resolved physics due to the coarse mesh. The CGCFD approach with SGM can be applied to simulate complete fuel assemblies or even complete cores capturing the unique features of the complex flow induced by the fuel assembly geometry and its spacers. In such a case, grids with a very low grid resolution are employed. The current paper discusses and presents both, the CGCFD and the LRGR approaches. (author)

  6. Calibration of the TVO spent BWR reference fuel assembly

    International Nuclear Information System (INIS)

    In 1989 the Support Programmes of Finland (FSP) and Sweden (SSP) initiated a joint task to cross calibrate the burnup of the IAEA spent BWR reference fuel assembly at the TVO AFR storage facility (TVO KPA-STORE) in Finland. The reference assembly, kept separately under the IAEA seal, is used for verification measurements of spent fuel by GBUV method (SG-NDA-38). The cross calibration was performed by establishing a calibration curve, 244Cm neutron rate versus burnup, using passive neutron assay (PNA) measurements. The declared burnup of the reference assembly was compared with the burnup value deduced from the calibration curve. A calibration line was also established by using the GBUV method with the aid of high resolution gamma ray spectrometry (HRGS). Normalization between the two different facilities was performed using sealed neutron and gamma calibration sources. The results of the passive neutron assay show consistency, better than 1 %, between the declared mean burnup of the reference assembly and the burnup deduced from the calibration curve. The corresponding consistency is within +-2 % for the HRGS measurements

  7. The Dit nuclear fuel assembly physics design code

    International Nuclear Information System (INIS)

    DIT is the Combustion Engineering, Inc. (C-E) nuclear fuel assembly design code. It belongs to a class of codes, all similar in structure and strategy, which may be characterized by the spectrum and spatial calculations being performed in 2D and in a single job step for the entire assembly. The forerunner of this class of codes is the U.K.A.E.A. WIMS code, the first version of which was completed 25 years ago. The structure and strategy of assembly spectrum codes have remained remarkably similar to the original concept thus proving its usefulness. As other organizations, including C-E, have developed their own versions of the concept, many important variations have been added which significantly influence the accuracy and performance of the resulting computational tool. This paper describes and discusses those features which are unique to the DIT code and which might be of interest to the community of fuel assembly physics design code users and developers

  8. The DIT nuclear fuel assembly physics design code

    International Nuclear Information System (INIS)

    The DIT code is the Combustion Engineering, Inc. (C-E) nuclear fuel assembly design code. It belongs to a class of codes, all similar in structure and strategy, that may be characterized by the spectrum and spatial calculations being performed in two dimensions and in a single job step for the entire assembly. The forerunner of this class of codes is the United Kingdom Atomic Energy Authority WIMS code, the first version of which was completed 25 yr ago. The structure and strategy of assembly spectrum codes have remained remarkably similar to the original concept thus proving its usefulness. As other organizations, including C-E, have developed their own versions of the concept, many important variations have been added that significantly influence the accuracy and performance of the resulting computational tool. Those features, which are unique to the DIT code and which might be of interest to the community of fuel assembly physics design code users and developers, are described and discussed

  9. Structural investigation of fuel rods basing on dynamic model of heat flow phenomena in fuel assemblies

    International Nuclear Information System (INIS)

    The structural investigation of reactor materials are usually by calculations determining the working conditions of particulars elements of fuel assemblies or the hole reactor core. For this analysis the mathematical model of heat flow phenomena was proposed which enable the calculations of temperature field within the assembly. The differential equations for mass, energy and momentum of cooling medium conservation in coaxial and transversal flow direction enable the steady state and transient analysis for the cases of change in heat flow in cooling medium velocity and the pressure in the assembly. The introduced empire correlation which are completing the set of equations make possible the analysis for violent changes of cladding temperature of fuel elements for cooling medium in two-phase flow. The computer program basing on the presented model was prepared for the calculations of initial parameters necessary for beginning the cladding and fuel material structural investigations. (author)

  10. Debris-retaining trap for a fuel assembly

    International Nuclear Information System (INIS)

    A fuel assembly is described for a nuclear reactor including nuclear fuel rods, at least one grid supporting the fuel rods in an organized array, an end flowing through the end nozzle and into the fuel assembly, a trap for capturing and retaining debris carried by the flowing coolant to prevent entry of debris into the fuel assembly, the debris trap comprising: (a) a hollow enclosure disposed adjacent the end nozzle on an opposite side from the grid, the enclosure being composed of a material which is permeable to the liquid coolant but impermeable to debris carried by the coolant; (b) the hollow enclosure has upper and lower walls spaced apart and interconnected at their peripheries so as to extend across the direction of liquid coolant flow through the end nozzle and define a debris capturing the retaining chamber within the enclosure; (c) means on the hollow enclosure defining at least one opening into the chamber of the enclosure through the lower wall; (d) lower debris-retaining means located within the chamber of the hollow enclosure and surrounding the opening into the chamber, the lower debris-retaining means has a configuration which serves to retain debris carried by coolant into the chamber through the opening from exiting through the opening; and (e) upper flow-diffusing means is located within the chamber of the hollow enclosure and is spaced generally above and aligned with the lower debris-retaining means and the opening. The upper flow-diffusing means has a configuration which substantially uniformly distributes across the bottom nozzle the flow of coolant into the chamber through the opening

  11. Vibration characteristics of the KSNP fuel assembly with newly developed top and bottom end pieces

    International Nuclear Information System (INIS)

    Nuclear fuel assembly is exposed to various exciting sources such as fluid induced vibration, circulating pump, earthquake, and loss of coolant accident. To maintain its integrity under these vibratory circumstances, vibration characteristics of fuel assembly should be thoroughly understood, and should be well reflected into fuel assembly design. In this study, the fuel assembly for Korea Standard Nuclear Plants (KSNP) is modeled as a uniform beam with reactor end condition and, based on the model, the vibration characteristics of the fuel assemblies with not only conventional upper and lower end fittings but also newly developed ones are evaluated by using the frequency equation which was derived by Fourier Sine series. In the case of introducing newly developed upper and lower end fittings to the fuel assembly for KSNP, it is expected that natural frequency of the fuel assembly be lowered a little due to the boundary condition change, but the difference is negligible

  12. Optimized fuel assemblies for modern PWRs - design features and operating experience

    International Nuclear Information System (INIS)

    Reliability, cost effectiveness, increased operational flexibility and easy handling are the focal aspects of the Siemens fuel assembly generation FOCUS. In the plants of the latest S MENS PWR generation, Konvoi, the 18x18 FOCUS fuel assemblies in particular have demonstrated consistently reliable performance. A modern fuel assembly design is characterized by its modular structure consisting of indispensable and optional technical features, adapted to the needs and wishes of the customers. FOCUS (Fuel assembly with Optimized Cladding and Upgraded Structure) stands for such a fuel assembly design, marking fuel assemblies equipped with an advanced cladding material and Zircaloy spacer grids in the active area featuring mixing vanes and a hang-up resistant envelope as the essential parts. Options exist for other fuel assembly components

  13. An Enhancement of Visual Test Performance for Nuclear Fuel Assembly

    International Nuclear Information System (INIS)

    In the overhaul period of the nuclear power plant, integrity of the neutron-irradiated fuel assembly is evaluated. Nuclear regulations require that nuclear power plants meet the design, operation, and inspection requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (B and PV). Section XI of the ASME B and PV Code provides the specific requirements for inspecting the systems, structures, and components; Section V of the ASME Code provides requirements for inspection methods, including volumetric (e.g., ultrasonic testing), surface (e.g., eddy current testing), and visual testing (VT). Visual testing of neutron irradiated fuel assembly is conducted generally for a variety of purposes, for example to detect discontinuities and imperfections on the surface of fuel rods, to detect evidence of leakage from end-cap welds, and to determine the general mechanical and structural condition of one. VT is performed remotely using video camera. As the neutron-irradiated fuel assembly is a high dose-rate gamma-ray source, approximately a few kGy, radiation hardened underwater camera is used in the VT of the fuel assembly. Utilities today follow the EPRI guidelines for VT-1 tests on nuclear components (BWR Vessel and Internals Project-3 1995). The VT-1 guidelines specify which areas around a weld should be examined, how to measure the sizes of indications found, and how to test the resolving power of the visual equipment used for the test. The EPRI guidelines use two 12μm (0.0005-in.) wires or notches as a resolution calibration standard. According to the EPRI guidelines (BWRVIP-03 1995), the camera systems employed were marginally able to detect the 0.0005-inch (12-μm) diameter wire on a steel background. In the some future, it is required that the VT of nuclear fuel assembly follows the EPRI VT-1 guideline. In order to meet the VT-1 guideline, any system used in VT (ranging from the naked eye to a digital closed-circuit TV system

  14. An Enhancement of Visual Test Performance for Nuclear Fuel Assembly

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Jai Wan; Choi, Young Soo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Shin, Jung Cheol [Korea Nuclear Fuel, Daejeon (Korea, Republic of)

    2009-05-15

    In the overhaul period of the nuclear power plant, integrity of the neutron-irradiated fuel assembly is evaluated. Nuclear regulations require that nuclear power plants meet the design, operation, and inspection requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (B and PV). Section XI of the ASME B and PV Code provides the specific requirements for inspecting the systems, structures, and components; Section V of the ASME Code provides requirements for inspection methods, including volumetric (e.g., ultrasonic testing), surface (e.g., eddy current testing), and visual testing (VT). Visual testing of neutron irradiated fuel assembly is conducted generally for a variety of purposes, for example to detect discontinuities and imperfections on the surface of fuel rods, to detect evidence of leakage from end-cap welds, and to determine the general mechanical and structural condition of one. VT is performed remotely using video camera. As the neutron-irradiated fuel assembly is a high dose-rate gamma-ray source, approximately a few kGy, radiation hardened underwater camera is used in the VT of the fuel assembly. Utilities today follow the EPRI guidelines for VT-1 tests on nuclear components (BWR Vessel and Internals Project-3 1995). The VT-1 guidelines specify which areas around a weld should be examined, how to measure the sizes of indications found, and how to test the resolving power of the visual equipment used for the test. The EPRI guidelines use two 12{mu}m (0.0005-in.) wires or notches as a resolution calibration standard. According to the EPRI guidelines (BWRVIP-03 1995), the camera systems employed were marginally able to detect the 0.0005-inch (12-{mu}m) diameter wire on a steel background. In the some future, it is required that the VT of nuclear fuel assembly follows the EPRI VT-1 guideline. In order to meet the VT-1 guideline, any system used in VT (ranging from the naked eye to a digital closed-circuit TV

  15. Inter fuel-assembly thermal-hydraulics for the ELSY square open reactor core design

    International Nuclear Information System (INIS)

    The lead-cooled reactor is one of the six proposed innovative reactor types by the Generation IV International Forum (GIF). In Europe, the lead-cooled reactor design is known as the European Lead-cooled System (ELSY), which is a 600 MWe medium size fast reactor. The reference design of the ELSY core foresees square open (wrapper-less) fuel-assemblies with a staggered arrangement. In this design, the fuel rods in a fuel-assembly are separated by 3.4 mm. The gap between fuel rods of neighboring fuel-assemblies is 5.5 mm. In other words, the reference gap size between fuel-assemblies is larger than the gap between fuel rods within a fuel-assembly. This article discusses the involved inter fuel-assembly thermal-hydraulics between neighboring fuel-assemblies in the ELSY core. For this purpose as a starting point a validated Reynolds Averaged Navier Stokes (RANS)-based Computational Fluid Dynamics (CFD) approach is adopted. Moreover, bare fuel rods are considered in the present analyses that serve as a step towards inclusion of a spacer grid when its design is fixed. As the next step, the fuel-assemblies are numerically arranged with different gap sizes of 2.1 mm and 3.4 mm in order to analyze the influence of gap size on the inter fuel-assembly thermal-hydraulics. As a final step, analyses on the influence of different power levels of neighboring fuel-assemblies in the ELSY core are presented based on the reference ELSY core design. These inter-fuel assembly thermal hydraulic analyses lead to a conservative Nusselt number correlation for calculating maximum surface temperature of bare fuel rods that are located in the gap region between neighboring fuel-assemblies having different power levels. Such correlations, when implemented, will improve the applicability of system codes.

  16. Development of an ultrasonic cleaning method for fuel assemblies

    International Nuclear Information System (INIS)

    Almost all radiation buildup in light water reactors is the result of the deposition of activated corrosion and wear products in out-of-core areas. After operation, a significant quantity of corrosion and wear products is deposited on the fuel rods as crud. At refueling shutdowns, these activation products are available for removal. If they can be quickly and easily removed, buildup of radioactivity on out-of-core surfaces and individual exposure dose can be greatly reduced. After studying various physical cleaning methods (e.g., water jet and ultrasonic), the ultrasonic cleaning method was selected as the most effective for fuel assembly cleaning. The ultrasonic cleaning method is especially able to efficiently clean the fuel without removing the channel box. The removed crud in the channel box would be swept out to the filtration unit. Parameter survey tests were carried out to evaluate the optimum conditions for ultrasonic cleaning using a mock-up of a short section of fuel assembly with the channel box. The ultrasonic device used was a 600-W ultrasonic transducer operating at 26-kHz ultrasonic frequency

  17. Development of an ultrasonic cleaning method for fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Heki, H.; Komura, S.; Kato, H.; Sakai, H. (Toshiba Corp., Kawasaki City (Japan)); Hattori, T. (Tokyo Electric Power Co., Kashiwazaki-shi (Japan))

    1991-01-01

    Almost all radiation buildup in light water reactors is the result of the deposition of activated corrosion and wear products in out-of-core areas. After operation, a significant quantity of corrosion and wear products is deposited on the fuel rods as crud. At refueling shutdowns, these activation products are available for removal. If they can be quickly and easily removed, buildup of radioactivity on out-of-core surfaces and individual exposure dose can be greatly reduced. After studying various physical cleaning methods (e.g., water jet and ultrasonic), the ultrasonic cleaning method was selected as the most effective for fuel assembly cleaning. The ultrasonic cleaning method is especially able to efficiently clean the fuel without removing the channel box. The removed crud in the channel box would be swept out to the filtration unit. Parameter survey tests were carried out to evaluate the optimum conditions for ultrasonic cleaning using a mock-up of a short section of fuel assembly with the channel box. The ultrasonic device used was a 600-W ultrasonic transducer operating at 26-kHz ultrasonic frequency.

  18. Tomographic techniques for safeguards measurements of nuclear fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Lundqvist Saleh, Tobias

    2007-10-15

    Nuclear power is currently experiencing increased interest over the world. New nuclear reactors are being built and techniques for taking care of the nuclear waste are being developed. This development puts new demands and standards to safeguards, i.e. the international efforts for ensuring the non-proliferation of nuclear weapons. New measuring techniques and devices are continuously being developed for enhancing the ability to detect diversion of fissile material. In this thesis, tomographic techniques for application in safeguards are presented. Tomographic techniques can non-destructively provide information of the inner parts of an object and may thus be used to control that no material is missing from a nuclear fuel assembly. When using the tomographic technique described in this thesis, the radiation field around a fuel assembly is first recorded. In a second step, the internal source distribution is mathematically reconstructed based on the recorded data. In this work, a procedure for tomographic safeguards measurements is suggested and the design of a tomographic measuring device is presented. Two reconstruction algorithms have been specially developed and evaluated for the application on nuclear fuel; one algorithm for image reconstruction and one for reconstructing conclusive data on the individual fuel rod level. The combined use of the two algorithms is suggested. The applicability for detecting individual removed or replaced rods has been demonstrated, based on experimental data

  19. Tomographic techniques for safeguards measurements of nuclear fuel assemblies

    International Nuclear Information System (INIS)

    Nuclear power is currently experiencing increased interest over the world. New nuclear reactors are being built and techniques for taking care of the nuclear waste are being developed. This development puts new demands and standards to safeguards, i.e. the international efforts for ensuring the non-proliferation of nuclear weapons. New measuring techniques and devices are continuously being developed for enhancing the ability to detect diversion of fissile material. In this thesis, tomographic techniques for application in safeguards are presented. Tomographic techniques can non-destructively provide information of the inner parts of an object and may thus be used to control that no material is missing from a nuclear fuel assembly. When using the tomographic technique described in this thesis, the radiation field around a fuel assembly is first recorded. In a second step, the internal source distribution is mathematically reconstructed based on the recorded data. In this work, a procedure for tomographic safeguards measurements is suggested and the design of a tomographic measuring device is presented. Two reconstruction algorithms have been specially developed and evaluated for the application on nuclear fuel; one algorithm for image reconstruction and one for reconstructing conclusive data on the individual fuel rod level. The combined use of the two algorithms is suggested. The applicability for detecting individual removed or replaced rods has been demonstrated, based on experimental data

  20. Thermal Hydraulic Analysis Of Thorium-Based Annular Fuel Assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Han, Kyu Hyun [Korea Institute of Nuclear Safety, 19, Guseong-dong, Yuseong-gu, Daejeon, 305-338 (Korea, Republic of)

    2008-07-01

    Thermal hydraulic characteristics of thorium-based fuel assemblies loaded with annular seed pins have been analyzed using AMAP combined with MATRA, and compared with those of the existing thorium-based assemblies. MATRA and AMAP showed good agreements for the pressure drops at the internal sub-channels. The pressure drop generally increased in the cases of the assemblies loaded with annular seed pins due to the larger wetted perimeter, but an exception existed. In the inner sub-channels of the seed pins, mass fluxes were high due to the grid form losses in the outer sub-channels. About 43% of the heat generated from the seed pin flowed into the inner sub-channel and the rest into the outer sub-channel, which implies the inner to outer wall heat flux ratio was approximately 1.2. The maximum temperatures of the annular seed pins were slightly above 500 deg. C. The MDNBRs of the assemblies loaded with annular seed pins were higher than those of the existing assemblies. Due to the fact that inter-channel mixing cannot occur in the inner sub-channels, temperatures and enthalpies were higher in the inner sub-channels. (author)

  1. Thermal Hydraulic Analysis Of Thorium-Based Annular Fuel Assemblies

    International Nuclear Information System (INIS)

    Thermal hydraulic characteristics of thorium-based fuel assemblies loaded with annular seed pins have been analyzed using AMAP combined with MATRA, and compared with those of the existing thorium-based assemblies. MATRA and AMAP showed good agreements for the pressure drops at the internal sub-channels. The pressure drop generally increased in the cases of the assemblies loaded with annular seed pins due to the larger wetted perimeter, but an exception existed. In the inner sub-channels of the seed pins, mass fluxes were high due to the grid form losses in the outer sub-channels. About 43% of the heat generated from the seed pin flowed into the inner sub-channel and the rest into the outer sub-channel, which implies the inner to outer wall heat flux ratio was approximately 1.2. The maximum temperatures of the annular seed pins were slightly above 500 deg. C. The MDNBRs of the assemblies loaded with annular seed pins were higher than those of the existing assemblies. Due to the fact that inter-channel mixing cannot occur in the inner sub-channels, temperatures and enthalpies were higher in the inner sub-channels. (author)

  2. Impact of the moderation ratio over the performance of different BWR fuel assemblies

    International Nuclear Information System (INIS)

    Highlights: • Performance of fuel assemblies is assessed using moderation ratio as a merit figure. • Burnup changes moderation ratio operating conditions for the fuel assembly. • After 30 GWd/MT fuel assemblies are working in the over-moderated region. • For an 18-month cycle discharge fuel assembly burnup is over 40 GWd/MT. • For extended cycles or up-rate conditions use of these FA could result in reduced margins to meet safety constraints. - Abstract: Fuel assembly design plays a very important role in the reactor core performance. A fuel assembly has to be designed to achieve safe and efficient performance during its active life inside the nuclear reactor core. Fuel assemblies are designed to be under-moderated to produce a negative moderator temperature coefficient under all operational circumstances. This study assesses the behavior of the infinite multiplication factor (k∞) as a function of the moderation ratio and its dependence on the burnup, for several BWR fuel assemblies. The results show that the moderation ratio at which the fuel assembly transitions from under-moderated to over-moderated changes through the life of the fuel assembly (i.e. with burnup). This study shows that the fuel assembly designs considered, operate in the over-moderated region for burnups over 30 GWd/MT. In a typical 18-month cycle BWR core, even though the fraction of fuel assemblies with burnups over 40 GWd/MT can reach about 50% at the end of cycle the core still meets safety constraints. However, if the fuel assembly designs used were to experience burnups over 45 GWd/MT, the fraction of fuel assemblies operating in the over-moderated region would be high enough to compromise the safety performance of the core

  3. Operation experience of WWER-440 fuel assemblies and measures to increase fuel reliability

    International Nuclear Information System (INIS)

    The paper presents technical data for the fuel cycles used in 14 WWER-440 reactors of B-213 type situated outside CIS-territory on the basis of the 2001 operational results. The paper reflects the dynamics of average and maximum fuel burnup as well as information on the annual rate of the leaking fuel rods for the above reactor group identified during the 1997-2001 discharge period. As an example of work performed by RIAR in 2001 the paper brings forth the PIE-results of a leaking WWER-440 fuel assemblies (FAs). It is reported that the reason behind the leaking and failed fuel rods of the FA was interaction with a foreign object being in the coolant flow. The paper describes the measures taken by the NPPs together with the Supplier (JSC TVEL) and Manufacturer (JSC MSZ) to enhance the fuel operational safety. (author)

  4. Irradiation-induced dimensional changes of fuel compacts and graphite sleeves of OGL-1 fuel assemblies

    International Nuclear Information System (INIS)

    Experimental data are summarized on irradiation-induced dimensional changes of fuel compacts and graphite sleeves of the first to ninth OGL-1 fuel assemblies. The range of fast-neutron fluence is up to 4 x 1024 n/m2 (E > 0.18 MeV); and that of irradiation temperature is 900 - 1400 deg C for fuel compacts and 800 - 1050 deg C for graphite sleeves. The dimensional change of the fuel compacts was shrinkage under these test conditions, and the shrinkage fraction increased almost linearly with fast-neutron fluence. The shrinkage fraction of the fuel compacts was larger by 20 % in the axial direction than in the radial direction. Influence of the irradiation temperature on the dimensional-change behavior of the fuel compacts was not observed clearly; presumably the influence was hidden by scatter of the data because of low level of the fast-neutron fluence and the resultant small dimensional changes. (author)

  5. Changes to Irradiation Conditions of VVER-1000 Surveillance Specimens Resulting from Fuel Assemblies with Greater Fuel Height

    OpenAIRE

    Panferov Pavel; Kochkin Viacheslav; Erak Dmitry; Makhotin Denis; Reshetnikov Alexandr; Timofeev Andrey

    2016-01-01

    The goal of the work was to obtain experimental data on the influence of newtype fuel assemblies with higher fuel rods on the irradiation conditions of surveillance specimens installed on the baffe of VVER-1000. For this purpose, two surveillance sets with container assemblies of the same design irradiated in reactors with different fuel assemblies in the core were investigated. Measurements of neutron dosimeters from these sets and retrospective measurements of 54Mn activity accumulated in ea...

  6. Nuclear material attractiveness: an assessment of used-fuel assemblies

    International Nuclear Information System (INIS)

    This paper examines the material attractiveness of used-fuel assemblies in a hypothetical scenario in which terrorists steal one or more assemblies in order to use the special nuclear materials (SNM) within an assembly in a nuclear explosive device. For assessing material attractiveness, this paper uses the Figure of Merit (FOM) that was used in earlier studies to examine the attractiveness of the SNM associated with the reprocessing of used light water reactor (LWR) fuel by various reprocessing schemes. However, for a theft scenario the mass used in the Acquisition Factor of the FOM is the mass of the stolen object conta ining SNM ; whereas the mass used for analyzing the material attractiveness of the products of various reprocessing schemes in the earlier studies was a fraction of the bare critical mass in recognition that a successful proliferator must avoid a criticality accident. This paper will indicate how long after discharge the radiation emanating from a cooling assembly is no longer self-protecting. Additionally, this paper will give the time scale for the SNM within the assembly to become more attractive. These studies were performed at the request of the United States Department of Energy (DOE), and are based on the calculation of ''attractiveness levels'' that has been couched in terms chosen for consistency with those normally used for nuclear materials in DOE nuclear facilities. The methodology and key findings will be presented. Additionally, this paper discusses how the results presented herein impact the application of safeguards and the securitization of SNM, and how they could be used to help inform policy makers.

  7. In-core sipping method for the identification of failed fuel assemblies

    International Nuclear Information System (INIS)

    The failed fuel assembly identification system is an important safety system which ensures safe operations of reactor and immediate treatment of failed fuel rod cladding. The system uses an internationally recognized method to identify failed fuel assemblies in a reactor with fuel element cases. The in-core sipping method is customary used to identify failed fuel assemblies during refueling or after fuel rod cladding failure accidents. The test is usually performed after reactor shutdown by taking samples from each fuel element case while the cases are still in their original core positions. The sample activity is then measured to identify failed fuel assemblies. A failed fuel assembly identification system was designed for the NHR-200 based on the properties of the NHR-200 and national requirements. the design provides an internationally recognized level of safety to ensure the safety of NHR-200

  8. Revaluation on measured burnup values of fuel assemblies by post-irradiation experiments at BWR plants

    International Nuclear Information System (INIS)

    Fuel composition data for 8x8 UO2, Tsuruga MOX and 9x9-A type UO2 fuel assemblies irradiated in BWR plants were measured. Burnup values for measured fuels based on Nd-148 method were revaluated. In this report, Nd-148 fission yield and energy per fission obtained by burnup analyses for measured fuels were applied and fuel composition data for the measured fuel assemblies were revised. Furthermore, the adequacies of revaluated burnup values were verified through the comparison with burnup values calculated by the burnup analyses for the measured fuel assemblies. (author)

  9. Control assembly for controlling a fuel cell system during shutdown and restart

    Science.gov (United States)

    Venkataraman, Ramki; Berntsen, George; Carlson, Glenn L.; Farooque, Mohammad; Beachy, Dan; Peterhans, Stefan; Bischoff, Manfred

    2010-06-15

    A fuel cell system and method in which the fuel cell system receives and an input oxidant gas and an input fuel gas, and in which a fuel processing assembly is provided and is adapted to at least humidify the input fuel gas which is to be supplied to the anode of the fuel cell of the system whose cathode receives the oxidant input gas via an anode oxidizing assembly which is adapted to couple the output of the anode of the fuel cell to the inlet of the cathode of the fuel cell during normal operation, shutdown and restart of the fuel cell system, and in which a control assembly is further provided and is adapted to respond to shutdown of the fuel cell system during which input fuel gas and input oxidant gas cease to be received by the fuel cell system, the control assembly being further adapted to, when the fuel cell system is shut down: control the fuel cell system so as to enable a purging gas to be able to flow through the fuel processing assembly to remove humidified fuel gas from the processing assembly and to enable a purging gas to be able to flow through the anode of the fuel cell.

  10. Handling process of assemblies and fuel pins during the re-loading of a nuclear reactor

    International Nuclear Information System (INIS)

    The objective of this invention is to propose a process of handling assemblies and fuel pins, when reloading a nuclear reactor enclosing assemblies comprising a skeleton closed at the two ends inside of which fuel pins are disposed in vertical position. The reloading is made with the reactor vessel opened, and comprises: the transfer of fuel assemblies from a position to another in the reactor, the replacements of defective or spent assemblies by new assemblies and different controls using the surrounding swimming pool. Every replaced assembly is taken from the reactor vessel, put in a transfer container and transported in horizontal position in the fuel swimming pool near the reactor, this process allows a better re-use of the fuel pins which have not been completely spent in the changed assemblies using the skeletons of this assemblies during unloading

  11. Natural convection heat transfer within horizontal spent nuclear fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Canaan, R.E.

    1995-12-01

    Natural convection heat transfer is experimentally investigated in an enclosed horizontal rod bundle, which characterizes a spent nuclear fuel assembly during dry storage and/or transport conditions. The basic test section consists of a square array of sixty-four stainless steel tubular heaters enclosed within a water-cooled rectangular copper heat exchanger. The heaters are supplied with a uniform power generation per unit length while the surrounding enclosure is maintained at a uniform temperature. The test section resides within a vacuum/pressure chamber in order to subject the assembly to a range of pressure statepoints and various backfill gases. The objective of this experimental study is to obtain convection correlations which can be used in order to easily incorporate convective effects into analytical models of horizontal spent fuel systems, and also to investigate the physical nature of natural convection in enclosed horizontal rod bundles in general. The resulting data consist of: (1) measured temperatures within the assembly as a function of power, pressure, and backfill gas; (2) the relative radiative contribution for the range of observed temperatures; (3) correlations of convective Nusselt number and Rayleigh number for the rod bundle as a whole; and (4) correlations of convective Nusselt number as a function of Rayleigh number for individual rods within the array.

  12. Natural convection heat transfer within horizontal spent nuclear fuel assemblies

    International Nuclear Information System (INIS)

    Natural convection heat transfer is experimentally investigated in an enclosed horizontal rod bundle, which characterizes a spent nuclear fuel assembly during dry storage and/or transport conditions. The basic test section consists of a square array of sixty-four stainless steel tubular heaters enclosed within a water-cooled rectangular copper heat exchanger. The heaters are supplied with a uniform power generation per unit length while the surrounding enclosure is maintained at a uniform temperature. The test section resides within a vacuum/pressure chamber in order to subject the assembly to a range of pressure statepoints and various backfill gases. The objective of this experimental study is to obtain convection correlations which can be used in order to easily incorporate convective effects into analytical models of horizontal spent fuel systems, and also to investigate the physical nature of natural convection in enclosed horizontal rod bundles in general. The resulting data consist of: (1) measured temperatures within the assembly as a function of power, pressure, and backfill gas; (2) the relative radiative contribution for the range of observed temperatures; (3) correlations of convective Nusselt number and Rayleigh number for the rod bundle as a whole; and (4) correlations of convective Nusselt number as a function of Rayleigh number for individual rods within the array

  13. Nuclear mass inventory, photon dose rate and thermal decay heat of spent research reactor fuel assemblies

    International Nuclear Information System (INIS)

    This document has been prepared to assist research reactor operators possessing spent fuel containing enriched uranium of United States origin to prepare part of the documentation necessary to ship this fuel to the United States. Data are included on the nuclear mass inventory, photon dose rate, and thermal decay heat of spent research reactor fuel assemblies. Isotopic masses of U, Np, Pu and Am that are present in spent research reactor fuel are estimated for MTR, TRIGA and DIDO-type fuel assembly types. The isotopic masses of each fuel assembly type are given as functions of U-235 burnup in the spent fuel, and of initial U-235 enrichment and U-235 mass in the fuel assembly. Photon dose rates of spent MTR, TRIGA and DIDO-type fuel assemblies are estimated for fuel assemblies with up to 80% U-235 burnup and specific power densities between 0.089 and 2.857 MW/kg[sup 235]U, and for fission product decay times of up to 20 years. Thermal decay heat loads are estimated for spent fuel based upon the fuel assembly irradiation history (average assembly power vs. elapsed time) and the spent fuel cooling time

  14. Manufacture of a Dual-Cooled Fuel Assembly Mockup for Mechanical Characterization Tests

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jaeyong; Kim, Hyungkyu; Yoon, Kyungho; Lee, Kanghee; Kang, Heungseok [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    All components were made of stainless steel 304 for research. A DUO fuel assembly mockup was assembled by mechanical fastening and laser welding methods with them. The conceptual feasibility of each component was checked through it. In this paper, manufactured items for a DUO fuel and a DUO fuel assembly are briefly described. Although the research of a DUO fuel has been done by USA, they have just focused on pellets, not mechanical parts such as TEP/BEP, GTs, and SGs. We designed and manufactured them and assembled a DUO fuel assembly. The realizable possibility of a DUO fuel assembly was checked. Mechanical characterization tests will be performed to measure the DUO fuel's mechanical properties such as bending rigidity, modal characteristics, impact durability, etc.

  15. Evaluation of the fuel-element assembly non-hermeticity at its early stage

    International Nuclear Information System (INIS)

    The given paper deals with control of the fuel-element assembly shell state at the early stage of failure development. Technique for the fuel-element assembly shell state evaluation are described. A method for assembly failure detection, used at WWR of the Institute for Nuclear Research is described also

  16. Design report of the disposal canister for twelve fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Raiko, H. [VTT Energy, Espoo (Finland); Salo, J.P. [Posiva Oy, Helsinki (Finland)

    1999-05-01

    The report provides a summary of the design of the canister for final disposal of spent nuclear fuel. The canister structure consists of a cylindrical massive nodular graphite cast iron insert covered by a 50 mm thick copper overlay. The capacity of the canister is 12 assemblies of BWR or VVER 440 fuel. The canister shall be tight with a high probability for about 100 000 years. The good and long lasting tightness requires: (1) The good initial tightness that is achieved by high quality requirements and extensive quality control, (2) The good corrosion resistance, which is obtained by the overpack of oxygen free copper, and (3) Mechanical strength of the canister, that is ensured by analyses (the following loads are considered: hydrostatic pressure, even and uneven swelling pressure of bentonite, thermal effects, and elevated hydrostatic pressure during glaciation. The allowed stresses and strains are set in such a way that reasonable engineering safety factors are obtained in all assessed design base loading cases). The canister shall limit the radiation dose rate outside the canister to minimise the radiolysis of the water in the vicinity of the canister. The canister insert shall keep the fuel assemblies in a subcritical configuration even if the void in the canister is filled with water due to postulated leakage. The design basis of the canister is set, the performed analyses are summarised and the results are assessed and discussed in the report. (orig.) 35 refs.

  17. Design report of the disposal canister for twelve fuel assemblies

    International Nuclear Information System (INIS)

    The report provides a summary of the design of the canister for final disposal of spent nuclear fuel. The canister structure consists of a cylindrical massive nodular graphite cast iron insert covered by a 50 mm thick copper overlay. The capacity of the canister is 12 assemblies of BWR or VVER 440 fuel. The canister shall be tight with a high probability for about 100 000 years. The good and long lasting tightness requires: (1) The good initial tightness that is achieved by high quality requirements and extensive quality control, (2) The good corrosion resistance, which is obtained by the overpack of oxygen free copper, and (3) Mechanical strength of the canister, that is ensured by analyses (the following loads are considered: hydrostatic pressure, even and uneven swelling pressure of bentonite, thermal effects, and elevated hydrostatic pressure during glaciation. The allowed stresses and strains are set in such a way that reasonable engineering safety factors are obtained in all assessed design base loading cases). The canister shall limit the radiation dose rate outside the canister to minimise the radiolysis of the water in the vicinity of the canister. The canister insert shall keep the fuel assemblies in a subcritical configuration even if the void in the canister is filled with water due to postulated leakage. The design basis of the canister is set, the performed analyses are summarised and the results are assessed and discussed in the report. (orig.)

  18. quality assurance in procurement, design and fabrication of nuclear fuel assemblies

    International Nuclear Information System (INIS)

    In this study, the scope of a Safety Guide which specified certain principles related with quality assurance in the procurement, design and manufacture of nuclear fuel assemblies is discussed. Requirements and recommendations for quality assurance programmes that are relevant with the procurement, design, manufacture, inspection, testing, packaging, shipping, storage and receiving inspection of fuel assemblies in nuclear power plants are provided in this paper. In addition items that require particular consideration in a typical fuel assembly manufacture and some processes which have been adopted in fuel manufacturing and quality verification are explained in a general context. Items in an identification system of a fuel assembly are given

  19. Study on pressure drop prediction of Tight Lattice Fuel Assembly using CFD

    International Nuclear Information System (INIS)

    This paper presents a prediction method about pressure drop of the tight lattice fuel assembly. To evaluate the core performance, the pressure drop of fuel assembly is important parameter for the reliability and safety. The shape of fuel assembly is complicated and the shape has a strong effect on pressure drop. Therefore, to obtain the pressure drop of fuel assembly, the experiments are needed. However the experiments need a lot of time and money. The purpose of this study is to predict the pressure drop of the tight lattice fuel assembly using CFD (computational fluid dynamics) and law of two phase flow similarity without experiments. Prediction method for pressure drop is, first the shape of fuel assembly was reproduced by 3D-CAD, then to evaluate the parameters depending on the shape using CFD analysis under the single phase flow condition. And then, to calculate pressure drop of two phase flow using law of two phase flow similarity. The predicted results by this method were compared with the experimental results of 3 types of the tight lattice fuel assemblies, 7-rod, 14-rod, and 37-rod fuel assembly. Fluid conditions (pressure, mass flux and quality) of test results covered the typical operating range of the LWR fuel assembly. It was found from comparison result that this method can predict the pressure drop of the LWR fuel assembly with sufficient accuracy. The prediction accuracy (predicted value/test value) is about 10%. (author)

  20. Criteria for removal of defective fuel rod from fuel assembly under repair without cladding rupture

    International Nuclear Information System (INIS)

    During repair of a failed fuel assembly (FA) there is a risk of cladding rupture while a defective fuel rod is forced out of the assembly skeleton. To reduce the corresponding risks, a program of experimental and analytical studies for WWER fuel was performed. It resulted in formulation of criteria for successful removal of the defective fuel rod from the FA under repair. 'Successful' means that no cladding rupture occurs. The paper summarizes the available data of post-irradiation examinations of WWER FAs with leaking fuel rods. A technique for express estimation of hydrogen content in cladding of a defective fuel rod is presented. The degradation of cladding mechanical properties can be estimated with this technique as well. A criterion of severe secondary hydriding involved in the risk analysis is also discussed. Finally, it is shown how the information on operation conditions may be used for prompt evaluation of the limiting force for successful removal of a defective fuel rod during FA repair in the inspection stand. (author)

  1. Development of Tools for Treating an Irradiated Fuel Rod Assembly in the Pool of HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Hong, J. T.; Ahn, S. H.; Kim, K. H.; Joung, C. Y. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-07-01

    To inspect a fuel rod during irradiation testing at the test loop of a research reactor, the test rig should be disassembled from the IPS (In-pile test section), and the targeted fuel rod assembly should be disassembled from the test rig and encapsulated in a cask to deliver the assembly to the hot cell. In addition, the fuel rod assembly under inspection in the hot cell should be delivered to the reactor pool and reassembled into the test rig to resume the irradiation test. Because the irradiated fuel rod is highly radioactive, all of the assembly and disassembly operations should be carried out in the reactor pool. Therefore, special tools need to be developed to treat the test rig in the pool of a research reactor. In this study, a new mechanically detachable fuel rod assembly has been developed for intermediate inspection during irradiation test at HANARO. A fuel rod assembly can be divided into two parts, such as an instrumented fuel rod assembly and a non-instrumented fuel rod assembly. In particular, an instrumented fuel rod assembly is assembled at the lower part of the test rig, and a non-instrumented fuel rod assembly is assembled at the bottom of the instrumented fuel rod assembly. The non-instrumented fuel rod assembly is locked in the test rig during irradiation test, and is easily disassembled from the instrumented fuel rod assembly by pushing the anchor button and twisting the non-instrumented fuel rod assembly. In addition, because a test rig is 5.4 meters long and the disassembling operation should be carried out at 6 meters deep in the pool of HANARO, tools to help disassemble and assemble the non-instrumented fuel rod assembly have also been developed. All components were designed to operate mechanically and are made of stainless steel and Al 6061 to minimize the effects from the radioactivity. The performance of the developed fuel rod assembly and tools have been verified through an out pile test.

  2. Operational experience of fuel assembly TVS-2 at Balakovo NPP

    International Nuclear Information System (INIS)

    TVS-2 operational history, main goals of TVS-2 development like: creation of fuel assembly (FA) with hard casing, able to resist bending and rotating loads at loose interaction of fuel elements with spacer grids (SG) cells during operation and creation of FA, able to work under prolonged fuel cycles, i.e., permitting to achieve more depth of burn up and also main solutions, accepted for TVS-2 construction like: 1) Hard armed casing, created by direct welding of spacer grids (SG) to control rod guides (CRG); 2) Hard to bending SG increased up to 30 mm high (instead of 20 mm for UTVS); 3) Special construction of SG cells to reduce forces on fuel elements sliding; 4) For CRG was used alloy E635 to improve casing stability; 5) Enforced end parts to improve FAs stability are presented in this paper. Types of TVS-2, used at Balakovo NPP, RPS CR drop time at imitation of EP actuation after reactor shutdown for reloading (Unit 1), RPS CR withdrawal forces before reactor startup after reloading (Unit 1), FAs bents after reactor shutdown for reloading (Unit 1) and FA clad condition of reactors at Units 2, 3, 4 as well as at Unit 1 are shown. At the end the following further activities at Balakovo NPP are listed: 1) Continue industrial operation of TVS-2; 2) Shift units to fuel cycle 3 * 16-18 months on basis months on basis of TVS of TVS-2 -2 with the purpose improve capacity factor; 3) Enhance units capacity up to 104 % Nnom; 4) Increase vertical speed of FAs, movement in reactor, SFP, FAC up to 1 and and 2 m/min. accordingly; 5) Implement TVS-2M, having big mass of fuel due to increase of the fuel pillar (on 150 mm) and decrease of pellet hole (up to 1,2 mm)

  3. Out-of-pile Verifying Test for the Hydraulic Stability of the CARR Standard Fuel Assembly

    Institute of Scientific and Technical Information of China (English)

    2002-01-01

    The CARR standard fuel element is a flat-plate-type assembly. A fuel plate consists of 0.6 mmthickness layer of uranium- silicon - aluminum fuel (U3Si2-Al) and 0.38 mm thickness of aluminumcladding. The fuel plates are attached to aluminum alloy side plates by a "roll swaging" technique. Thistype of fuel assembly is first used in China. The testing simulates the in-pile thermal-hydraulic operating conditions except for neutron

  4. Analysis of Unplated Subcritical Experiments Using Fresh Fuel Assemblies

    International Nuclear Information System (INIS)

    The number of spent nuclear fuel assemblies taken from nuclear power plants and to be stored in existing storage pools is increasing. Therefore, there is a need to optimize the storage configurations. The computer codes and cross sections used to analyze proposed storage configurations must be validated through comparison with experimental data. Restrictive values of ksafe, caused by limited data, can prevent optimal storage utilization. As a collaborative effort between Westinghouse Safety Management Solutions, Oak Ridge National Laboratory (ORNL), Georgia Institute of Technology, and the University of Missouri Research Reactor (MURR), more than 120 experiments were performed using four highly enriched MURR fuel assemblies. The 252Cf-source-driven noise analysis technique developed at ORNL was used as the measurement method for these experiments. This method is based on calculating a specific ratio of measured auto-power and cross-power spectral densities. Twenty-two unique configurations from the MURR experimental program were analyzed for benchmarking purposes.These subcritical experiments were described and analyzed in this paper to provide new measurements to increase the amount of data available for benchmarking criticality codes and cross sections for systems that are far from critical (keff eff values. Inferred keff values ranged from 0.648 ± 0.005 to 0.860 ± 0.006. A simplified benchmark model is described that consists of the four fuel assemblies, four 3He detectors, detector drywells, and the water reflector. For these measurements, the calculated ratio and keff values agreed with the measurement results within the measurement uncertainty. All of the analyzed configurations were considered acceptable for validation of computer codes and cross sections

  5. Development of a reference spent fuel library of 17x17 PWR fuel assemblies

    International Nuclear Information System (INIS)

    One of the most common ways to investigate new Non-Destructive Assays (NDA) for the spent fuel assemblies are Monte Carlo simulations. In order to build realistic models the user must define in an accurate way the material compositions and the source terms in the system. This information can be obtained using burnup codes such as ORIGEN-ARP and ALEPH2.2, developed at SCK-CEN. These software applications allow the user to select the irradiation history of the fuel assembly and to calculate the corresponding isotopic composition and neutron/gamma emissions as a function of time. In the framework of the development of an innovative NDA for spent fuel verifications, SCK•CEN built an extensive fuel library for 17x17 PWR assemblies, using both ORIGEN-ARP and ALEPH2.2. The parameters considered in the calculations were initial enrichment, discharge burnup, and cooling time. The combination of these variables allows to obtain more than 1500 test cases. Considering the broad range of the parameters, the fuel library can be used for other purposes apart from spent fuel verifications, for instance for the direct disposal in geological repositories. In addition to the isotopic composition of the spent fuel, the neutron and photon emissions were also calculated and compared between the two codes. The comparison of the isotopic composition showed a good agreement between the codes for most of the relevant isotopes in the spent fuel. However, specific isotopes as well as neutron and gamma spectra still need to be investigated in detail.

  6. Development of a reference spent fuel library of 17x17 PWR fuel assemblies

    International Nuclear Information System (INIS)

    One of the most common ways to investigate new Non- Destructive Assays (NDA) for the spent fuel assemblies are Monte Carlo simulations. In order to build realistic models the user must define in an accurate way the material compositions and the source terms in the system. This information can be obtained using burnup codes such as ORIGEN-ARP and ALEPH2.2, developed at SCK•CEN. These software applications allow the user to select the irradiation history of the fuel assembly and to calculate the corresponding isotopic composition and neutron/gamma emissions as a function of time. In the framework of the development of an innovative NDA for spent fuel verifications, SCK•CEN built an extensive fuel library for 17x17 PWR assemblies, using both ORIGENARP and ALEPH2.2. The parameters considered in the calculations were initial enrichment, discharge burnup, and cooling time. The combination of these variables allows to obtain more than 1500 test cases. Considering the broad range of the parameters, the fuel library can be used for other purposes apart from spent fuel verifications, for instance for the direct disposal in geological repositories. In addition to the isotopic composition of the spent fuel, the neutron and photon emissions were also calculated and compared between the two codes. The comparison of the isotopic composition showed a good agreement between the codes for most of the relevant isotopes in the spent fuel. However, specific isotopes as well as neutron and gamma spectra still need to be investigated in detail.

  7. Experimental and calculational study of temperature distributions in deformed model fuel assemblies of fast reactors

    International Nuclear Information System (INIS)

    Experimental and calculational data tastify to absence of temperature nonuniformity stabilization in fuel assembly peripheral area. The effect of fuel lattice deformation on the fuel assembly temperature field at shroud crushing in the core centre is demonstrated. 17 refs.; 21 figs

  8. Complete characterization of spent fuel assemblies and alpha waste packages

    International Nuclear Information System (INIS)

    One center of interest of the research program at the Karlsruhe Nuclear Research Centre concentrates on neutron detection systems for special applications. Neutron measurements have the advantage because of high transparency of the waste material and simple detectability of neutrons. The determination of neutron radiation is advantageous for measuring the fissile material content and multiplication effects. Furthermore, radionuclides showing an alpha decay can be detected when the process is accompanied by neutron emission. Two measuring systems developed together with NUKEM are presented in this document: FAMOS, the fuel assembly monitoring system, and the alpha waste monitoring system

  9. Welding fixture for nuclear reactor fuel assembly grid

    International Nuclear Information System (INIS)

    A welding fixture for a nuclear fuel assembly grid having rigid top and bottom members and having apparatus for releasably securing them together. The bottom and top members each have an egg-crate configuration of interleaved fixture straps. Top notches on the bottom fixture straps' top edges are placed to engage the grid straps' bottom edges when the grid straps are aligned for welding. Likewise, bottom notches on the top fixture straps' bottom edges are placed to engage the grid straps' top edges when the grid straps are aligned for welding

  10. A spacer grid hysteretic model for the structural analysis of spent fuel assemblies under impact

    International Nuclear Information System (INIS)

    This paper presents a methodology for determining the response of spent fuel assembly spacer grids subjected to transport cask impact loading. The spacer grids and their interaction with rod-to-rod loading are the most critical components governing the structural response of spent fuel assemblies. The purpose of calculating the assembly response is to determine the resistance to failure of spent fuel during regulatory transport. The failure frequency computed from these analyses is used in calculating category B spent fuel cask containment source term leakage rates for licensing calculations. Without defensible fuel rod failure frequency prediction calculations, assumptions of 100% fuel failure must be made, leading to leak tight cask design requirements

  11. Simplified process for leaching precious metals from fuel cell membrane electrode assemblies

    Science.gov (United States)

    Shore, Lawrence; Matlin, Ramail

    2009-12-22

    The membrane electrode assemblies of fuel cells are recycled to recover the catalyst precious metals from the assemblies. The assemblies are cryogenically embrittled and pulverized to form a powder. The pulverized assemblies are then mixed with a surfactant to form a paste which is contacted with an acid solution to leach precious metals from the pulverized membranes.

  12. Photon dose rates from ET-RR-2 spent fuel assemblies

    International Nuclear Information System (INIS)

    The external dose rates from spent fuel assemblies consist of gamma ray and neutron components. The source of gamma ray is from fission products and actinides in the spent fuel and from activation products in structural components of the fuel assembly. Neutrons originate from spontaneous fission in actinides within the spent fuel and from ( ,n) reactions in oxide fuel. However, a significant number of neutrons are produced due to further fission within the fuel. The degree of physical protection given to spent fuel assemblies is largely dependent upon the photon dose rate of the spent fuel material. These data are not always know or readily determined. The photon dose rate from spent fuel assemblies is calculated for the purpose of estimating the radiation level. For safe spent fuel assembly containment, the thermal heat load generated by the decay of fission products in the spent fuel material is an important consideration. The dose rate estimate for an MTR type fuel assembly can be made based upon the line source model dose rate. Photon dose rates as a function of fission product decay times have been calculated for the above spent fuel assembly knowing the mass of fuel burned, fraction of fuel burned, and the fuel assembly specific power. The fission products selected in this study are: Sr-90, Y-90,Cs-133, Cs-134, Cs-137, Ba-137m, Ce-144, Pr-144, Ru-106, Rh-106, Zr-95 and Nb-95. This calculations performed by PHDOSE code, which has been complied to work in Pc environments. Correlations between the activity ratio of some fission products and the burn up and cooling times of the spent fuel assembly have been tested

  13. Integrity assessment of test fuel assemblies of the High Temperature Engineering Test Reactor

    International Nuclear Information System (INIS)

    Assessment of integrity has been made on the B-type fuel assemblies, which will be loaded in the High Temperature Engineering Test Reactor (HTTR) as test fuel assemblies. Specifications of coated fuel particles for the B-1 type fuel assembly have been slightly changed in the fuel kernel diameter and thickness of coating layers from those for the A-type fuel assembly, which is employed as the driver fuel. These changes have been directed toward safer side in developing this advanced fuel for use up to higher burnups at higher temperatures. The B-2 type fuel assembly uses the zirconium-carbide (ZrC) coating layer with excellent high-temperature chemical stability, instead of the silicon carbide (SiC) layer. This change has lead to demonstration of its better performance than the A-type fuel assembly in the kernel migration, corrosion by fission products including palladium, and coating failure at extremely high temperatures. The B-3 type fuel assembly adopts the (U,Th)O2 kernel - SiC TRISO coated fuel articles. The service condition (1000degC and 22,000 MWd/t) of the B-3 type fuel assembly is decided as the range within which the performance data of the fuel have been sufficiently obtained. Thus, it has been judged that the integrity of these B-type fuel assemblies will be maintained under the normal operating conditions of the HTTR. Moreover, the validity of the permissible design limit of the fuel has been confirmed, which requires that the fuel temperature shall not exceed 1,600degC at anticipated operational transients. (author)

  14. Critical experiments for BWR fuel assemblies with cluster of gadolinia rods

    International Nuclear Information System (INIS)

    Gadolinia-bearing fuel rods are needed for high-burnup fuels. Strong neutron absorption of gadolinia makes an assembly heterogeneous from the viewpoint of reactor physics. The cluster of gadolinia-bearing fuel rods is useful for higher-burnup fuels than current fuels. Few critical experiments have been reported for fuel assemblies with the cluster of gadolinia-bearing fuel rods. We conducted critical experiments for BWR fuel assemblies with the cluster of gadolinia-bearing fuel rods in the Toshiba Nuclear Critical Assembly (NCA). Critical water level and power distribution were measured. Measurements were compared with analyses by a continuous-energy Monte Carlo code, MCNP, with the JENDL3.3 nuclear data library. (author)

  15. Membrane electrode assemblies for unitised regenerative polymer electrolyte fuel cells

    Science.gov (United States)

    Wittstadt, U.; Wagner, E.; Jungmann, T.

    Membrane electrode assemblies for regenerative polymer electrolyte fuel cells were made by hot pressing and sputtering. The different MEAs are examined in fuel cell and water electrolysis mode at different pressure and temperature conditions. Polarisation curves and ac impedance spectra are used to investigate the influence of the changes in coating technique. The hydrogen gas permeation through the membrane is determined by analysing the produced oxygen in electrolysis mode. The analysis shows, that better performances in both process directions can be achieved with an additional layer of sputtered platinum on the oxygen electrode. Thus, the electrochemical round-trip efficiency can be improved by more than 4%. Treating the oxygen electrode with PTFE solution shows better performance in fuel cell and less performance in electrolysis mode. The increase of the round-trip efficiency is negligible. A layer sputtered directly on the membrane shows good impermeability, and hence results in high voltages at low current densities. The mass transportation is apparently constricted. The gas diffusion layer on the oxygen electrode, in this case a titanium foam, leads to flooding of the cell in fuel cell mode. Stable operation is achieved after pretreatment of the GDL with a PTFE solution.

  16. SOURCE OF BURNUP VALUES FOR COMMERCIAL SPENT NUCLEAR FUEL ASSEMBLIES

    International Nuclear Information System (INIS)

    Waste packages are loaded with commercial spent nuclear fuel (SNF) that satisfies the minimum burnup requirements of a criticality loading curve. The burnup value assigned by the originating nuclear utility to each SNF assembly (assigned burnup) is used to load waste packages in compliance with a criticality loading curve. The burnup provided by a nuclear utility has uncertainties, so conservative calculation methods are used to characterize those uncertainties for incorporation into the criticality loading curves. Procedural safety controls ensure that the correct assembly is loaded into each waste package to prevent a misload that could create a condition affecting the safety margins. Probabilistic analyses show that procedural safety controls can minimize the chance of a misload but can not completely eliminate the possibility. Physical measurements of burnup with instrumentation in the surface facility are not necessary due to the conservative calculation methods used to produce the criticality loading curves. The reactor records assigned burnup of a commercial SNF assembly contains about two percent uncertainty, which is increased to five-percent to ensure conservatism. This five-percent uncertainty is accommodated by adjusting the criticality loading curve. Also, the record keeping methods of nuclear utilities are not uniform and the level of detail required by the NRC has varied over the last several decades. Thus, some SNF assemblies may have assigned burnups that are averages for a batch of assemblies with similar characteristics. Utilities typically have access to more detailed core-follow records that allow the batch average burnup to be changed to an assembly specific burnup. Alternatively, an additional safety margin is incorporated into the criticality loading curve to accommodate SNF assemblies with batch average burnups or greater uncertainties due to the methodology used by the nuclear utility. The utility records provide the assembly identifier

  17. Nuclear fuel assembly top nozzle with improved arrangement of hold-down leaf spring assemblies

    Energy Technology Data Exchange (ETDEWEB)

    De Mario, E.E.; Lawson, C.N.

    1991-10-15

    This patent describes a top nozzle for use in a fuel assembly having guide thimbles for mounting the top nozzle. It comprises: a lower adapter plate having a periphery bounding an interior thereof mountable to the guide thimbles: guide structures attached to and extending along the periphery of the adapter plate and upwardly therefrom; an upper hold-down plate mounted to the guide structures for slidable movement relative thereto such that the upper plate can move toward and away from the interior of the lower plate within the space bounded by the guide structures as the upper plate slidably moves along the guide structures; and leaf spring assemblies interposed between and engaged with the lower and upper plates so as to yieldably support the upper plate in spaced relation above the lower plate and bias the upper plate for movement away from the lower plate; the leaf spring assemblies being provided in a non-peripheral arrangement relative to the periphery of the lower plate in which the assemblies cross the interior of the lower plate in a diagonal fashion between adjacent ones of the guide structures.

  18. Experimental research of local hydrodynamic characteristics of fast reactor fuel assembly peripheral zone. 4

    International Nuclear Information System (INIS)

    Measurements were made of shear stress distribution and the velocity field of an aerodynamic model of the fast breeder reactor fuel assembly periphery. The effect was studied of a 50% disturbance of the geometry of a corner rod in a fuel assembly as against normal geometry. The coefficient of friction in the channel was assessed in dependence on the Reynolds number. The distribution of shear stresses in the walls of the fuel assembly and on spacers is graphically represented. (M.D.)

  19. Natural convection in the gap volume between fuel assembly boxes

    International Nuclear Information System (INIS)

    The aim of this study is the prediction of temperature and velocity of the moderator water in the gap region between the fuel assemblies for an advanced pressurized water reactor with supercritical water in the primary loop (PWR-SC). In addition the occurrence of unwanted natural convection phenomena in the moderator gaps is investigated. The simulation of the gap water flow between fuel assemblies in the reactor core is done by applying a porous media approach because size and geometric complexity of the flow region in the gap prevent a more detailed solution of the velocity and temperature fields. Instead, macroscopic conservation equations have been employed by volume averaging the microscopic conservation equations. The equations are derived in dependency of the porosity, the permeability and the resistance function of the porous medium. The latter two ones have been determined by applying the CFD software Star-CD on the detailed gap geometry. The results have been compared with correlations for similar flow cases and applied in the macroscopic conservation equations. Finally, the equation system is implemented in the commercial multiphysics tool COMSOL. A parametric study with different mass flow rates at a given coolant temperature distribution showed that a minimum moderator mass flow rate will be required to avoid natural convection in the gap volume, and thus to ensure a stable neutron flux condition. (author)

  20. Underwater remote repairing device for fuel assembly supporting lattice

    International Nuclear Information System (INIS)

    Purpose: To effectively repair a supporting lattice by employing a grinding tool mounted at the lower end of a tool supporting post and an underwater television camera disposed in the vicinity of the tool. Constitution: A grinding tool is mounted at the lower end of a tool supporting post supported by rotatable tool support post supporting arm or the like so as to be laterally and longitudinally rotatable, and an underwater television camera is mounted in the height of the grinding tool of the underwater television supporting post by mounting an underwater television supporting plate via a rotatable arm onto the underwater television turning post rotatably mounted at an upper stationary plate and lower trestle. The grinding tool is remotely operated while confirming the grinding position with the underwater television camera to grind and remove the outer peripheral side and the corner and the like minutely scratched or deformed of the supporting lattice of a fuel assembly. Thus, the fuel assembly can be used consecutively. (Sekiya, K.)

  1. Void fractions and pressure drops in reactor fuel assemblies

    International Nuclear Information System (INIS)

    Model fuel assemblies of the Advanced Thermal Reactor (2nd design) and the JPDR-II have been tested in an air-water test loop (FAT-I) to study two-phase flow characteristics at the system pressure of 3.5 kg/cm2 g, the void fraction of 10 - 50% and the water flow rate up to 60 t/h. Average void fractions and pressure drops due to spacers, base plate, tie plate as well as rod bundle were measured. The ratio of the average void fraction to the volumetric flow fraction of air was 0.95 for the ATR fuel assembly and 0.7 for the JPDR-II. The frictional pressure drop for the rod bundle was expressed as a function of the volumetric flow fraction. An estimation method of the total pressure drop at normal reactor operation pressure has been derived in the light of two-phase flow in pipes. Nearly one half of the total pressure drop was the pressure drop other than for the rod bundle friction. Observations through a window showed that bubble flow regime predominated throughout the experiments. (author)

  2. Pool inspection techniques for surveillance and further development of high burnup fuel assemblies

    International Nuclear Information System (INIS)

    The pool inspection techniques which have been used in fuel assembly surveillance programs for many years are suitable for high burnup fuel assemblies too. The techniques have been adapted to the requirements of new fuel assembly concepts with higher burnup potential. For high burnup, the emphasis within the scope of examination techniques available has shifted towards a characterization of the corrosion behaviour and surveillance of the geometrical dimensions of the fuel assemblies. In order to accomplish these tasks complementary techniques will have to be developed. (orig.)

  3. Lead test fuel assembly physics for erbium fuel at the San Onofre Nuclear Generating Station

    International Nuclear Information System (INIS)

    ABB Combustion Engineering Nuclear Fuel (ABB C-E) has completed a qualification program for the use of Er2O3 as a burnable absorber in pressurized water reactors. Neutronic properties and performance have been assessed in a development program that included the following: 1. evaluation and processing of basic cross-section data for naturally occurring erbium isotopes and capture/decay products of erbium and thulium; 2. performance of critical experiments by ABB C-E, Rensselaer Polytechnic Institute, and Southern California Edison (SCE); 3. comparison of methods between ABB C-E and SCE (Ref. 6); and 4. irradiation of lead test fuel assemblies (LTAs) and corefollow analyses in collaboration between ABB C-E and SCE (16 x 16 fuel) and Baltimore Gas ampersand Electric Company (14 x 14 fuel). The subject of this paper is the physics core-follow results for San Onofre unit 2

  4. The PS booster

    CERN Multimedia

    1972-01-01

    The PS booster which accelerates protons from the linac at an energy of 50 MeV to an energy of 800 MeV before injecting them into the main magnet ring of the synchrotron. The booster consists of four superposed rings. In the photograph can be seen the input beam line from the linac and the output beam lines, where beams from the four booster levels have been combined into two beams before final recombination.

  5. Irradiation experiments of the 13th-15th OGL-1 fuel assemblies

    International Nuclear Information System (INIS)

    The Oarai Gas Loop-1, OGL-1, which had been installed in the Japan Materials Testing Reactor (JMTR) of the Oarai Research Establishment of the Japan Atomic Energy Research Institute (JAERI), was an in-pile helium loop for irradiation of high-temperature gas-cooled reactor fuels at high pressure and temperature. This report describes results of fabrication, irradiation and post-irradiation examinations (PIEs) of the 13th-15th OGL-1 fuel assemblies. The 13th and 15th fuel assemblies employed the first-charge fuel of the High Temperature Engineering Test Reactor (HTTR). The 13th assembly was loaded with a high quality fuel, whose as-produced failure fraction had been drastically decreased, compared with that for fuels before that time. The 15th assembly was loaded with a fuel, which had been produced by the same apparatus that was used afterwards for the first charge fuel of the HTTR. Both of these fuel assemblies gave good results in PIEs as well as in the fission-gas release rates during irradiation. The 14th fuel assembly used a trial product of an advanced fuel for high burnup utilization, which employed coated fuel particles (CFPs) with thicker coating layers than those for the first charge fuel. This fuel assembly indicated a spike release of fission gas during irradiation at 1500degC after a transient temperature increase up to this value. As a whole, all of the 13th - 15th assemblies demonstrated good performance of the loaded fuels, giving significantly lower values in fission-gas release rates during irradiation and in failure fractions of CFPs after irradiation, than the corresponding design limit values for the first charge fuel of the HTTR. (author)

  6. Irradiation experiments of the 13th-15th OGL-1 fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Hayashi, Kimio; Sawa, Kazuhiro; Kitajima, Toshio [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Shiratori, Tetsuo; Kikuchi, Hironobu; Fukuda, Kousaku; Itoh, Tadaharu; Waragai, Heita [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2000-01-01

    The Oarai Gas Loop-1, OGL-1, which had been installed in the Japan Materials Testing Reactor (JMTR) of the Oarai Research Establishment of the Japan Atomic Energy Research Institute (JAERI), was an in-pile helium loop for irradiation of high-temperature gas-cooled reactor fuels at high pressure and temperature. This report describes results of fabrication, irradiation and post-irradiation examinations (PIEs) of the 13th-15th OGL-1 fuel assemblies. The 13th and 15th fuel assemblies employed the first-charge fuel of the High Temperature Engineering Test Reactor (HTTR). The 13th assembly was loaded with a high quality fuel, whose as-produced failure fraction had been drastically decreased, compared with that for fuels before that time. The 15th assembly was loaded with a fuel, which had been produced by the same apparatus that was used afterwards for the first charge fuel of the HTTR. Both of these fuel assemblies gave good results in PIEs as well as in the fission-gas release rates during irradiation. The 14th fuel assembly used a trial product of an advanced fuel for high burnup utilization, which employed coated fuel particles (CFPs) with thicker coating layers than those for the first charge fuel. This fuel assembly indicated a spike release of fission gas during irradiation at 1500degC after a transient temperature increase up to this value. As a whole, all of the 13th - 15th assemblies demonstrated good performance of the loaded fuels, giving significantly lower values in fission-gas release rates during irradiation and in failure fractions of CFPs after irradiation, than the corresponding design limit values for the first charge fuel of the HTTR. (author)

  7. Experimental study of hydrodynamically induced vibrational processes in VVER-440 fuel assemblies

    International Nuclear Information System (INIS)

    Investigations are described of hydrodynamically induced vibrations in a single fuel assembly of a VVER-440 reactor, performed on a full-scale model installed in a closed loop filled with distilled water; the model fuel elements contained simulators of fuel pellets. Data on hydrodynamic loads were obtained by measuring pressure oscillations along the height of the fuel assembly case. Results of the measurements are presented in graphs and are discussed in some detail. (A.K.)

  8. Between-cycle laser system for depressurization and resealing of modified design nuclear fuel assemblies

    International Nuclear Information System (INIS)

    A laser beam is used to puncture fuel cladding for release of contained pressurized fission gas from plenum sections or irradiated fuel pins. Exhausted fission gases are collected and trapped for safe disposal. The laser beam, adjusted to welding mode, is subsequently used to reseal the puncture holes. The fuel assembly is returned to additional irradiation or, if at end of reactivity lifetime, is routed to reprocess. The fuel assembly design provides graded cladding lengths, by rows or arrays, such that the cladding of each component fuel element of the assembly is accessible to laser beam reception

  9. TRU transmutation using ThO2-UO2 and fully ceramic micro-encapsulated fuels in LWR fuel assemblies

    International Nuclear Information System (INIS)

    The objective of this work is to design new LWR fuel assemblies which are able to efficiently destroy TRU (transuranics) nuclide without degradation of safety aspects by using ThO2-UO2 fuel pins and FCM (Fully Ceramic Micro-encapsulated) fuel pins containing TRU fuel particles. Thorium was mixed to UO2 in order to reduce the generation of plutonium nuclides and to save the uranium resources in the UO2 pins. Additionally, the use of thorium contributes to the extension of the fuel cycle length. All calculations were performed by using DeCART (Deterministic Core Analysis based on Ray Tracing) code. The results show that the new concept of fuel assembly has the TRU destruction rates of ∼40% and ∼25% per 1200 EFPD (Effective Full Power Day) over the TRU FCM pins and the overall fuel assembly, respectively, without degradation of FTC and MTC

  10. Numerical study of assembly pressure effect on the performance of proton exchange membrane fuel cell

    International Nuclear Information System (INIS)

    The performance of the fuel cell is affected by many parameters. One of these parameters is assembly pressure that changes the mechanical properties and dimensions of the fuel cell components. Its first duty, however, is to prevent gas or liquid leakage from the cell and it is important for the contact behaviors of fuel cell components. Some leakage and contact problems can occur on the low assembly pressures whereas at high pressures, components of the fuel cell, such as bipolar plates (BPP), gas diffusion layers (GDL), catalyst layers, and membranes, can be damaged. A finite element analysis (FEA) model is developed to predict the deformation effect of assembly pressure on the single channel PEM fuel cell in this study. Deformed fuel cell single channel model is imported to three-dimensional, computational fluid dynamics (CFD) model which is developed for simulating proton exchange membrane (PEM) fuel cells. Using this model, the effect of assembly pressure on fuel cell performance can be calculated. It is found that, when the assembly pressure increases, contact resistance, porosity and thickness of the gas diffusion layer (GDL) decreases. Too much assembly pressure causes GDL to destroy; therefore, the optimal assembly pressure is significant to obtain the highest performance from fuel cell. By using the results of this study, optimum fuel cell design and operating condition parameters can be predicted accordingly.

  11. Changes to Irradiation Conditions of VVER-1000 Surveillance Specimens Resulting from Fuel Assemblies with Greater Fuel Height

    Directory of Open Access Journals (Sweden)

    Panferov Pavel

    2016-01-01

    Full Text Available The goal of the work was to obtain experimental data on the influence of newtype fuel assemblies with higher fuel rods on the irradiation conditions of surveillance specimens installed on the baffe of VVER-1000. For this purpose, two surveillance sets with container assemblies of the same design irradiated in reactors with different fuel assemblies in the core were investigated. Measurements of neutron dosimeters from these sets and retrospective measurements of 54Mn activity accumulated in each irradiated specimen allow a detailed distribution of the fast neutron flux in the containers to be obtained. Neutron calculations have been done using 3D discrete ordinate code KATRIN. On the basis of the obtained results, a change of the lead factor due to newtype fuel assemblies was evaluated for all types of VVER-1000 container assemblies.

  12. Changes to Irradiation Conditions of VVER-1000 Surveillance Specimens Resulting from Fuel Assemblies with Greater Fuel Height

    Science.gov (United States)

    Panferov, Pavel; Kochkin, Viacheslav; Erak, Dmitry; Makhotin, Denis; Reshetnikov, Alexandr; Timofeev, Andrey

    2016-02-01

    The goal of the work was to obtain experimental data on the influence of newtype fuel assemblies with higher fuel rods on the irradiation conditions of surveillance specimens installed on the baffe of VVER-1000. For this purpose, two surveillance sets with container assemblies of the same design irradiated in reactors with different fuel assemblies in the core were investigated. Measurements of neutron dosimeters from these sets and retrospective measurements of 54Mn activity accumulated in each irradiated specimen allow a detailed distribution of the fast neutron flux in the containers to be obtained. Neutron calculations have been done using 3D discrete ordinate code KATRIN. On the basis of the obtained results, a change of the lead factor due to newtype fuel assemblies was evaluated for all types of VVER-1000 container assemblies.

  13. Irradiation experiment of the first and the second OGL-1 fuel assemblies

    International Nuclear Information System (INIS)

    The gas loop, OGL-1, installed in JMTR is the only facility for irradiation of full-size fuel rods under a simulated condition of the experimental reactor. One fuel assembly is irradiated every year in OGL-1. This report describes the irradiation experiment fof the first and the second fuel assemblies. Each fuel assembly has three fuel rods inserted in a graphite block. Irradiation periods are two reactor cycles for the first and four reactor cycles for the second, maximum burnups are 4500 MWD/T for the first and 8700 MWD/T for the second and estimated maximum compact temperatures are 13800C for the first and 13700C for the second fuel assemblies. Post irradiation examination revealed a bowing of the sleeves to a small extent. It was also revealed that the fuel compacts have neither cracks nor defect and that the particles were not broken during irradiation. (author)

  14. Integrated Three-Voltage-Booster DC-DC Converter to Achieve High Voltage Gain with Leakage-Energy Recycling for PV or Fuel-Cell Power Systems

    OpenAIRE

    Chih-Lung Shen; Hong-Yu Chen; Po-Chieh Chiu

    2015-01-01

    In this paper, an integrated three-voltage-booster DC-DC (direct current to direct current) converter is proposed to achieve high voltage gain for renewable-energy generation systems. The proposed converter integrates three voltage-boosters into one power stage, which is composed of an active switch, a coupled-inductor, five diodes, and five capacitors. As compared with conventional high step-up converters, it has a lower component count. In addition, the features of leakage-energy recycling...

  15. Cap assembly for a bundled tube fuel injector

    Energy Technology Data Exchange (ETDEWEB)

    LeBegue, Jeffrey Scott; Melton, Patrick Benedict; Westmoreland, III, James Harold; Flanagan, James Scott

    2016-04-26

    A cap assembly for a bundled tube fuel injector includes an impingement plate and an aft plate that is disposed downstream from the impingement plate. The aft plate includes a forward side that is axially separated from an aft side. A tube passage extends through the impingement plate and the aft plate. A tube sleeve extends through the impingement plate within the tube passage towards the aft plate. The tube sleeve includes a flange at a forward end and an aft end that is axially separated from the forward end. A retention plate is positioned upstream from the impingement plate. A spring is disposed between the retention plate and the flange. The spring provides a force so as to maintain contact between at least a portion of the aft end of the tube sleeve and the forward side of the aft plate.

  16. Fuel assembly leakage, Unit 4, Cycle 22, Paks NPP

    International Nuclear Information System (INIS)

    At the beginning of Cycle 22, Unit 4, Paks NPP the Iodine isotopes activity concentrations raised irregularly in the water of the primary circuit. Analysis supposed that from 1 to 10 fuel rods in one or more newly loaded follower assemblies had lost their integrity. Due to the fact it was not necessary to shut down the reactor, but at the end of the cycle sipping tests were performed for the entire core to find out the facts using a telescope sipping device supplied by H and B Co., Germany. This paper describes the circumstances of the emergence of the problem, the operational inspection and limitation rules in the Paks NPP, the theoretical analysis to estimate the scope and kind of the problem, the sipping device and the measurement/evaluation methods applied for the practical tests, fulfilment the tests, the results and their evaluation and the conclusions regarding the event. (authors)

  17. Analysis of Coolant Mixing in the VVER-1000 Fuel Assembly

    International Nuclear Information System (INIS)

    Thermohydrodynamics is numerically analyzed for turbulent flows through a VVER-1000 rod bundle. The steady-state form of the Reynolds-averaged Navier-Stokes (RANS), mass, energy and turbulence equations was discretized and solved using ANSYS CFX 12.1. The flow domain and the heated tube diameter are 700 mm and 9.1 mm, respectively. The standard Launder and Spalding K-ε turbulence model was used in the current numerical analysis since it is practically useful and converges well for the complex turbulent flow in the subchannel geometry. Results indicate that the behavior of the pressure drop in the rod bundle between the spacer grids is linear, and the spacer grids deviated from linearity and increased the pressure drops. The normalized Nusselt numbers decreased with the increasing axial distance downstream of the spacer grid. The coupled thermohydrodynamics and neutronics would indeed constitute more rigorous analyses of the fuel assembly design, heat transfer correlation and mixing coefficient for VVER

  18. Reactor vessel model flow tests for 145-fuel assembly core

    International Nuclear Information System (INIS)

    Hydraulic tests on a one-sixth-scale model of a two-loop pressurized water reactor with 145 fuel assemblies are described. Core inlet and outlet flow distributions and reactor vessel pressure drop were investigated. The core inlet flow distribution was developed to be independent of the flow conditions in the inlet annulus. A flow distribution system, consisting of several flow splitters in the inlet annulus and a spherical plate flow distributor in the lower head region, was developed to obtain a symmetric and stable core inlet flow distribution. A minimum core inlet flow factor of 0.99 was established in the core. Reactor vessel unrecoverable pressure drops were measured on the model to predict losses that will occur in the prototype

  19. Analysis of Coolant Mixing in the VVER-1000 Fuel Assembly

    Energy Technology Data Exchange (ETDEWEB)

    Nazififard, Mohammad; Suh, Kune Y. [Seoul National Univ., Seoul (Korea, Republic of); Suh, Kune Y. [PHILOSOPHIA, Seoul (Korea, Republic of); Nazififard, Mohammad; Nematollahi, Mohammadreza; Jafarpour, Khosrow [Shiraz Univ., Shiraz (Indonesia)

    2012-03-15

    Thermohydrodynamics is numerically analyzed for turbulent flows through a VVER-1000 rod bundle. The steady-state form of the Reynolds-averaged Navier-Stokes (RANS), mass, energy and turbulence equations was discretized and solved using ANSYS CFX 12.1. The flow domain and the heated tube diameter are 700 mm and 9.1 mm, respectively. The standard Launder and Spalding K-{epsilon} turbulence model was used in the current numerical analysis since it is practically useful and converges well for the complex turbulent flow in the subchannel geometry. Results indicate that the behavior of the pressure drop in the rod bundle between the spacer grids is linear, and the spacer grids deviated from linearity and increased the pressure drops. The normalized Nusselt numbers decreased with the increasing axial distance downstream of the spacer grid. The coupled thermohydrodynamics and neutronics would indeed constitute more rigorous analyses of the fuel assembly design, heat transfer correlation and mixing coefficient for VVER.

  20. Modernization of OIK unit for fabrication of mixed-oxide fuel, vibrocompacted fuel elements and fuel-containing assemblies of BN-600 reactor using plutonium of weapon quality

    International Nuclear Information System (INIS)

    In the framework of participation in international project of weapon plutonium utilization modernization of the technological complex for fabrication of granulated fuel, vibrocompacted fuel elements and fuel-containing assemblies is realizing. Taking into account domestic and foreign experience of MOX-fuel fabrication different versions of equipment are examined

  1. Spent fuel assembly hardware: Characterization and 10 CFR 61 classification for waste disposal

    International Nuclear Information System (INIS)

    Consolidation of spent fuel is under active consideration as the US Department of Energy plans to dispose of spent fuel. During consolidation, the fuel pins are removed from an intact fuel assembly and repackaged into a more compact configuration. After repackaging, approximately 30 kg of residual spent fuel assembly hardware per assembly remains that is also radioactive and requires disposal. Understanding the nature of this secondary waste stream is critical to designing a system that will properly handle, package, store, and dispose of the waste. This report presents a methodology for estimating the radionuclide inventory in irradiated spent fuel hardware. Ratios are developed that allow the use of ORIGEN2 computer code calculations to be applied to regions that are outside the fueled region. The ratios are based on the analysis of samples of irradiated hardware from spent fuel assemblies. The results of this research are presented in three volumes. In Volume 1, the development of scaling factors that can be used with ORIGEN2 calculations to estimate activation of spent fuel assembly hardware is documented. The results from laboratory analysis of irradiated spent-fuel hardware samples are also presented in Volume 1. In Volumes 2 and 3, the calculated flux profiles of spent nuclear fuel assemblies are presented for pressurized water reactors and boiling water reactors, respectively. The results presented in Volumes 2 and 3 were used to develop the scaling factors documented in Volume 1. 5 refs., 4 figs., 21 tabs

  2. Spent fuel assembly hardware: Characterization and 10 CFR 61 classification for waste disposal

    International Nuclear Information System (INIS)

    Consolidation of spent fuel is under active consideration as the US Department of Energy plans to dispose of spent fuel as required by the Nuclear Waste Policy Act of 1982. During consolidation, the fuel pins are removed from an intact fuel assembly and repackaged into a more compact configuration. After repackaging, approximately 30 kg of residual spent fuel assembly hardware per assembly that is also radioactive and required disposal. Understanding the nature of this secondary waste stream is critical to designing a system that will properly handle, package, store, and dispose of the waste. This report presents a methodology for estimating the radionuclide inventory in irradiated spent fuel hardware. Ratios are developed that allow the use of ORIGEN2 computer code calculations to be applied to regions that are outside the fueled region. The ratios are based on the analysis of samples of irradiated hardware from spent fuel assemblies. The results of this research are presented in three volumes. In Volume 1, the development of scaling factors that can be used with ORIGEN2 calculations to estimate activation of spent fuel assembly hardware is documented. The results from laboratory analysis of irradiated spent-fuel hardware samples are also presented in Volume 1. In Volume 2 and 3, the calculated flux profiles of spent nuclear fuel assemblies are presented for pressurized water reactors and boiling water reactors, respectively. The results presented in Volumes 2 and 3 were used to develop the scaling factors documented in Volume 1

  3. PS Booster - Festive colloquium

    CERN Document Server

    2012-01-01

    A festive colloquium will be held to celebrate the 40th anniversary of the PS Booster on Friday, 28 September at 2 p.m. in the CERN council chamber. The meeting will be open to everybody. Read more on the PS Booster in the CERN Bulletin and in the CERN Courier.

  4. Single rod leak detection and repair of leaking or damaged fuel assemblies

    International Nuclear Information System (INIS)

    In some circumstances, it is necessary to perform rework operations on some fuel assemblies in order to make them reusable in reactors, movable, transportable or consistent with fuel reprocessor specifications, depending on the plant utility policy. These rework operations are of two types: - Those which consist in restoring the leak tightness of the fuel assemblies. They are made after a series of tests allowing the localization of the failed fuel rods: at first, overall leak detection is provided by monitoring primary coolant activity during reactor operation; then, during refuelling, leaking assemblies are identified by subjecting each of the assemblies scheduled for reloading to a sipping test; finally individual leaking fuel rods are singled out before the defective assemblies can be repaired, i.e. failed rods can be replaced. - Those which involve replacement of part of or the whole assembly structure (combined or not with replacement of failed fuel rods). In order to meet these two needs for rework operations, FRAGEMA has developed a full range of test and tooling systems for detecting single leaking rods in irradiated fuel assemblies and for restoring fuel assemblies to be used in PWR nuclear power plants. As an illustration of the means available, two of these systems are described

  5. Separator assembly for use in spent-nuclear-fuel shipping cask. [Patent application

    Energy Technology Data Exchange (ETDEWEB)

    Bucholz, J.A.

    1981-04-24

    A separator assembly for use in a spent-nuclear-fuel shipping cask has a honeycomb-type wall structure defining parallel cavities for holding nuclear fuel assemblies. Tubes formed of an effective neutron-absorbing material are embedded in the wall structure around each of the cavities and provide neutron flux traps when filled with water.

  6. Combined fuel assembly and thimble plug gripper for a nuclear reactor

    International Nuclear Information System (INIS)

    This invention relates to an apparatus for loading and unloading a fuel assembly into and from the core of a nuclear reactor and for removing and inserting control rod guide thimble plugs from and into the fuel assembly during a reactor refueling operation in substantially less time than that presently required and in a more reliable, safe and efficient manner. (UK)

  7. A spray cooling technique for spent fuel assembly stored in pool

    International Nuclear Information System (INIS)

    For the safety of spent nuclear fuel assemblies stored in storage pool in the extreme condition where the water is lost completely, a passive spray cooling technique was designed, and its effectiveness has been validated by a functional experiment. The spray cooling characteristics of the spent fuel assembly have also been investigated by the experiment.

  8. Modeling the effect of engine assembly mass on engine friction and vehicle fuel economy

    Science.gov (United States)

    An, Feng; Stodolsky, Frank

    An analytical model is developed to estimate the impact of reducing engine assembly mass (the term engine assembly refers to the moving components of the engine system, including crankshafts, valve train, pistons, and connecting rods) on engine friction and vehicle fuel economy. The relative changes in frictional mean effective pressure and fuel economy are proportional to the relative change in assembly mass. These changes increase rapidly as engine speed increases. Based on the model, a 25% reduction in engine assembly mass results in a 2% fuel economy improvement for a typical mid-size passenger car over the EPA Urban and Highway Driving Cycles.

  9. BROOKHAVEN: Booster boost

    International Nuclear Information System (INIS)

    After three months of intensive dedicated machine studies, Brookhaven's new Booster accelerated 5 x 1013 protons over four cycles, about 85% of the design intensity. This was made possible by careful matching of Linac beam into the Booster and by extensive resonance stop band corrections implemented during Booster acceleration. The best single cycle injection into the AGS Alternating Gradient Synchrotron was 1.14 x 1013 protons from the Booster. 1.05 x 1013 protons were kept in the AGS, a 92% combined efficiency of extraction, transfer, and injection. The maximum injected 1994 shutdown period, enabling the 1994 physics run to make use of the full Booster intensity and go for the stated AGS objective of 4x1013 protons per pulse

  10. Buoyancy-driven flow excursions in fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Laurinat, J.E.; Paul, P.K.; Menna, J.D. [Westinghouse Savannah River Company, Aiken, SC (United States)

    1995-09-01

    A power limit criterion was developed for a postulated Loss of Pumping Accident (LOPA) in one of the recently shut down heavy water production reactors at the Savannah River Site. These reactors were cooled by recirculating heavy water moderator downward through channels in cylindrical fuel tubes. Powers were limited to safeguard against a flow excursion in one of more of these parallel channels. During-full-power operation, limits safeguarded against a boiling flow excursion. At low flow rates, during the addition of emergency cooling water, buoyant forces reverse the flow in one of the coolant channels before boiling occurs. As power increased beyond the point of flow reversal, the maximum wall temperature approaches the fluid saturation temperature, and a thermal excursion occurs. The power limit criterion for low flow rates was the onset of flow reversal. To determine conditions for flow reversal, tests were performed in a mock-up of a fuel assembly that contained two electrically heated concentric tubes surrounded by three flow channels. These tests were modeled using a finite difference thermal-hydraulic code. According to code calculations, flow reversed in the outer flow channel before the maximum wall temperature reached the local fluid saturation temperature. Thermal excursions occurred when the maximum wall temperature approximately equaled the saturation temperature. For a postulated LOPA, the flow reversal criterion for emergency cooling water addition was more limiting than the boiling excursion criterion for full power operation. This criterion limited powers to 37% of the limiting power for previous long-term reactor operations.

  11. DNB analysis with mechanistic models for PWR fuel assemblies

    International Nuclear Information System (INIS)

    In order to predict the DNB heat flux of PWR fuel assemblies and the critical power of BWR fuel bundles, the Boiling Transition Analysis Code CAPE' has been developed in the IMPACT project. The CAPE code for PWR includes three analysis modules, subchannel analysis module, three-dimensional two-phase flow analysis module, and DNB evaluation module. The subchannel module uses drift-flux model to identify the hottest subchannel. The three-dimensional two-phase flow analysis module uses nonhomogeneous and nonequilibrium two fluid model to analyze the detailed three-dimensional two-phase flow behaviors such as void distribution. For DNB heat flux prediction, the DNB evaluation module uses the Weisman model in which is a mechanistic DNB evaluation model. This paper describes the analysis models, analysis techniques and the results of validation by rod bundle test analysis. To date, the average difference between calculated and 11 measured values was -0.6% with a standard deviation of 7.0%. (author)

  12. ROSA-IV Large Scale Test Facility (LSTF) system description for second simulated fuel assembly

    International Nuclear Information System (INIS)

    The ROSA-IV Program's Large Scale Test Facility (LSTF) is a test facility for integral simulation of thermal-hydraulic response of a pressurized water reactor (PWR) during small break loss-of-coolant accidents (LOCAs) and transients. In this facility, the PWR core nuclear fuel rods are simulated using electric heater rods. The simulated fuel assembly which was installed during the facility construction was replaced with a new one in 1988. The first test with this second simulated fuel assembly was conducted in December 1988. This report describes the facility configuration and characteristics as of this date (December 1988) including the new simulated fuel assembly design and the facility changes which were made during the testing with the first assembly as well as during the renewal of the simulated fuel assembly. (author)

  13. Calculation of the dose rate for the spent fuel assembly from VVER-1000 type reactors

    International Nuclear Information System (INIS)

    The dose rate from the spent fuel assembly presented in this paper was obtained by using of the SAS2H control module in Version 4.3 of the SCALE code system. The calculations were performed for a VVER-1000 type fuel assembly with 4.4wt% initial 235U enrichment and different storage time. The calculations may be divided into two steps: calculation of the characteristics - concentrations of fission products, actinides, neutron and gamma source emissions, and determination of the equivalent dose rate on the surface and at some distance from a single assembly. SAS2H computes neutron and gamma source spectrum and evaluates equivalent dose rates from spent nuclear fuel assembly using a 1-D transport shielding analysis. Obtained results presented in this paper are necessary step in shielding analysis of spent fuel casks and interim storage facilities. (author) Keywords: equivalent dose rate, spent fuel assembly, VVER-1000

  14. Sensitivity and System Response of Pin Power Peaking in VVER-1000 Fuel Assembly Using TSUNAMI-2D

    Science.gov (United States)

    Frybort, J.

    2014-04-01

    Pin power peaking in a VVER-1000 fuel assembly and its sensitivity and uncertainty was analyzed by TSUNAMI-2D code. Several types of fuel assemblies were considered. They differ in number and position of gadolinium fuel pins. The calculations were repeated for several fuel compositions obtained by fuel depletion calculation. The results are quantified sensitivity data, which can be used for enrichment profiling.

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

  17. Analysis of dismantling possibility and unloading efforts of fuel assemblies from core of WWER

    International Nuclear Information System (INIS)

    The computation methods of optimal dismantling sequence of fuel assemblies (FA) from core of WWER after different operating periods and accident conditions are considered. The algorithms of fuel dismantling sequence are constructed both on the basis of analysis of mutual spacer grid overlaps of adjacent fuel assemblies and numerical structure analysis of efforts required for FA removal as FA heaving from the core. Computation results for core dismantling sequence after 3-year operating period and LB LOCA are presented in the paper

  18. Neutronic and thermohydraulic characteristics of a new breeding thorium–uranium mixed SCWR fuel assembly

    International Nuclear Information System (INIS)

    Highlights: • A new Th–U mixed fuel assembly for SCWR has been introduced and investigated. • Neutronic and thermohydraulic characteristics of the new assembly have been studied. • The new fuel assembly satisfies design rules of SCWR. • The introduced fuel assembly can fulfill the sustainable breeding Th–U cycle. • The new fuel assembly also has advantages with respect to lower generation of minor actinides and reactor safety. - Abstract: The exploitation of thorium fuel is a promising way to overcome the pressing problems of nuclear fuel supply, nuclear waste and nuclear proliferation. In this paper, a novel conceptual design of a breeding thorium–uranium (Th–U) mixed fuel assembly in SCWR is proposed, which is aimed to achieve the breeding ratio bigger than 1.0, so as to fulfill the sustainable breeding thorium–uranium cycle. Through the calculations of neutronics and neutronic/thermohydraulic (N–T) coupling, the results indicate that the introduced conceptual design of a breeding Th–U mixed fuel assembly in SCWR satisfies design rules of SCWR, with considerable advantages with respect to breeding performance, lower minor actinide generation and reactor safety

  19. CFD study on coolant mixing in VVER-440 fuel rod bundles and fuel assembly heads

    International Nuclear Information System (INIS)

    A CFD model of VVER-440 fuel assembly heads was developed based on the technical documentation of a full-scale test facility built in the Kurchatov Institute, Russia. Steady-state and transient calculations were performed to validate the model with a measurement set. Effects of the spatial resolution, turbulence models, difference schemes and different inlet boundary conditions were investigated. Inlet boundary conditions were determined with both the COBRA subchannel code and a fuel rod bundle CFD model that was built for this special purpose. The results were compared against experimental data. The sensitivity studies showed that a grid of about 8 million cells, high resolution scheme and BSL Reynolds stress model are suitable sets to provide accurate prediction for the signal of the in-core thermocouple. The best prediction was achieved with transient calculation using inlet boundary conditions generated with the CFD fuel rod bundle model. The results indicated that the coolant mixing is intensive but not perfect in the assembly head. Besides, the significant role of the outflow from the central tube was also proven. The transient runs revealed relatively large temperature fluctuations near the in-core thermocouple housing.

  20. Measurement of uranium and plutonium content in a fuel assembly using the RPI spent fuel assay device

    International Nuclear Information System (INIS)

    In this paper we report measurements of the significant parameters, the sensitivities of the slowing-down-time assay device to the fissile contents of a boiling water reactor (BWR) assembly mock-up of fresh fuel

  1. Drop impact analysis of plate-type fuel assembly in research reactor

    International Nuclear Information System (INIS)

    In this research, a drop impact analysis of a fuel assembly in a research reactor is carried out to determine whether the fuel plate integrity is maintained in a drop accident. A fuel assembly drop accident is classified based on where the accident occurs, i.e., inside or outside the reactor, since each occasion results in a different impact load on the fuel assembly. An analysis procedure suitable for each drop situation is systematically established. For an accident occurring outside the reactor, the direct impact of a fuel assembly on the pool bottom is analyzed using implicit and explicit approaches. The effects of the key parameters, such as the impact velocity and structural damping ratios, are also studied. For an accident occurring inside the reactor, the falling fuel assembly may first hit the fixing bar at the upper part of the standing fuel assembly. To confirm the fuel plate integrity, a fracture of the fixing bar should be investigated, since the fixing bar plays a role in protecting the fuel plate from the external impact force. Through such an analysis, the suitability of an impact analysis procedure associated with the drop situation in the research reactor is shown.

  2. Variants of WWER-440 in-core fuel management with low burned assemblies reuse

    International Nuclear Information System (INIS)

    Not concerning technical and regulatory problems of the unburned fuel transition from stopped to working WWER-440 reactors in the paper are considered only ideas of the methodical approach to mixed core loadings formation with purpose to reach the residual fuel design burnup at preservation standard operational reactor parameters. With this purpose a common model is developed on the basis of a steady state mode of reloading with three cycles of use for fuel assemblies with enrichment of 3.6%. So at the end of each cycle 1/3 assemblies are removed from reactor core and each assembly with enrichment of 3.6% works 3 cycles. In the presented model it is supposed also to use fuel assemblies with enrichment of 2.4% working 2 cycles and with enrichment of 1.6% - operated one cycle. In the model it is supposed also that for this time of operation the assemblies of the corresponding enrichment reach the design burnup

  3. Mark C (17 x 17) fuel assembly. Research and development. Revision 2

    International Nuclear Information System (INIS)

    The research and development programs related to the B and W Mark C (17 x 17) fuel assembly are described. The programs described are designed to provide data for analytical models, confirm predictions, and verify the adequacy of the design. A general description of the Mark C assembly is presented along with comparisons to the Mark B (15 x 15) fuel assembly (the Mark C fuel assembly is an extension of B and W's existing Mark B fuel assembly technology and the planned testing programs draw on the experience gained in the Mark B development program). The report is organized in three main test areas: (1) mechanical and hydraulic, (2) critical heat flux, and (3) rod swelling and burst. The test program scopes and objectives are discussed along with the test methods

  4. Critical Experiments to Determine Amount of U-235 in Research Reactor Fuel Assemblies

    International Nuclear Information System (INIS)

    Seven different critical core configurations of the IRR1 have been used to determine experimentally the reactivity changes caused by interchanging different fuel assemblies in the same core position(1'2). The data obtained with the first four critical configurations, Cores 1, 2, 3, and 4, are analyzed in this report; the other three will be analyzed and reported in a later paper. The different fuel assemblies (highly enriched MTR type fuel assemblies) interchanged in these cores varied in their burnup or loss in U-235, thus changing the core keff w i t h each interchange. The k∞ of a fuel assembly is a function of its U-235 inventory. The fuel assemblies interchange forced the control blade to alter its height in order to compensate for the change in core reactivity. The control blade reactivity worth as a function of its height was measured by period measurements so the reactivity changes due to the fuel assemblies interchange could be determined. Several different fuel assemblies in each core were measured in this manner. This experimental data can be calculated using reactor physics codes

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

    International Nuclear Information System (INIS)

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

  6. Estimation of the nuclear fuel assembly eigenfrequencies in the probability sense

    Directory of Open Access Journals (Sweden)

    Zeman V.

    2014-12-01

    Full Text Available The paper deals with upper and lower limits estimation of the nuclear fuel assembly eigenfrequencies, whose design and operation parameters are random variables. Each parameter is defined by its mean value and standard deviation or by a range of values. The gradient and three sigma criterion approach is applied to the calculation of the upper and lower limits of fuel assembly eigenfrequencies in the probability sense. Presented analytical approach used for the calculation of eigenfrequencies sensitivity is based on the modal synthesis method and the fuel assembly decomposition into six identical revolved fuel rod segments, centre tube and load-bearing skeleton linked by spacer grids. The method is applied for the Russian TVSA-T fuel assembly in the WWER1000/320 type reactor core in the Czech nuclear power plant Temelín.

  7. Development in the manufacture of fuel assembly components at Nuclear Fuel Complex

    International Nuclear Information System (INIS)

    The integrity of the fuel bundle and pellet-clad mechanical and chemical interaction (PCMCI) is the major limiting factor in achieving high burn up in thermal as well as fast reactors. Zircaloy based fuel bundle used for Indian pressurized heavy water reactor consists of number of components such as fuel clad tube, end cap bearing pad and spacer pad. These tubular, bar and sheet components are manufactured at Nuclear Fuel Complex using a series of thermomechanical processes involving hot and cold working with intermediate heat treatment. This paper is aimed at bringing out recent advances in NFC in the manufacture of fuel assembly components. Zircaloy based double clad tube adopting co-extrusion route followed by cold pilgering was successfully produced for its potential usage for high burnup in advance thermal reactors such as Advanced Heavy Water Reactors, This paper also includes process modifications carried out in the manufacture of clad tube and end cap components based on in-depth metallurgical studies. A radial forging process was established for primary breakdown of arc melted ingot which allows for better soundness and homogeneous microstructure. Manufacturing route of bar components for end caps was suitably modified by adopting only barrel straightening to minimize the residual stress and thereby increasing the recovery appreciably. NFC also supplies clad tube for fast breeder reactors where limiting factor for burn up are void swelling and fuel-clad interaction. In view of this, advance claddings such as P/M based 9Cr - Oxide Dispersion strengthened (ODS) steel clad and Zirconium lined T91 (9Cr-1 Mo) steel double clad have been successfully produced. Zirconium lined T91 (9Cr-1 Mo) double clad tubes required was successfully produced by adopting the method of co-pilgering, as a candidate material for clad tubes of Fast Breeder Reactors. (author)

  8. Cask containing method for spent fuel assembly and subcriticality measuring device for a cask containing system

    International Nuclear Information System (INIS)

    An area for a spent fuel storage pool is sectioned into an ordinary rack area for disposing spent fuel assemblies taken out from a reactor core and a preliminary storage rack area having the same constitution as a cask for containing spent fuel assemblies. Preceding to cask-containment, the spent fuel assemblies are temporarily transferred once in the preliminary storing rack area from the ordinary rack area to ensure subcriticality and then contained in casks. In addition, those fuels having a higher burn-up degree are disposed coaxially to the central portion and those having not higher burn-up degree are disposed at the outer circumferential portion. The spent fuel assemblies can surely be contained in the casks, or the process of containing the spent fuel assemblies to the casks or the subcriticality after the containment can be evaluated thereby capable of further ensuring the subcriticality. The spent fuel assemblies can be transferred or stored safely and reliably at a good efficiency. (N.H.)

  9. Effects of the chemical decontamination on the component parts of the ATR fuel assembly

    International Nuclear Information System (INIS)

    The chemical decontamination technique has been developed in order to remove the crud adhering to the surface of the components constructing the primary coolant system, as a part of the measure to decrease the exposure in the annual inspection. The technique has been already applied to the prototype reactor 'Fugen', in the core of which the fuel assemblies were not loaded. The chemical decontamination, for the core in which the fuel assemblies are loaded, has been planned for the purpose of improving the utilization factor. It is necessary to confirm, through the test before putting the plan into practice, that the decontamination reagent does not exert a bad influence upon the components constructing the fuel assembly. This report describes the test results which have been carried out so as to investigate the influence of the reagent on the components constructing the fuel assembly. The outline of the results is as follows: (1) The susceptibility to stress corrosion cracking of the chemical decontamination treatment and the residual decontamination reagent on the components constructing the fuel assembly is low enough. (2) The chemical decontamination treatment and the residual decontamination reagent do not exert a bad influence upon the integrity of the fuel assembly concerning the fuel rod holding function of the spacer and the characteristics of the fretting wear caused on the fuel claddings. (author)

  10. Effects of the chemical decontamination on the component parts of the ATR fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Kodaka, Kozo; Tendo, Masayuki; Sugawara, Masayuki; Koike, Mitsutaka; Matsuda, Masanori; Endo, Kazuo; Iba, Toshi [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1996-03-01

    The chemical decontamination technique has been developed in order to remove the crud adhering to the surface of the components constructing the primary coolant system, as a part of the measure to decrease the exposure in the annual inspection. The technique has been already applied to the prototype reactor `Fugen`, in the core of which the fuel assemblies were not loaded. The chemical decontamination, for the core in which the fuel assemblies are loaded, has been planned for the purpose of improving the utilization factor. It is necessary to confirm, through the test before putting the plan into practice, that the decontamination reagent does not exert a bad influence upon the components constructing the fuel assembly. This report describes the test results which have been carried out so as to investigate the influence of the reagent on the components constructing the fuel assembly. The outline of the results is as follows: (1) The susceptibility to stress corrosion cracking of the chemical decontamination treatment and the residual decontamination reagent on the components constructing the fuel assembly is low enough. (2) The chemical decontamination treatment and the residual decontamination reagent do not exert a bad influence upon the integrity of the fuel assembly concerning the fuel rod holding function of the spacer and the characteristics of the fretting wear caused on the fuel claddings. (author)

  11. Integrated Radiation Transport and Nuclear Fuel Performance for Assembly-Level Simulations

    Energy Technology Data Exchange (ETDEWEB)

    Clarno, Kevin T [ORNL; Hamilton, Steven P [ORNL; Philip, Bobby [ORNL; Berrill, Mark A [ORNL; Sampath, Rahul S [ORNL; Allu, Srikanth [ORNL; Pugmire, Dave [ORNL; Dilts, Gary [Los Alamos National Laboratory (LANL); Banfield, James E [ORNL

    2012-02-01

    The Advanced Multi-Physics (AMP) Nuclear Fuel Performance code (AMPFuel) is focused on predicting the temperature and strain within a nuclear fuel assembly to evaluate the performance and safety of existing and advanced nuclear fuel bundles within existing and advanced nuclear reactors. AMPFuel was extended to include an integrated nuclear fuel assembly capability for (one-way) coupled radiation transport and nuclear fuel assembly thermo-mechanics. This capability is the initial step toward incorporating an improved predictive nuclear fuel assembly modeling capability to accurately account for source-terms and boundary conditions of traditional (single-pin) nuclear fuel performance simulation, such as the neutron flux distribution, coolant conditions, and assembly mechanical stresses. A novel scheme is introduced for transferring the power distribution from the Scale/Denovo (Denovo) radiation transport code (structured, Cartesian mesh with smeared materials within each cell) to AMPFuel (unstructured, hexagonal mesh with a single material within each cell), allowing the use of a relatively coarse spatial mesh (10 million elements) for the radiation transport and a fine spatial mesh (3.3 billion elements) for thermo-mechanics with very little loss of accuracy. In addition, a new nuclear fuel-specific preconditioner was developed to account for the high aspect ratio of each fuel pin (12 feet axially, but 1 4 inches in diameter) with many individual fuel regions (pellets). With this novel capability, AMPFuel was used to model an entire 17 17 pressurized water reactor fuel assembly with many of the features resolved in three dimensions (for thermo-mechanics and/or neutronics), including the fuel, gap, and cladding of each of the 264 fuel pins; the 25 guide tubes; the top and bottom structural regions; and the upper and lower (neutron) reflector regions. The final, full assembly calculation was executed on Jaguar using 40,000 cores in under 10 hours to model over 162

  12. Dendritic assembly of gold nanoparticles during fuel-forming electrocatalysis.

    Science.gov (United States)

    Manthiram, Karthish; Surendranath, Yogesh; Alivisatos, A Paul

    2014-05-21

    We observe the dendritic assembly of alkanethiol-capped gold nanoparticles on a glassy carbon support during electrochemical reduction of protons and CO2. We find that the primary mechanism by which surfactant-ligated gold nanoparticles lose surface area is by taking a random walk along the support, colliding with their neighbors, and fusing to form dendrites, a type of fractal aggregate. A random walk model reproduces the fractal dimensionality of the dendrites observed experimentally. The rate at which the dendrites form is strongly dependent on the solubility of the surfactant in the electrochemical double layer under the conditions of electrolysis. Since alkanethiolate surfactants reductively desorb at potentials close to the onset of CO2 reduction, they do not poison the catalytic activity of the gold nanoparticles. Although catalyst mobility is typically thought to be limited for room-temperature electrochemistry, our results demonstrate that nanoparticle mobility is significant under conditions at which they electrochemically catalyze gas evolution, even in the presence of a high surface area carbon and binder. A careful understanding of the electrolyte- and polarization-dependent nanoparticle aggregation kinetics informs strategies for maintaining catalyst dispersion during fuel-forming electrocatalysis. PMID:24766431

  13. Preliminary Design of U-Mo Alloy Dispersion Fuel Assembly

    Institute of Scientific and Technical Information of China (English)

    2008-01-01

    <正>As a kind of new type fuel for research reactor, high density U-Mo alloy dispersion fuel which will substitute current fuel in the future is being studied and developed by RERTR. There are two characteristics

  14. Structural integrity assessment and stress measurement of CHASNUPP-1 fuel assembly

    OpenAIRE

    Waseem; Murtaza Ghulam; Siddiqui Ashfaq Ahmad; Akhtar Syed Waseem

    2016-01-01

    Fuel assembly of the PWR nuclear power plant is a long and flexible structure. This study has been made in an attempt to find the structural integrity of the fuel assembly (FA) of Chashma Nuclear Power Plant-1 (CHASNUPP-1) at room temperature in air. The non-linear contact and structural tensile analysis have been performed using ANSYS 13.0, in order to determine the fuel assembly (FA) elongation behaviour as well as the location and values of the stress intensity and stresses developed in ax...

  15. Hydraulic Design of the CARA Fuel Assembly for Atucha-I

    International Nuclear Information System (INIS)

    In this paper a hydraulic model of the CARA fuel assembly within the Atucha I fuel channel is developed. Besides, a experimental test running in the CBP low pressure loop have been designed.This model is used for design purpose of the assembly system such as the whole channel pressure drop remains the same that it is at the present.It is observed that choosing the right thickness and hole surface of the assembly system, it is possible tune up the CARA pressure drop, releases the azimuth alignment condition on the fuel element neighbors

  16. Numerical simulation of gas dynamics and heat exchange tasks in fuel assemblies of the nuclear reactors

    International Nuclear Information System (INIS)

    This report presents a PC-based program for solution gas dynamics and heat exchange mathematical tasks in fuel assemblies of the fast-neutron nuclear reactors. A fuel assembly consisting of bulk heat-generating elements, which are integrated together by the system of supply and pressure manifolds, is examined. Spherical heat-generating microelements, which contain nuclear fuel, are pulled into the heat-generating elements. Gaseous coolant proceed from supply manifolds to heat-generating elements, where it withdraws the nuclear reaction heat and assembles in pressure manifolds

  17. Power distribution control in BN-600 reactor by method of gamma-scanning of fuel assemblies

    International Nuclear Information System (INIS)

    Acceptability, convenience and reliability of γ-scanning of fuels assembles at fast reactor NPP have been analyzed and demonstrated. Error of the procedure is amount 3-6% for different fuel assemblies. The procedure is recommended as optimum one for the constructed BN-800 and perspective fast reactors. Findings allow conclusion on the accordance of BN-600 fuel assemblies powers with design parameters and insignificant (in the limits of observation accuracy) changing power distribution in new BN-600 01M2 reactor core. Experimental procedure is modernized and optimized, three cycles of measurement are realized, new experimental data on the character of radial and axial distributions of neutron field are received

  18. BROOKHAVEN: Booster commissioned

    International Nuclear Information System (INIS)

    The construction and first commissioning phase of the Booster synchrotron to inject into Brookhaven's veteran Alternating Gradient Synchrotron (AGS) were completed last year. Scheduled to come into operation this year, the new Booster will extend the research capabilities AGS, and with its ability to accelerate partially stripped heavy ions will play an essential role in the chain of accelerators serving the Relativistic Heavy Ion Collider (RHIC)

  19. Differential Die-Away Instrument: Report on Fuel Assembly Mock-up Measurements with Neutron Generator

    Energy Technology Data Exchange (ETDEWEB)

    Goodsell, Alison Victoria [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Swinhoe, Martyn Thomas [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Henzl, Vladimir [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Rael, Carlos D. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Desimone, David J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2014-09-18

    Fresh fuel experiments for the differential die-away (DDA) project were performed using a DT neutron generator, a 15x15 PWR fuel assembly, and nine 3He detectors in a water tank inside of a shielded cell at Los Alamos National Laboratory (LANL). Eight different fuel enrichments were created using low enriched (LEU) and depleted uranium (DU) dioxide fuel rods. A list-mode data acquisition system recorded the time-dependent signal and analysis of the DDA signal die-away time was performed. The die-away time depended on the amount of fissile material in the fuel assembly and the position of the detector. These experiments were performed in support of the spent nuclear fuel Next Generation Safeguards Initiative DDA project. Lessons learned from the fresh fuel DDA instrument experiments and simulations will provide useful information to the spent fuel project.

  20. Structural integrity assessment and stress measurement of CHASNUPP-1 fuel assembly

    Directory of Open Access Journals (Sweden)

    Waseem

    2016-01-01

    Full Text Available Fuel assembly of the PWR nuclear power plant is a long and flexible structure. This study has been made in an attempt to find the structural integrity of the fuel assembly (FA of Chashma Nuclear Power Plant-1 (CHASNUPP-1 at room temperature in air. The non-linear contact and structural tensile analysis have been performed using ANSYS 13.0, in order to determine the fuel assembly (FA elongation behaviour as well as the location and values of the stress intensity and stresses developed in axial direction under applied tensile load of 9800 N or 2 g being the fuel assembly handling or lifting load [Y. Zhang et al., Fuel assembly design report, SNERDI, China, 1994]. The finite element (FE model comprises spacer grids, fuel rods, flexible contacts between the fuel rods and grid's supports system and guide thimbles with dash-pots and flow holes, in addition to the spot welds between spacer grids and guide thimbles, has been developed using Shell181, Conta174 and Targe170 elements. FA is a non-straight structure. The actual behavior of the geometry is non-linear due to its curvature or design tolerance. It has been observed that fuel assembly elongation values obtained through FE analysis and experiment [SNERDI Tech. Doc., Mechanical strength and calculation for fuel assembly, Technical Report, F3.2.1, China, 1994] under applied tensile load are comparable and show approximately linear behaviors. Therefore, it seems that the permanent elongation of fuel assembly may not occur at the specified load. Moreover, the values of stresses obtained at different locations of the fuel assembly are also comparable with the stress values of the experiment determined at the same locations through strain gauges. Since the results of both studies (analytical and experimental are comparable, therefore, validation of the FE methodology is confirmed. The stress intensity of the FE model and maximum stresses developed along the guide thimbles in axial direction are

  1. Single PWR spent fuel assembly heat transfer data for computer code evaluations

    International Nuclear Information System (INIS)

    The descriptions and results of two separate heat transfer tests designed to investigate the dry storage of commercial PWR spent fuel assemblies are presented. Presented first are descriptions and selected results from the Fuel Temperature Test performed at the Engine Maintenance and Disassembly facility on the Nevada Test Site. An actual spent fuel assembly from the Turkey Point Unit Number 3 Reactor with a decay heat level of 1.17 KW, was installed vertically in a test stand mounted canister/liner assembly. The boundary temperatures were controlled and the canister backfill gases were alternated between air, helium and vacuum to investigate the primary heat transfer mechanisms of convection, conduction and radiation. The assembly temperature profiles were experimentally measured using installed thermocouple instrumentation. Also presented are the results from the Single Assembly Heat Transfer Test designed and fabricated by Allied General Nuclear Services, under contract to the Department of Energy, and ultimately conducted by the Pacific Northwest Laboratory. For this test, an electrically heated 15 x 15 rod assembly was used to model a single PWR spent fuel assembly. The electrically heated model fuel assembly permitted various ''decay heat'', levels to be tested; 1.0 KW and 0.5 KW were used for these tests. The model fuel assembly was positioned within a prototypic fuel tube and in turn placed within a double-walled sealed cask. The complete test assembly could be positioned at any desired orientation (horizontal, vertical, and 250 from horizontal for the present work) and backfilled as desired (air, helium, or vacuum). Tests were run for all combinations of ''decay heat,'' backfill, and orientation. Boundary conditions were imposed by temperature controlled guard heaters installed on the cask exterior surface

  2. Dynamic characteristics and design criteria of fuel rod assemblies in a baffle jet flow

    International Nuclear Information System (INIS)

    During recent refuelling operations of a PWR type nuclear power plant it was found that several fuel assemblies located at baffle joint were damaged. It was assumed that the damage had been caused by severe vibration of fuel rods induced by coolant leakage through baffle joint. Model testing was con-ducted to identify the vibration mechanism and to obtain the safety criteria for fuel assemblies in a baffle jet flow. Fuel rods are long beams supported along their length by seven grid assemblies. Those prototype rods were simulated as single span simply supported beams. Model assemblies are 4 x 4 and 5 x 4 bundles of simply supported beams with a pitch ratio of 1.33. Flow tests were carried out in a water loop of 40 GPM. It was found that rod assemblies in a jet flow experience large amplitude vibration caused by jet induced instability. The stability boundary of rod assemblies is determined to be Vc/fD=2.3 √(D/h)(msub(o) deltasub(o))/(rho D2). Based on the stability boundary provided, safety limit of baffle gap is calculated as to be 1.6 x 10-3 in. The effect of the position of fuel rod assemblies relative to the baffle joint was investigated. And it was found that the susceptibility of rod assemblities to vibration increases as the stand off distance shortens. (Author)

  3. New knowledge and experiences of flow induced fretting in PWR fuel assemblies

    International Nuclear Information System (INIS)

    One of the most important demands for a fretting free nuclear fuel assembly design is to verify that the fuel assembly is able to resist any fretting damages during its lifetime. The reactor experience and laboratory investigations have made clear to distinguish between 2 types of flow induced excitation: Turbulent Excitation (TE) and Self Induced Excitation (SIE). The former is the 'Conventional' flow excitation which is traditionally verified through integral endurance test; the latter, for which the rod (or assembly) vibration influences significantly the flow excitation and which may lead to a resonance for given flow conditions, requires a specific test protocol which allows to identify this phenomenon safely. For this purpose, an improved experimental approach to investigate the behavior of a fuel assembly design towards SIE has been developed: an integral fuel assembly flow test is performed using an assembly 'optimized' in terms of vibratory characteristics and a specially designed loop (PETER loop), able to simulate different flow boundary conditions, such as different inlet flow conditions, neighboring fuel assembly influence (different grid or bottom nozzle pressure loss coefficients, intermediate flow mixing grids), fuel assembly bow, baffle shaking or vibrations of neighboring fuel assembly. During this flow test, the resulting fuel rod and fuel assembly vibration behavior are measured under many parameterized flow conditions and the verifications of 'non resonance' or instability are done. Moreover, the maximum rod vibration amplitudes are recorded for the conditions of a flow bounding those expected for the reactor. In a second step, the design can be verified towards TE through two different approaches which are considered as complementary or alternative: 1)The conventional approach which consists of introducing a fuel assembly prototype into a flow loop for 1000 hours under bounding flow conditions and then to inspect the fuel rods for potential

  4. GNS experience of CASTOR cask loading for storage and transport of spent fuel assemblies

    International Nuclear Information System (INIS)

    With over 25 years of experience in the design, manufacturing, assembly and loading of CASTOR registered casks, GNS is one of the worldwide leading suppliers of casks for the transport and storage of spent fuel assemblies as well as for canisters with vitrified high level wastes. GNS products are used at around 30 sites worldwide for a wide range of inventories from pressurized and boiling water reactor fuels (PWR and BWR), thorium high-temperature reactor fuels (THTR) and research reactor fuels to vitrified high-active wastes (HAW) from reprocessing plants

  5. Burn up Analysis for Fuel Assembly Unit i n a Pressurized Heavy Water CANDU Reactor

    International Nuclear Information System (INIS)

    MCNPX code has been used for modeling a nd simulation of an assembly of CANDU Fuel bundle . The assembly is composed of a heterogeneous lattice of 37-element natural Uranium fuel, heavy water moderator and coolant. The fuel bundle is burnt in normal operation conditions of CANDU reactors. The effective multiplication factor (Keff ) of the bundle is calculated as a function of fuel burnup. The flux and power distribution are determined. Comparing t he concentrations of both Uranium and Plutonium isotopes are analyzed in the bundle. The results of the present model with the results of a benchmark problem, a good agreement was found PWR

  6. Fuel nozzle assembly for use as structural support for a duct structure in a combustor of a gas turbine engine

    Science.gov (United States)

    Wiebe, David J; Fox, Timothy A

    2015-03-31

    A fuel nozzle assembly for use in a combustor apparatus of a gas turbine engine. An outer housing of the fuel nozzle assembly includes an inner volume and provides a direct structural connection between a duct structure and a fuel manifold. The duct structure defines a flow passage for combustion gases flowing within the combustor apparatus. The fuel manifold defines a fuel supply channel therein in fluid communication with a source of fuel. A fuel injector of the fuel nozzle assembly is provided in the inner volume of the outer housing and defines a fuel passage therein. The fuel passage is in fluid communication with the fuel supply channel of the fuel manifold for distributing the fuel from the fuel supply channel into the flow passage of the duct structure.

  7. Proving test on thermal-hydraulic performance of BWR fuel assemblies

    International Nuclear Information System (INIS)

    Nuclear Power Engineering Corporation (NUPEC) has conducted a proving test for thermal-hydraulic performance of BWR fuel (high-burnup 8 x 8, 9 x 9) assemblies entrusted by the Ministry of Economy, Trade and Industry (NUPEC-TH-B Project). The high-burnup 8 x 8 fuel (average fuel assembly discharge burnup: about 39.5 GWd/t), has been utilized from 1991. And the 9 x 9 fuel (average fuel assembly discharge burnup: about 45 GWd/t), has started to be used since 1999. There are two types (A-type and B-type) of fuel design in 9 x 9 fuel assembly. Using an electrically heated test assembly which simulated a BWR fuel bundle on full scale, flow induced vibration, pressure drop, critical power under steady state condition and post-boiling transition (post-BT) tests were carried out in an out-of pile test facility that can simulate the high pressure and high temperature conditions of BWRs. This paper completed the results of 9 x 9 fuel combined with the previously reported results of high-burnup 8 x 8 fuel. As a result of NUPEC-TH-B Project, the validity of the current BWR thermal-hydraulic design method was confirmed and the reliability of BWR thermo-hydraulic fuel performance was demonstrated. Based on the test data, a new correlation of the estimation of fuel rod vibration amplitude, new post-BT heat transfer and rewet correlations for the estimation of fuel rod surface temperature were developed. (author)

  8. The computational fluid dynamics (CFD) modeling of coolant flow in fuel assembly - current results and their application

    International Nuclear Information System (INIS)

    The paper summarises present works performed by VUJE Inc in the field of computational fluid dynamics (CFD) simulation of coolant flow in fuel assemblies used in the WWER-440 reactors. These works are extension of the previous calculations presented on Acer Symposium 2004 and their aim is the better understanding of coolant flow in different parts of fuel assembly. While the previous CFD simulations were focused on study of individual parameters influences on the coolant flow mixing in the upper part of fuel assembly (pin-wise power distribution, geometry of fuel assembly, by-pass flow and coolant flow in central tube), the present works are directed more to verification of previous calculations and joining of individual effects on coolant flow mixing in the upper part of fuel assembly, mainly in the thermocouple position. The main purpose of the CFD simulations of coolant flow in fuel assembly is to increase accuracy of temperature measurements placed at the output of fuel assemblies (Authors)

  9. The underwater coincidence counter for plutonium measurements in mixed-oxide fuel assemblies manual

    Energy Technology Data Exchange (ETDEWEB)

    G. W. Eccleston; H. O. Menlove; M. Abhold; M. Baker; J. Pecos

    1999-05-01

    This manual describes the Underwater Coincidence Counter (UWCC) that has been designed for the measurement of plutonium in mixed-oxide (MOX) fuel assemblies prior to irradiation. The UWCC uses high-efficiency {sup 3}He neutron detectors to measure the spontaneous-fission and induced-fission rates in the fuel assembly. Measurements can be made on MOX fuel assemblies in air or underwater. The neutron counting rate is analyzed for singles, doubles, and triples time correlations to determine the {sup 240}Pu effective mass per unit length of the fuel assembly. The system can verify the plutonium loading per unit length to a precision of less than 1% in a measurement time of 2 to 3 minutes. System design, components, performance tests, and operational characteristics are described in this manual.

  10. VANTAGE 5 PWR fuel assembly demonstration program at Virgil C. Summer nuclear station

    International Nuclear Information System (INIS)

    VANTAGE 5 is an improved PWR fuel product designed and manufactured by Westinghouse Electric Corporation. The VANTAGE 5 fuel design features integral fuel burnable absorbers, intermediate flow mixer grids, axial blankets, high burnup capability, and a reconstitutable top nozzle. A demonstration program for this fuel design commenced in late 1984 in cycle 2 of the Virgil C. Summer Nuclear Station. Objectives for VANTAGE 5 fuel are reduced fuel cycle costs, better core operating margins, and increased design and operating flexibility. Inspections of the VANTAGE 5 demonstration assemblies are planned at each refueling outage

  11. Improvement of the design model for SMART fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Zee, Sung Kyun; Yim, Jeong Sik

    2001-04-01

    A Study on the design improvement of the TEP, BEP and Hoddown spring of a fuel assembly for SMART was performed. Cut boundary Interpolation Method was applied to get more accurate results of stress and strain distribution from the results of the coarse model calculation. The improved results were compared with that of a coarse one. The finer model predicted slightly higher stress and strain distribution than the coarse model, which meant the results of the coarse model was not converged. Considering that the test results always showed much less stress than the FEM and the location of the peak stress of the refined model, the pressure stress on the loading point seemed to contribute significantly to the stresses. Judging from the fact that the peak stress appeared only at the local area, the results of the refined model were considered enough to be a conservative prediction of the stress levels. The slot of the guide thimble screw was ignored to get how much thickness of the flow plate can be reduced in case of optimization of the thickness and also cut off the screw dent hole was included for the actual geometry. For the BEP, the leg and web were also included in the model and the results with and without the leg alignment support were compared. Finally, the holddown spring which is important during the in-reactor behavior of the FA was modeled more realistic and improved to include the effects of the friction between the leaves and the loading surface. Using this improved model, it was possible that the spring characteristics were predicted more accurate to the test results. From the analysis of the spring characteristics, the local plastic area controled the characteristics of the spring dominantly which implied that it was necessary for the design of the leaf to be optimized for the improvement of the plastic behavior of the leaf spring.

  12. Improvement of the design model for SMART fuel assembly

    International Nuclear Information System (INIS)

    A Study on the design improvement of the TEP, BEP and Hoddown spring of a fuel assembly for SMART was performed. Cut boundary Interpolation Method was applied to get more accurate results of stress and strain distribution from the results of the coarse model calculation. The improved results were compared with that of a coarse one. The finer model predicted slightly higher stress and strain distribution than the coarse model, which meant the results of the coarse model was not converged. Considering that the test results always showed much less stress than the FEM and the location of the peak stress of the refined model, the pressure stress on the loading point seemed to contribute significantly to the stresses. Judging from the fact that the peak stress appeared only at the local area, the results of the refined model were considered enough to be a conservative prediction of the stress levels. The slot of the guide thimble screw was ignored to get how much thickness of the flow plate can be reduced in case of optimization of the thickness and also cut off the screw dent hole was included for the actual geometry. For the BEP, the leg and web were also included in the model and the results with and without the leg alignment support were compared. Finally, the holddown spring which is important during the in-reactor behavior of the FA was modeled more realistic and improved to include the effects of the friction between the leaves and the loading surface. Using this improved model, it was possible that the spring characteristics were predicted more accurate to the test results. From the analysis of the spring characteristics, the local plastic area controled the characteristics of the spring dominantly which implied that it was necessary for the design of the leaf to be optimized for the improvement of the plastic behavior of the leaf spring

  13. Nondestructive verification with minimal movement of irradiated light-water-reactor fuel assemblies

    International Nuclear Information System (INIS)

    Nondestructive verification of irradiated light-water reactor fuel assemblies can be performed rapidly and precisely by measuring their gross gamma-ray and neutron signatures. A portable system measured fuel assemblies with exposures ranging from 18.4 to 40.6 GWd/tU and with cooling times ranging from 1575 to 2638 days. Differences in the measured results for side or corner measurements are discussed. 25 figures, 20 tables

  14. Positioning of Nuclear Fuel Assemblies by Means of Image Analysis on Tomographic Data

    OpenAIRE

    Troeng, Mats

    2004-01-01

    A tomographic measurement technique for nuclear fuel assemblies has been developed at the Department of Radiation Sciences at Uppsala University [1]. The technique requires highly accurate information about the position of the measured nuclear fuel assembly relative to the measurement equipment. In experimental campaigns performed earlier, separate positioning measurements have therefore been performed in connection to the tomographic measurements. In this work, another positioning approach h...

  15. Determination of weight factors for VVER-440 fuel assemblies with burnable poison

    International Nuclear Information System (INIS)

    Detailed CFD model for the head parts of the VVER-440 fuel assemblies with burnable poison has been developed. The coolant mixing was analyzed in some typical assemblies with this model and the signals of the in-core thermocouples above the selected assemblies were calculated. The investigations pointed out that the mixing is intensive in these assembly heads but the coolant is not perfectly mixed before reaching the thermocouples. Significant differences between the outlet average coolant temperatures and the thermocouple signals were revealed in the case of the fresh fuels. These deviations can cause about 6% underestimations in the online monitored assembly powers unless a proper correction is introduced. The coolant mixing was also studied by means of numerical tracers and weight factors of selected rod bundle regions for the in-core thermocouple were determined. Using these weight factors and the outlet enthalpies of the assemblies' subchannels, the thermocouple signals can be corrected. (authors)

  16. A CAREM fuel assembly prototype construction in order to verify its mechanical design using hydrodynamic testing

    International Nuclear Information System (INIS)

    The scope of this paper is to describe the activities of several Groups from three Atomic Centers (C. A. Bariloche, C. A. Ezeiza and C. A. Constituyentes), involved in the manufacturing of a CAREM fuel assembly prototype. The Design Group (UAIN-CAB) carried out the fuel assembly engineering. Cladding components were constructed by the Special Alloys Pilot Factory (UAMCN-CAE). Engineering Group (UACN-CAC) manufactured the parts to be processed, resorting to qualified suppliers. Elastic spacers were completely designed and constructed by this Group, and fuel rods, control rods, guide tubes and spacers were also welded here. Research Reactors Fuels Group (UACN-CAC) carried out the dimensional control of the elaborated parts, while Postirradiation Testing Group (UACN-CAC) performed the assembling of the fuel element. This paper also refers to the design and development of special equipment and devices, all of them required for the prototype construction. (author)

  17. Hinkley Point CAGR - fuel assembly vibration and charge route interaction during on-load refuelling

    International Nuclear Information System (INIS)

    A key feature of the UK advanced gas cooled reactor system is the ability to refuel while producing power. To achieve this the fuel and plug units are built up into a long slender fuel assembly, and an associated charge route constructed for each fuel assembly location in core. Currently, flow induced vibration of the fuel assembly limits the operating power during refuelling, with the disadvantages of lost power production, and reactor power cycling. The test work, analysis and subsequent theoretical appraisals carried out on the Hinkley Point B reactor systems (Hunterston, Heysham II and Torness being of equivalent design) with two specific aims: are described firstly to assess the impact velocities of the fuel against the charge-route wall, and secondly to consider possible methods of reducing these velocities. (author)

  18. Application of PLUTO Test Facility for U. S. NRC Licensing of a Fuel Assembly

    International Nuclear Information System (INIS)

    The fuel assembly of the PLUS-7 loaded in the APR-1400 follows the same schedule. Meanwhile, In July 1998, the U.S. NRC adopted a research plan to address the effects of high burnup from a Loss of Coolant Accident (LOCA). From these programs, several important technical findings for rule revision were obtained. Based on the technical findings, the U. S. NRC has amended the 10 CFR 50.46 which will be proclaimed sooner or later. Through the amendment, a LOCA analysis on the fuel assembly has to show the safety at both a fresh and End of Life (EOL) state. The U. S. NRC has already required EOL effects on seismic/LOCA performance for a fuel assembly since 1998. To obtain U.S NRC licensing of a fuel assembly, based on the amendment of 10CFR50.46, a LOCA analysis of the fuel assembly has to show safety both fresh and EOL states. The proper damping factor of the fuel assembly measured at the hydraulic test loop for a dynamic model in a LOCA and a seismic analysis code are at least required. In this paper, we have examined the damping technologies and compared the test facility of PLUTO with others in terms of performance. PLUTO has a better performance on the operating conditions than any others

  19. Application of PLUTO Test Facility for U. S. NRC Licensing of a Fuel Assembly

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Dongseok; Shin, Changhwan; Lee, Kanghee; Kang, Heungseok [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    The fuel assembly of the PLUS-7 loaded in the APR-1400 follows the same schedule. Meanwhile, In July 1998, the U.S. NRC adopted a research plan to address the effects of high burnup from a Loss of Coolant Accident (LOCA). From these programs, several important technical findings for rule revision were obtained. Based on the technical findings, the U. S. NRC has amended the 10 CFR 50.46 which will be proclaimed sooner or later. Through the amendment, a LOCA analysis on the fuel assembly has to show the safety at both a fresh and End of Life (EOL) state. The U. S. NRC has already required EOL effects on seismic/LOCA performance for a fuel assembly since 1998. To obtain U.S NRC licensing of a fuel assembly, based on the amendment of 10CFR50.46, a LOCA analysis of the fuel assembly has to show safety both fresh and EOL states. The proper damping factor of the fuel assembly measured at the hydraulic test loop for a dynamic model in a LOCA and a seismic analysis code are at least required. In this paper, we have examined the damping technologies and compared the test facility of PLUTO with others in terms of performance. PLUTO has a better performance on the operating conditions than any others.

  20. Development of mechanical test techniques on the irradiated grid elements in PWR fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, Sang Bok; Kim, Do Sik; Baik, Seung Jai; Choo, Yong Sun; Baek, Sang Youl; Ryu, Woo Seok [Korea Atomic Energy Research Institute, Daejon (Korea, Republic of); Ha, Dong Keun; Seo, Jeong Min [Korea Nuclear Fuel Company, Daejon (Korea, Republic of)

    2008-11-15

    The mechanical properties of the parts of a nuclear fuel assembly are degraded during the operation of the reactor, through the mechanism of irradiation damage. The properties changes of the parts of the fuel assembly should be quantitatively estimated to ensure the safety of the fuel assembly and rod during the operation. The test techniques developed in this paper are used to produce the irradiation data of the grid 1x1 cell, the inner/outer strip of the grid and the welded part. The specimens were irradiated in the CT test hole of HANARO of a 30MW thermal output at 300.deg.C. From the spring tests of mid grid 1x1 cell and grid plate, the irradiation effects can be examined. The irradiation effects on the irradiation growth also were occurred. The buckling load of mid grid 1x1 cell does not change with a neutron irradiation. From the tensile tests, the strengths increased but the elongations decreased due to an irradiation. From the tensile tests, the strengths increased but the elongations decreased due to an irradiation. the tensile test and microstructure examination of the spot and fillet welded parts are performed for the evaluation of an irradiation effects. Through these tests of components, the essential data on the fuel assembly design could be obtained. These results will be used to update the irradiation behavior databases, to improve the performance of fuel assembly, and to predict the service life of the fuel assembly in a reactor.

  1. Vibration Monitoring Using Fiber Optic Sensors in a Lead-Bismuth Eutectic Cooled Nuclear Fuel Assembly

    Directory of Open Access Journals (Sweden)

    Ben De Pauw

    2016-04-01

    Full Text Available Excessive fuel assembly vibrations in nuclear reactor cores should be avoided in order not to compromise the lifetime of the assembly and in order to prevent the occurrence of safety hazards. This issue is particularly relevant to new reactor designs that use liquid metal coolants, such as, for example, a molten lead-bismuth eutectic. The flow of molten heavy metal around and through the fuel assembly may cause the latter to vibrate and hence suffer degradation as a result of, for example, fretting wear or mechanical fatigue. In this paper, we demonstrate the use of optical fiber sensors to measure the fuel assembly vibration in a lead-bismuth eutectic cooled installation which can be used as input to assess vibration-related safety hazards. We show that the vibration characteristics of the fuel pins in the fuel assembly can be experimentally determined with minimal intrusiveness and with high precision owing to the small dimensions and properties of the sensors. In particular, we were able to record local strain level differences of about 0.2 μϵ allowing us to reliably estimate the vibration amplitudes and modal parameters of the fuel assembly based on optical fiber sensor readings during different stages of the operation of the facility, including the onset of the coolant circulation and steady-state operation.

  2. Vibration Monitoring Using Fiber Optic Sensors in a Lead-Bismuth Eutectic Cooled Nuclear Fuel Assembly.

    Science.gov (United States)

    De Pauw, Ben; Lamberti, Alfredo; Ertveldt, Julien; Rezayat, Ali; van Tichelen, Katrien; Vanlanduit, Steve; Berghmans, Francis

    2016-01-01

    Excessive fuel assembly vibrations in nuclear reactor cores should be avoided in order not to compromise the lifetime of the assembly and in order to prevent the occurrence of safety hazards. This issue is particularly relevant to new reactor designs that use liquid metal coolants, such as, for example, a molten lead-bismuth eutectic. The flow of molten heavy metal around and through the fuel assembly may cause the latter to vibrate and hence suffer degradation as a result of, for example, fretting wear or mechanical fatigue. In this paper, we demonstrate the use of optical fiber sensors to measure the fuel assembly vibration in a lead-bismuth eutectic cooled installation which can be used as input to assess vibration-related safety hazards. We show that the vibration characteristics of the fuel pins in the fuel assembly can be experimentally determined with minimal intrusiveness and with high precision owing to the small dimensions and properties of the sensors. In particular, we were able to record local strain level differences of about 0.2 μϵ allowing us to reliably estimate the vibration amplitudes and modal parameters of the fuel assembly based on optical fiber sensor readings during different stages of the operation of the facility, including the onset of the coolant circulation and steady-state operation. PMID:27110782

  3. Studies of residual heat release of the BN-600 reactor spent assemblies in a fuel cooling pond

    International Nuclear Information System (INIS)

    A method was developed, and a facility was manufactured for measuring residual heat release of spent fuel assemblies directly in the fuel pool. The operations involving rearrangement of spent fuel assemblies are performed using the standard hardware and technologies, which is the main advantage of the method. Thus, the design safety of handling operations is provided. Direct measurements of residual heat release of essential number of spent fuel assemblies of diverse types have been taken for the first time

  4. AllianceTM, the fuel assembly on the threshold of the third millennium

    International Nuclear Information System (INIS)

    Nuclear energy competitiveness has become a major challenge of the end of this century. In order to respond, Framatome has developed a new nuclear fuel assembly for an enhanced utilisation of existing PWRs, and also for the future European Pressurized Reactor (EPR). This development is the result of a joint Framatome and Framatome Cogema Fuels (FCF) strategy to propose today a fuel product meeting their customer's future needs. ALLIANCE is a fuel assembly designed for assembly burnup of at least 70 GWd/t. This allows the utilities to consider modes of operation of their reactors which were not technically accessible until now. ALLIANCE is based on an in depth analysis of the market needs of Framatome and Fragema, and also of FCF. ALLIANCE benefits from Framatome long term commitment in a large R and D program, which has provided significant outcomes such as new alloys, new components and new fuel assembly concepts. This R and D program allows direct access to all the CEA expertise and facilities. As a result, ALLIANCE is a fuel assembly with unmatched and totally demonstrated performance. ALLIANCE takes also advantage of the extended operational experience available through all the reactors supplied with Framatome fuel: number of irradiated assemblies reactor types operational conditions - sometimes very demanding. Framatome major customers have been directly involved in the development of this assembly. All the available operational feedback has been taken into account at the early stages of the ALLIANCE design. The main features of the ALLIANCE assembly are: cladding tubes and an assembly structure in M5 alloy, a mono-metallic mixing grid with enhanced thermo-hydraulic performance, the possibility to add mid span mixing grids, the Monobloc guide tube, the Trapper bottom nozzle and a new structure designed for assembly burnups of at least 70 GW d/t. The first ALLIANCE assemblies will be loaded in 1999 in an EdF reactor. In 2000, ALLIANCE assemblies will be loaded

  5. Improvement of a verification device for fresh fuel assemblies in a shipping container

    International Nuclear Information System (INIS)

    Fresh fuel shipping containers which are consisted with the outer wooden container and inner metal container are stored as stacking four or five stories with approximately 10 cm of space between each containers until the shipment to the reactor facility. Normally, two fresh fuel assemblies are placing in the inner container with the distance of approximately 4 cm side by side and the distance from the surface of the fuel assembly to the outer surface of the container is approximately 30 cm. To satisfy the safeguards requirements, the gross defect test have to be performed. Currently, the number of fuel assemblies in the container is verified by sliding the NaI(Tl) detector crossing the surface of the shipping container. However, capability to discrimination of the position of the fuel assembly based on changing the specific U235 gamma intensity by position is insufficient due to the lack of background shielding of the lead collimator. To solve this problem, improve the detection efficiency and energy resolution by introduce CdTe detector, also improve the discrimination capability of fuel position by strengthen of the shielding of the lead collimator. The outline of the development plan to establish the verification technique for the number of fuel assembly in the shipping container will be presented. (author)

  6. Irradiation experiments of the 6th-12th OGL-1 fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Hayashi, Kimio; Minato, Kazuo; Kobayashi, Fumiaki; Kikuchi, Hironobu; Fukuda, Kousaku; Kikuchi, Teruo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Saruta, Tohru; Kitajima, Toshio

    1994-10-01

    The Oarai Gas Loop-1, OGL-1, is an in-pile helium gas loop, installed in the Japan Materials Testing Reactor (JMTR), for irradiation of high-temperature gas-cooled reactor fuels at high pressure and temperature. The present report describes results of fabrication, irradiation and post-irradiation examinations (PIE) of the 6th-12th OGL-1 fuel assemblies. The 6th-8th assemblies used coated fuel particles produced by a small-scale fluidized bed. On the other hand, the 9th-12th assemblies used trial manufacturing fuels, produced with a large-scale fluidized bed for mass production of the fuel for the High Temperature Engineering Test Reactor (HTTR) being constructed. For the 9th assembly loaded with the first mass-product fuel, the fission gas release, R/B of {sup 88}Kr, was relatively high, 1.5x10{sup -5}, and various defects were observed in the ceramography of the irradiated coating layers. Afterwards, a decrease was achieved in the through-coating failure fractions at the fabrication. Correspondingly, the R/B of {sup 88}Kr for the 12th assembly was reduced to an excellent value of 2x10{sup -6}. Thus, the production technology and the irradiation performance of the HTTR design fuels were successfully demonstrated. (author).

  7. Irradiation experiments of the 6th-12th OGL-1 fuel assemblies

    International Nuclear Information System (INIS)

    The Oarai Gas Loop-1, OGL-1, is an in-pile helium gas loop, installed in the Japan Materials Testing Reactor (JMTR), for irradiation of high-temperature gas-cooled reactor fuels at high pressure and temperature. The present report describes results of fabrication, irradiation and post-irradiation examinations (PIE) of the 6th-12th OGL-1 fuel assemblies. The 6th-8th assemblies used coated fuel particles produced by a small-scale fluidized bed. On the other hand, the 9th-12th assemblies used trial manufacturing fuels, produced with a large-scale fluidized bed for mass production of the fuel for the High Temperature Engineering Test Reactor (HTTR) being constructed. For the 9th assembly loaded with the first mass-product fuel, the fission gas release, R/B of 88Kr, was relatively high, 1.5x10-5, and various defects were observed in the ceramography of the irradiated coating layers. Afterwards, a decrease was achieved in the through-coating failure fractions at the fabrication. Correspondingly, the R/B of 88Kr for the 12th assembly was reduced to an excellent value of 2x10-6. Thus, the production technology and the irradiation performance of the HTTR design fuels were successfully demonstrated. (author)

  8. Development status and research directions on the structural components of the fuel assembly

    International Nuclear Information System (INIS)

    Survey on the structural components of the state-of-the art of the PWR fuel assembly developed by various nuclear fuel vendors has been performed. As a result, some developmental directions and mechanical/structural basic technology to be established for these structural components have been drawn out. The developmental directions are as follows; The top end piece shall be designed in shape to reduce its height to accommodate the fuel rod growth for high burnup and to have a function for easy reconstitution of the fuel assembly. The bottom end piece shall be designed in shape to reduce its height to accommodate the fuel rod growth for high burnup and to have a function for easy reconstitution of the fuel assembly. The bottom end piece shall be designed in shape to reduce its height to accommodate the fuel rod growth for high burnup and to have a function of debris protection. The spacer grid shall be designed in shape to have a function of enhancing the thermal margin and maintaining the fuel rod integrity without fuel failure due to fuel rod fretting and vibration. The mechanical/structural basic technology which must be established is as follows; The stress analysis results shall comply with the stress criteria specified in the ASME code stress limits and the shape optimization technology shall be developed for the top/bottom end pieces. For the spacer grid cell, the nonlinear analysis model of the fuel rod and the analysis model on the flow-induced fuel rod vibration, and a study of the mechanism and a quantified model on the fuel rod fretting wear shall be developed. In addition, numerical analysis model to estimate the buckling strength of the spacer grid assembly shall be developed. Besides above technology, technology related the verification test should be developed. (author). 30 figs., 54 refs

  9. SICOM, An equipment for very accurate dimensional and corrosion inspection of irradiated fuel assemblies

    International Nuclear Information System (INIS)

    Nuclear fuel undergo evolution in its properties and characteristics as result of the operation in the nuclear reactors. Knowledge of this evolution is necessary to guaranty the behaviour and safety in the operation, as well as to optimize the economic performance of the fuel cycle. When, with the objective of improving the performance of fuel, some important modifications are made, it is necessary to prove the effect on the behaviour and evaluate the influence in each of the critical characteristics of the fuel assemblies. In this cases, it could be necessary to burn a reduced number of fuel assemblies of the new design (demonstration assemblies). During the demonstration process, some characterization of the behaviour of the fuel is made, usually at the end of each cycle. An important part of this characterization is made using NDE methods, applied at the nuclear plant during the refuelling outages. Within the Electrotechnical Research and Development Program (PIE), and with participation by IBERDROLA, TECNATOM and ENUSA, an inspection system (SICOM) has been developed for spent fuel assemblies from pressurized water plants, the aim being to check the following features: general condition, apply dimensional controls and measure the oxide layer on the peripheral fuel rods. This equipment was qualified at Tecnatom and Almaraz I NPP during the first quarter of 1995. Subsequently, in September, it was validated for the EDF P'4 plants at C.N.P.E. Belleville

  10. Measurement of gamma and neutron radiations inside spent fuel assemblies with passive detectors

    Czech Academy of Sciences Publication Activity Database

    Viererbl, L.; Lahodová, Z.; Voljanskij, A.; Klupák, V.; Koleška, M.; Cabalka, M.; Turek, Karel

    2011-01-01

    Roč. 652, č. 1 (2011), s. 90-93. ISSN 0168-9002 Institutional research plan: CEZ:AV0Z10480505 Keywords : fuel assembly * spent fuel * track detector Subject RIV: JF - Nuclear Energetics Impact factor: 1.207, year: 2011

  11. Analysis of Spent Fuel Assembly Thermal Behaviors in Boil-off Accident Scenarios

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hye-Min; Chun, Tae-Hyun; Kim, Sun-Ki [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    The spent fuel pool (SFP) accidents would occur due to many different postulated scenarios, for example a SBO (Station Black Out) at SFP storage or an attack from external factor. In this study, we focused on the SFP boil off accident and analyzed the thermal behaviors of spent fuels following this accident, using MELCOR 1.8.6. version. MELCOR, originally the severe accident code, has been developed to also be appropriate to the SFP accident. This paper provides the spent fuel heatup characteristics in terms of decay heat, water level and fuel arrangement. The SFP model is based on 17x17 PWR assembly designed by Westinghouse. Spent fuel coolability has been analyzed with single and 1x4 assembly MELCOR models in the case of boil-off accident. It was shown that the low powered spent fuel assembly could be more vulnerable in the partial loss of coolant inventory because of lack of steam cooling and more fuel being uncovered. In addition, it was found that minimum water level has to be maintained above half of assembly height so as not to experience fuel failure, which depends on decay heat power.

  12. Neutron metrology in the fuel assemblies of a material test reactor

    International Nuclear Information System (INIS)

    Results are presented of detailed thermal and fast neutron measurements performed in all fuel and control assemblies of the HFR in Petten. The results give information about deviations of a general shape of vertical thermal and fast fluence rate distributions due to material transitions in the reactor core and different control assembly settings. Further it is demonstrated that the ratio of fast and thermal fluence rate at the various monitor positions in the assemblies give useful information for the (relative) local burn-up of the fuel. (orig.)

  13. Determination of the Effectiveness of Control Rods in the VVER Reactor Fuel Assemblies

    International Nuclear Information System (INIS)

    The paper describes experiments done in homogeneous mock-ups of the fuel assemblies from the VVER Reactor (at one level of enrichment) to determine the effectiveness of absorbing systems comprising shim fuel assemblies or water cavities and of absorbing rods clad in jackets made of differing materials. The paper also gives data on some experiments that have been done in mock-ups of assemblies with differing levels of enrichment. These experiments make it possible to verify the methods used in calculation and to evaluate the prospects of using them for heterogeneous reactors. (author)

  14. Welding device for nuclear fuel assembly structure elements

    International Nuclear Information System (INIS)

    The device has two parallel assembling positions next to each other. The welding robot is carried by a carriage with displacement parallel to the guide tubes and has enough degrees of freedom to move from one assembling position to the other and have access to the structural elements

  15. Field tests and examinations in cell of French fuel assemblies

    International Nuclear Information System (INIS)

    This paper presents the field tests carried out on more than 70 assemblies and those realized at the hot laboratory on a whole assembly of 25 rods taken from pressurized water power plants operating in France: dimension examination; fission gas release; oxidation and hydridation of cans

  16. Integrated Radiation Transport and Nuclear Fuel Performance for Assembly-Level Simulations

    International Nuclear Information System (INIS)

    The Advanced Multi-Physics (AMP) Nuclear Fuel Performance code (AMPFuel) is focused on predicting the temperature and strain within a nuclear fuel assembly to evaluate the performance and safety of existing and advanced nuclear fuel bundles within existing and advanced nuclear reactors. AMPFuel was extended to include an integrated nuclear fuel assembly capability for (one-way) coupled radiation transport and nuclear fuel assembly thermo-mechanics. This capability is the initial step toward incorporating an improved predictive nuclear fuel assembly modeling capability to accurately account for source-terms, such as neutron flux distribution, coolant conditions and assembly mechanical stresses, of traditional (single-pin) nuclear fuel performance simulation. A novel scheme is introduced for transferring the power distribution from the Scale/Denovo (Denovo) radiation transport code (structured, Cartesian mesh with smeared materials within each cell) to AMPFuel (unstructured, hexagonal mesh with a single material within each cell), allowing the use of a relatively coarse spatial mesh (10 million elements) for the radiation transport and a fine spatial mesh (3.3 billion elements) for thermo-mechanics with very little loss of accuracy. With this novel capability, AMPFuel was used to model an entire 1717 pressurized water reactor fuel assembly with many of the features resolved in three dimensions (for thermo-mechanics and/or neutronics). A full assembly calculation was executed on Jaguar using 40,000 cores in under 10 hours to model over 160 billion degrees of freedom for 10 loading steps. The single radiation transport calculation required about 50% of the time required to solve the thermo-mechanics with a single loading step, which demonstrates that it is feasible to incorporate, in a single code, a high-fidelity radiation transport capability with a high-fidelity nuclear fuel thermo-mechanics capability and anticipate acceptable computational requirements. The

  17. Histories of spent-nuclear-fuel assemblies while at the E-MAD facility, December 1978-September 1982

    International Nuclear Information System (INIS)

    This report documents the handling and storage histories of seventeen spent pressurized water reactor (PWR) fuel assemblies while they have been at the Engine Maintenance, Assembly and Disassembly (E-MAD) facility on the Nevada Test Site (NTS) through September 30, 1982. This report also presents thermal histories of the fuel assemblies, including predictions of peak clad temperatures

  18. End-of-irradiation data report for the instrumented fuel assembly (IFA)-527

    International Nuclear Information System (INIS)

    This report presents data obtained during the irradiation of the six-rod instrumented fuel assembly (IFA)-527 in the Halden Boiling Water Reactor (HBWR), Halden, Norway. This assembly is the last in a series of US Nuclear Regulatory Commission (NRC)-sponsored tests to obtain data for the development and verification of steady-state fuel performance computer codes. IFA-527 contains five identical rods with high-density stable fuel pellets and 230-μm diametral gaps and one rod with similar fuel pellets but with a 60-μm diametral gap. All six rods were xenon-filled to simulate the effects of fission gas and to enhance the observable effects of fuel cracking and relocation on fuel temperatures. This report presents both pre- and postfailure data for IFA-527

  19. Fuel assembly mechanical behavior with respect to center guide tube welding points

    International Nuclear Information System (INIS)

    A fuel assembly with modified welding procedure between the center guide tube and sleeves on grid structures is proposed in the work, and some mechanical behavior of the fuel assembly is presented. While all sleeves on grid assemblies are welded to center guide tube in the present design, the modified fuel assembly with reduced welding is proposed. To evaluate mechanical performance of the design, lateral bending deflection test and lateral vibration test were performed in a test facility. Applying a lateral static load at 6-th grid, the fuel assembly deflection at each grid was measured in the lateral static test. In the vibration test, displacement signal at every grid is measured while the fuel assembly is excited by a dynamic shaker at the 6-th grid. Processing the measured signals, modal properties such as natural frequency, damping, and mode shapes can be found based on frequency response functions. The test results, comparing to a previous design with all girds welding to the center guide tube, do not show any notable deviation in the structural behaviors

  20. Transport of fresh MOX fuel assemblies for the MONJU initial core

    International Nuclear Information System (INIS)

    Transport of fresh MOX fuel assemblies for the prototype FBR MONJU initial core started in July 1992 and ended in March 1994. As many as 205 fresh MOX fuel assemblies (109 assemblies for an inner core, 91 assemblies for an outer core and 5 assemblies for testing) were transported in nine transport missions. The packaging for fuel assemblies, which has shielding and shock absorbing material inside, meets IAEA regulatory requirements for Type B(U) packaging including hypothetical accident conditions such as the 9 m drop test, fire test, etc. Moreover, this packaging design features such advanced technologies as high performance neutron shielding material and an automatic hold-down mechanism for the fuel assemblies. Every effort was made to carry out safe transport in conjunction with the cooperation of every competent organisation. This effort includes establishment of the transport control centre, communication training, and accompanying the radiation monitoring expert. No transport accident occurred during the transport and all the transport missions were successfully completed on schedule. (Author)

  1. Fuel assembly cooling: grapple seals in the FFTF/IEM cell

    International Nuclear Information System (INIS)

    High temperature, extremely severe environment seals are vital for adequate cooling flow when transferring high decay heat driver fuel assemblies with the Interim Examination and Maintenance Cell at the Fast Flux Test Facility. The data from a multi-million dollar fuel assembly experiment would be invalidated if internal temperatures were allowed to exceed normal maximum operating temperatures. Cooling is accomplished by drawing cell atmosphere through the assembly by means of a ducted grapple. For this to be effective, an adequate grapple/fuel assembly interface seal had to be found. A silicone rubber seal was designed and tested that is both compliant (for leak tightness) and capable of surviving in enviroments not normally considered compatible with such materials. 4 figs., 1 tab

  2. Improvement of operation efficiency for WWER-440 and WWER-1000 for TRIGON fuel assembly design features

    International Nuclear Information System (INIS)

    TRIGON 440 and TRIGON 1000 fuel assemblies and their assembly matching counterparts are described. Their role in increasing the efficiency of WWER reactors is stressed. Special attention is paid to their design features as well as calibrated means of predicting behaviour under irradiation from light water reactor core operation. They reduce the fuel cycle cost as a result of the reduced need for natural uranium which have to be enriched and of the smaller number of fuel assemblies which have to be fabricated, stored or reprocessed. The improved control assemblies bring comfort to the plant operator due to intrinsic progress in safety with respect to accidental situation, trouble-free behaviour and long time utilization in the reactor. 14 figs

  3. Thermohydrodynamic investigations of the grid spacers of nuclear reactor rod fuel assemblies

    International Nuclear Information System (INIS)

    The studies on the velocity fields on the fuel assembly aerodynamic model were carried out with an account of the fact, that the grid spacers effect the redistribution of thermohydraulic parameters by the assembly cross section. It made it possible to determine the mechanisms of the grids impact on mixing, which may lead to leveling the nonuniformity of the thermohydraulic parameters distribution, which is typical for rod fuel assemblies, and thereby to the critical heat flux growth. The study results made it possible to determine the causes of the grid spacers effect on the heat transfer crisis and identify the most effective types of grids, which may be considered as the heat exchange intensifiers, leading to significant expansion of the area of the fuel assembly noncritical operation

  4. Estimation of Porous Media Approach for Thermal Hydraulics of Nuclear Fuel Assembly

    International Nuclear Information System (INIS)

    In many CFD studies, porous media assumption has been often used for thermal hydraulics of nuclear fuel assembly, e.g., reactor core, storage cask, spent fuel pool and etc. and it could be applied extensively as shown in Fig. 1. However, the assumption could not predict the local phenomena in a subchannel or the mixing effect between subchannels and did not consider distribution of variables. This work validates the porous media approach in nuclear fuel assembly from two aspects, friction factor and averaged temperature and discusses about appropriate use of the porous media approach at the various fluid conditions. Commercial CFD code CFX 12.0 was used

  5. Methods and programs of thermal hydraulic calculations of fast reactor fuel assemblies

    International Nuclear Information System (INIS)

    The methods and computer codes for calculating the velocity and temperature distributions in fast reactor fuel assemblies are described and analyzed. Three levels of thermal hydraulic analysis of fuel element bundles can be distinguished, viz.: analysis of local characteristics (finite element method, finite difference method), subchannel analysis (lumped parameter method), and analysis of characteristics averaged over volumes (porous body model). The possibilities of the existing computer codes and methods are demonstrated and conclusions regarding the future development of methods of and codes for thermal hydraulic analysis of fuel assemblies are presented. (author). 102 figs., 17 tabs., 256 refs

  6. High fidelity analysis of BWR fuel assembly with COBRA-TF/PARCS and trace codes

    International Nuclear Information System (INIS)

    The growing importance of detailed reactor core and fuel assembly description for light water reactors (LWRs) as well as the sub-channel safety analysis requires high fidelity models and coupled neutronic/thermalhydraulic codes. Hand in hand with advances in the computer technology, the nuclear safety analysis is beginning to use a more detailed thermal hydraulics and neutronics. Previously, a PWR core and a 16 by 16 fuel assembly models were developed to test and validate our COBRA-TF/PARCS v2.7 (CTF/PARCS) coupled code. In this work, a comparison of the modeling and simulation advantages and disadvantages of modern 10 by 10 BWR fuel assembly with CTF/PARCS and TRACE codes has been done. The objective of the comparison is making known the main advantages of using the sub-channel codes to perform high resolution nuclear safety analysis. The sub-channel codes, like CTF, permits obtain accurate predictions, in two flow regime, of the thermalhydraulic parameters important to safety with high local resolution. The modeled BWR fuel assembly has 91 fuel rods (81 full length and 10 partial length fuel rods) and a big square central water rod. This assembly has been modeled with high level of detail with CTF code and using the BWR modeling parameters provided by TRACE. The same neutronic PARCS's model has been used for the simulation with both codes. To compare the codes a coupled steady state has be performed. (author)

  7. Design of a full scale model fuel assembly for full power production reactor flow excursion experiments

    International Nuclear Information System (INIS)

    A novel full scale production reactor fuel assembly model was designed and built to study thermal-hydraulic effects of postulated Savannah River Site (SRS) nuclear reactor accidents. The electrically heated model was constructed to simulate the unique annular concentric tube geometry of fuel assemblies in SRS nuclear production reactors. Several major design challenges were overcome in order to produce the prototypic geometry and thermal-hydraulic conditions. The two concentric heater tubes (total power over 6 MW and maximum heat flux of 3.5 MW/m2) (1.1E+6 BTU/(ft2hr)) were designed to closely simulate the thermal characteristics of SRS uranium-aluminum nuclear fuel. The paper discusses the design of the model fuel assembly, which met requirements of maintaining prototypic geometric and hydraulic characteristics, and approximate thermal similarity. The model had a cosine axial power profile and the electrical resistance was compatible with the existing power supply. The model fuel assembly was equipped with a set of instruments useful for code analysis, and durable enough to survive a number of LOCA transients. These instruments were sufficiently responsive to record the response of the fuel assembly to the imposed transient

  8. Heat transfer in a dry, horizontal LWR spent fuel assembly

    International Nuclear Information System (INIS)

    A series of experiments has been conducted to measure the steady-state temperature distribution through a dry 16 x 16 dummy PWR assembly lying horizontally in a close-fitting box. The experimental results have been analysed to obtain the pin cladding temperature distribution across the centre of the assembly. These measured temperatures have been compared against calculations with the computer code RIGG and good agreement has been obtained over a wide range of assembly powers and temperatures and with different filling gases. The experiments showed that natural convection inside the air-filled assembly had a significant effect upon the temperature profile. The temperatures in the lower half of the assembly were reduced and those in the upper half increased. These effects were reproduced in a simple RIGG model demonstration that these convection currents reduce the peak pin temperature slightly. The helium-filled assembly experienced no significant effects due to convection. The agreement between measured and calculated peak pin temperatures was, in general, very good, especially when the effects of convection were modelled. If a reduction of about 50C due to convection is assumed for the tests in which the assembly was air filled, then the agreement over all the tests between the measured and calculated peak pin temperatures was +- 70C. A significant part of this error is attributed to random errors in experimental power measurement and measurement of material properties

  9. Advanced fuel assemblies for economic and flexible operation of light water reactors

    International Nuclear Information System (INIS)

    Increasing competition in the electricity market sets up a corresponding competition between the different electricity producing technologies. This makes further improvements in the economics of nuclear power generation a vital item for the future of nuclear energy. Though the costs for development, design and fabrication of fuel assemblies contribute only about 10% to the fuel cycle costs, the design and the performance of the fuel assemblies considerably influences total electricity generation cost. By the recent creation of Framatome ANP the nuclear activities of Framatome and Siemens were combined into one company. In the past, both had made considerable achievements in the development of fuel assemblies and related services supporting the goal of safe and economic electricity generation by light water reactors. The examples described in this paper cover former Siemens products and experience. In the future, our combined experience bases will be an ideal platform to offer further substantial improvements to our customers. (author)

  10. Application of the ballooning analysis code MATARE on a generic PWR fuel assembly

    International Nuclear Information System (INIS)

    The MATARE (MAbel-TAlink-RElap) code is a new multi-pin deformation analysis code created through the dynamic coupling between the thermal-hydraulic code RELAP5 and multiple instances of the single-pin thermal-mechanics code MABEL. A multi-pin representation of different zones of a typical PWR fuel assembly under post-LOCA reflooding conditions was analysed including some of the most relevant features that characterise a typical nuclear reactor fuel assembly and evaluate their effect on the behaviour of the fuel rods under conditions leading to clad ballooning. The code was able to simulate the deformation of wide regions of a fuel assembly under reflood conditions and has shown how differences in pin pressure and the presence of rod with burnable poisons and control rod guide thimbles also contribute to a substantial incoherent ballooning in agreement with the experimental data. (author)

  11. Apparatus for removing and/or positioning fuel assemblies of a nuclear reactor

    International Nuclear Information System (INIS)

    Apparatus for positioning fuel assemblies of a nuclear reactor includes a control for a crane comprising a strain gauge connected to the crane line which raises and lowers the load. The signal from the strain gauge is compared with setpoints; which if the strain gauge signal exceeds a high-level setpoint, indicating that the movement of a fuel assembly is obstructed, the line drive is disabled. The line drive is also disabled if the strain gauge signal is less than a low-level setpoint, indicating that a fuel being deposited contacts the bottom of its slot or an obstruction. To preclude lateral movement of the fuel assembly suspended from the crane line, the traverse drive of the crane is disabled once the strain-gauge signal exceeds the low-level setpoint. The traverse drive can only be enabled after the strain-gauge signal is less than a slack-line setpoint. (author)

  12. A Parametric Study on the Thermal Hydraulic Design for an Annular Fuel Assembly

    Energy Technology Data Exchange (ETDEWEB)

    Shin, C. H.; Seo, K. W.; Chun, T. H. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2008-10-15

    Recently, MIT proposed an internally and externally cooled annular fuel for an advanced PWR which can endure a substantial power uprating. To apply this annular fuel in the conventional reactors such as OPR-1000, it is desirable to investigate its a structural compatibility for its reloading to operating PWR reactors of OPR-1000 as well as other compatibilities like the fuel to moderator ratio, amount of fissile material and coolant flow area. Conventional fuel assembly has a 16x16 solid rod array with four big guide tubes and one instrumentation tube. A 12x12 annular fuel assembly design which can meet the above compatibilities was proposed, which is structurally compatible with the existing internals of OPR-1000. Actually the advantage of an annular fuel comes from the fuel performance and thermal hydraulics. In the thermal hydraulic analysis, the mixing effect between the neighboring channels has to be carried out in a subchannel analysis. A subchannel analysis code, MATRA has been developed by KAERI. However, MATRA dose not have the capability to model both an internally and externally cooled annular fuel. A subchannel code, MATRA-AF which can be coupled to MATRA and can calculate the coolant flow distribution and heat transfer fraction in the internal and external subchannels has been developed. In this paper, the characteristics and the verification of the MATRA-AF are described. The effects of the thermal hydraulic parameters are estimated through a single fuel assembly.

  13. A Parametric Study on the Thermal Hydraulic Design for an Annular Fuel Assembly

    International Nuclear Information System (INIS)

    Recently, MIT proposed an internally and externally cooled annular fuel for an advanced PWR which can endure a substantial power uprating. To apply this annular fuel in the conventional reactors such as OPR-1000, it is desirable to investigate its a structural compatibility for its reloading to operating PWR reactors of OPR-1000 as well as other compatibilities like the fuel to moderator ratio, amount of fissile material and coolant flow area. Conventional fuel assembly has a 16x16 solid rod array with four big guide tubes and one instrumentation tube. A 12x12 annular fuel assembly design which can meet the above compatibilities was proposed, which is structurally compatible with the existing internals of OPR-1000. Actually the advantage of an annular fuel comes from the fuel performance and thermal hydraulics. In the thermal hydraulic analysis, the mixing effect between the neighboring channels has to be carried out in a subchannel analysis. A subchannel analysis code, MATRA has been developed by KAERI. However, MATRA dose not have the capability to model both an internally and externally cooled annular fuel. A subchannel code, MATRA-AF which can be coupled to MATRA and can calculate the coolant flow distribution and heat transfer fraction in the internal and external subchannels has been developed. In this paper, the characteristics and the verification of the MATRA-AF are described. The effects of the thermal hydraulic parameters are estimated through a single fuel assembly

  14. Summary of Booster Development and Qualification Report

    Energy Technology Data Exchange (ETDEWEB)

    Francois, Elizabeth G. [Los Alamos National Laboratory; Harry, Herbert H. [Los Alamos National Laboratory; Hartline, Ernest L. [Los Alamos National Laboratory; Hooks, Daniel E. [Los Alamos National Laboratory; Johnson, Carl E. [Los Alamos National Laboratory; Morris, John S. [Los Alamos National Laboratory; Novak, Alan M. [Los Alamos National Laboratory; Ramos, Kyle J. [Los Alamos National Laboratory; Sanders, Victor E. [Los Alamos National Laboratory; Scovel, Christina A. [Los Alamos National Laboratory; Lorenz, Thomas [LLNL; Wright, Mark [AWE; Botcher, Tod [PANTEX; Marx, Erin [NSWC-IHDIV; Gibson, Kevin [NSWC-IHDIV

    2012-06-21

    This report outlines booster development work done at Los Alamos National Laboratory from 2007 to present. The booster is a critical link in the initiation train of explosive assemblies, from complex devices like nuclear weapons to conventional munitions. The booster bridges the gap from a small, relatively sensitive detonator to an insensitive, but massive, main charge. The movement throughout the explosives development community is to use more and more insensitive explosive components. With that, more energy is needed out of the booster. It has to initiate reliably, promptly, powerfully and safely. This report is divided into four sections. The first provides a summary of a collaborative effort between LANL, LLNL, and AWE to identify candidate materials and uniformly develop a testing plan for new boosters. Important parameters and the tests required to measure them were defined. The nature of the collaboration and the specific goals of the participating partners has changed over time, but the booster development plan stands on its own merit as a complete description of the test protocol necessary to compare and qualify booster materials, and is discussed in its entirety in this report. The second section describes a project, which began in 2009 with the Department of Defense to develop replacement booster formulations for PBXN-7. Replacement of PBXN-7 was necessary because it contained Triaminotrinitrobenzene (TATB), which was becoming unavailable to the DoD and because it contained Cyclotrimethylenetrinitramine (RDX), which was sensitive and toxic. A LANL-developed explosive, Diaminoazoxyfurazan (DAAF), was an important candidate. This project required any replacement formulation be a drop-in replacement in existing munitions. This project was timely, in that it made use of the collaborative booster development project, and had the additional constraint of matching shock sensitivity. Additionally it needed to be a safety improvement, and a performance

  15. Validation of PWR core seismic models with shaking table tests on interacting scale 1 fuel assemblies

    International Nuclear Information System (INIS)

    The fuel assembly mechanical strength must be justified with respect to the lateral loads under accident conditions, in particular seismic loads. This justification is performed by means of time-history analyses with dynamic models of an assembly row in the core, allowing for assembly deformations, impacts at grid locations and reactor coolant effects. Due to necessary simplifications, the models include 'equivalent' parameters adjusted with respect to dynamic characterisation tests of the fuel assemblies. Complementing such tests on isolated assemblies by an overall model validation with shaking table tests on interacting assemblies is obviously desirable. Seismic tests have been performed by French CEA (Commissariat a l'Energie Atomique) on a row of six full scale fuel assemblies, including two types of 17 x 17 12ft design. The row models are built according to the usual procedure, with preliminary characterisation tests performed on a single assembly. The test-calculation comparisons are made for two test configurations : in air and in water. The relatively large number of accelerograms (15, used for each configuration) is also favourable to significant comparisons. The results are presented for the impact forces at row ends, displacements at mid assembly, and also 'statistical' parameters. Despite a non-negligible scattering in the results obtained with different accelerograms, the calculations prove realistic, and the modelling process is validated with a good confidence level. This satisfactory validation allows to evaluate precisely the margins in the seismic design methodology of the fuel assemblies, and thus to confirm the safety of the plants in case of seismic event. (author)

  16. A US perspective on fast reactor fuel fabrication technology and experience part I: metal fuels and assembly design

    International Nuclear Information System (INIS)

    This paper is part I of a review focusing on the United States experience with metallic fast reactor fuel fabrication and assembly design for the Experimental Breeder Reactor-II (EBR-II) and the Fast Flux Test Facility (FFTF). Experience with metal fuel fabrication in the United States is extensive, including over 60 years of research conducted by the government, national laboratories, industry, and academia. This experience has culminated in a considerable amount of research that resulted in significant improvements to the technologies employed to fabricate metallic fast reactor fuel. This part of the review documents the current state of fuel fabrication technologies for metallic fuels, some of the challenges faced by previous researchers, and how these were overcome. Knowledge gained from reviewing previous investigations will aid both researchers and policy makers in forming future decisions relating to nuclear fuel fabrication technologies.

  17. Supercell burnup model for the physics design of BWR fuel assemblies

    International Nuclear Information System (INIS)

    A code called SUPERB has been developed for the BWR fuel assembly burnup analyses using supercell model. Each of the characteristic heterogeneities of a BWR fuel assembly like water gap, poisoned pins, control blade etc., is treated by invoking appropriate supercell concept. The burnup model of SUPERB is so devised as to strike a balance between accuracy and speed. This is achieved by building isotopic densities in each fuel pin separately while the depletion equations are solved only in a few groups of pins or burnup zones and the multigroup neutron spectra are differentiated in fewer group of pincell types. Multiple fuel ring burnup is considered only for Gd isotopes. A special empirical formula allows the microscopic cross section of Gd isotopes to be varied even during burnup integration. The supercell model has been tested against Monte Carlo results for the fresh cold clean Tarapur fuel assembly with two Gd fuel pins. The burnup model of SUPERB has been validated against one of the most sophisticated codes LWR-WIMS for a benchmark problem involving all the complexities of a BWR fuel assembly. The agreement of SUPERB results with both Monte Carlo and LWR-WIMS results is found to be excellent. (auth.)

  18. High-level neutron-coincidence-counter (HLNCC) implementation: assay of the plutonium content of mixed-oxide fuel assemblies

    International Nuclear Information System (INIS)

    The portable High-Level Neutron Coincidence Counter is used to assay the 240Pu-effective loading of a reference mixed-oxide fuel assembly by neutron coincidence counting. We have investigated the effects on the coincidence count rate of the total fuel loading (UO2 + PuO2), the fissile loading, the fuel rod diameter, and the fuel rod pattern. The coincidence count rate per gram of 240Pu-effective per centimeter is primarily dependent on the total fuel loading of the assembly; the higher the loading, the higher the coincidence count rate. Detailed procedures for the assay of mixed-oxide fuel assemblies are developed

  19. Applicability of a set of tomographic reconstruction algorithms for quantitative SPECT on irradiated nuclear fuel assemblies

    International Nuclear Information System (INIS)

    A fuel assembly operated in a nuclear power plant typically contains 100–300 fuel rods, depending on fuel type, which become strongly radioactive during irradiation in the reactor core. For operational and security reasons, it is of interest to experimentally deduce rod-wise information from the fuel, preferably by means of non-destructive measurements. The tomographic SPECT technique offers such possibilities through its two-step application; (1) recording the gamma-ray flux distribution around the fuel assembly, and (2) reconstructing the assembly's internal source distribution, based on the recorded radiation field. In this paper, algorithms for performing the latter step and extracting quantitative relative rod-by-rod data are accounted for. As compared to application of SPECT in nuclear medicine, nuclear fuel assemblies present a much more heterogeneous distribution of internal attenuation to gamma radiation than the human body, typically with rods containing pellets of heavy uranium dioxide surrounded by cladding of a zirconium alloy placed in water or air. This inhomogeneity severely complicates the tomographic quantification of the rod-wise relative source content, and the deduction of conclusive data requires detailed modelling of the attenuation to be introduced in the reconstructions. However, as shown in this paper, simplified models may still produce valuable information about the fuel. Here, a set of reconstruction algorithms for SPECT on nuclear fuel assemblies are described and discussed in terms of their quantitative performance for two applications; verification of fuel assemblies' completeness in nuclear safeguards, and rod-wise fuel characterization. It is argued that a request not to base the former assessment on any a priori information brings constraints to which reconstruction methods that may be used in that case, whereas the use of a priori information on geometry and material content enables highly accurate quantitative

  20. Applicability of a set of tomographic reconstruction algorithms for quantitative SPECT on irradiated nuclear fuel assemblies

    Science.gov (United States)

    Jacobsson Svärd, Staffan; Holcombe, Scott; Grape, Sophie

    2015-05-01

    A fuel assembly operated in a nuclear power plant typically contains 100-300 fuel rods, depending on fuel type, which become strongly radioactive during irradiation in the reactor core. For operational and security reasons, it is of interest to experimentally deduce rod-wise information from the fuel, preferably by means of non-destructive measurements. The tomographic SPECT technique offers such possibilities through its two-step application; (1) recording the gamma-ray flux distribution around the fuel assembly, and (2) reconstructing the assembly's internal source distribution, based on the recorded radiation field. In this paper, algorithms for performing the latter step and extracting quantitative relative rod-by-rod data are accounted for. As compared to application of SPECT in nuclear medicine, nuclear fuel assemblies present a much more heterogeneous distribution of internal attenuation to gamma radiation than the human body, typically with rods containing pellets of heavy uranium dioxide surrounded by cladding of a zirconium alloy placed in water or air. This inhomogeneity severely complicates the tomographic quantification of the rod-wise relative source content, and the deduction of conclusive data requires detailed modelling of the attenuation to be introduced in the reconstructions. However, as shown in this paper, simplified models may still produce valuable information about the fuel. Here, a set of reconstruction algorithms for SPECT on nuclear fuel assemblies are described and discussed in terms of their quantitative performance for two applications; verification of fuel assemblies' completeness in nuclear safeguards, and rod-wise fuel characterization. It is argued that a request not to base the former assessment on any a priori information brings constraints to which reconstruction methods that may be used in that case, whereas the use of a priori information on geometry and material content enables highly accurate quantitative assessment, which

  1. Applicability of a set of tomographic reconstruction algorithms for quantitative SPECT on irradiated nuclear fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Jacobsson Svärd, Staffan, E-mail: staffan.jacobsson_svard@physics.uu.se; Holcombe, Scott; Grape, Sophie

    2015-05-21

    A fuel assembly operated in a nuclear power plant typically contains 100–300 fuel rods, depending on fuel type, which become strongly radioactive during irradiation in the reactor core. For operational and security reasons, it is of interest to experimentally deduce rod-wise information from the fuel, preferably by means of non-destructive measurements. The tomographic SPECT technique offers such possibilities through its two-step application; (1) recording the gamma-ray flux distribution around the fuel assembly, and (2) reconstructing the assembly's internal source distribution, based on the recorded radiation field. In this paper, algorithms for performing the latter step and extracting quantitative relative rod-by-rod data are accounted for. As compared to application of SPECT in nuclear medicine, nuclear fuel assemblies present a much more heterogeneous distribution of internal attenuation to gamma radiation than the human body, typically with rods containing pellets of heavy uranium dioxide surrounded by cladding of a zirconium alloy placed in water or air. This inhomogeneity severely complicates the tomographic quantification of the rod-wise relative source content, and the deduction of conclusive data requires detailed modelling of the attenuation to be introduced in the reconstructions. However, as shown in this paper, simplified models may still produce valuable information about the fuel. Here, a set of reconstruction algorithms for SPECT on nuclear fuel assemblies are described and discussed in terms of their quantitative performance for two applications; verification of fuel assemblies' completeness in nuclear safeguards, and rod-wise fuel characterization. It is argued that a request not to base the former assessment on any a priori information brings constraints to which reconstruction methods that may be used in that case, whereas the use of a priori information on geometry and material content enables highly accurate quantitative

  2. Design of a mixed-oxide fuel assembly to be assessed as a lead test assembly in a BWR reactor

    International Nuclear Information System (INIS)

    The open and the close cycle are the two alternatives to pursue during power generation. The reprocessing is a mature process that now shows a more competitive economic aspect, making it more attractive than ever. Mexico has not decided what to do with the existing and future depleted fuel assemblies that will be generated from the power operation, thus the direct disposal and the reprocessing are still being considered. To have enough arguments in one or the other alternatives it is necessary to make an assessment of both. This investigation focus in the MOX fuel design assuming that the reprocessing is the option to follow and looking for the lowest impact in power generation. The first step in a reprocessing program is to analyze the performance of four lead test assemblies (LTA's), thus in this investigation we design the corresponding MOX to be used as LTA's and assess their performance through one operational cycle. (author)

  3. Experimental investigations for determination of heat-transfer coefficients and temperature fields in simulated fuel assemblies of BREST reactor with fuel elements spaced by transverse grids

    International Nuclear Information System (INIS)

    The consideration is given to heat transfer and temperature fields in fuel pin bundles with transverse spacer grids (s/d =1.33) equally spaced along energy deposition length. Experimental data are obtained on two simulated 37-rod core assemblies: one assembly is with uniform geometry along the cross-section and in the other there is nonheated rod simulating supporting pipe in fuel assembly of reactor with heavy coolant. Eutectic Na-K alloy is used as coolant. Nusselt numbers and temperature nonuniformity along the perimeter of measurement fuel element simulator obtained in these assemblies are compared as well as available data for finned (wire to wire) fuel rods

  4. The conceptual design for fuel assembly and reactor structures of a new research reactor

    International Nuclear Information System (INIS)

    A new Research Reactor(ARR) is being designed by KAERI since 2002. The final goal of the project is to develop a new and unique research reactor model which is superior in safety and economical aspect. In this work, as a first step for the design of the fuel assembly and reactor structures of the ARR, the conceptual design has been carried out. The vibration characteristics of the tubular fuel model and the locking performance of the preliminary designed locking devices were investigated. The basic design of the reactor core, reactor structure, and CRDM of the ARR has been carried out. In order to investigate the effects of the stiffener on the vibration characteristics of the tubular fuel, modal analysis was performed for the finite element models of the tubular fuel with stiffener and without stiffener. The analysis results show that the vibration characteristics of the tubular fuel with stiffener is better than those of the tubular fuel without stiffener. To investigate the locking performance of the preliminary designed locking devices for the fuel assembly of the ARR, the elements of the locking devices were fabricated. Then the rotating resistance, fixing status and vibration characteristics of the locking devices were tested. The test results show that using the locking device with fins on the bottom guide can prevent the rotating motion of the fuel assembly, and additional spring or guide on the top of the fuel assembly is needed to suppress the lateral motion of the fuel assembly. Based on the modal analysis and experimental results, the fuel assembly and locking device of the ARR were newly designed. The locking performance of the newly designed locking device, pressure drop characteristics and vibration characteristics of the newly designed fuel assembly will be tested in the near future. In addition, based on the technical experiences on the design and operation of the HANARO, the reactor core, reactor structures, and CRDM of the ARR were conceptually

  5. Usual control of RA-3 fuel assemblies for the upgrading the reactor power and fuel density

    International Nuclear Information System (INIS)

    In recent years, the Argentine Atomic Energy Commission has been working actively on the upgrading of its RA-3 research reactor and the development of new fuel compounds. The last and most important milestones were the irradiation of fuel prototypes with higher U mass and higher density that standard, and the upgrading of the reactor power from 5 to 10 MW. The RA-3 is a typical Material Testing and Research Reactor (MTR), built and operated by CNEA, and located at Ezeiza Atomic Center in Buenos Aires, Argentina. It is a pool type reactor, moderated and cooled with light water. Cooling is provided by down going forced convection. Since its commissioning, the reactor had mainly operated at a thermal power of 5 MW until a power increase program was initiated. Recently, the target of 10 MW was reached successfully. This involved mainly the improvement of primary and secondary cooling and water purification systems. The primary coolant flow was increased from 750 to 1350 m3/h, this was achieved through a third pipeline with its corresponding pump and heat exchanger. The current core consists of 21 standard fuel elements (Fe) and 4 control fuel elements. These fuel assemblies are Mtr-type box plates of low enrichment Uranium (Leu), with an U3O8-Al fuel meat matrix with a meat density of 3.2 gu/cm3 and Aluminum-6061 cladding, having 19 and 14 fuel plates respectively. Since 1997, three full scale prototypes of U3Si2-Al matrix with a meat density of 4.8 gU/cm3 and having up to 30% higher U mass have been irradiated. The irradiation of FE with even higher U mass and increased density compounds is about to start. All of these modifications brought about the need of a more demanding control of the core components, the fuel elements, and the radiological and chemical quality of water of the primary cooling system (PWS). Before the power increase program was initiated, all the graphite reflector boxes were inspected in the annex built-in hot cell of the reactor in order to

  6. Simulation of Spent PWR Fuel Assembly Behavior Under Normal Conditions of Transport

    International Nuclear Information System (INIS)

    The behavior of a PWR high-burnup spent fuel assembly under normal conditions of transport is simulated in a dynamic analysis of a 0.3-m free drop of a transportation cask unprotected by impact limiters striking a flat rigid surface in the horizontal orientation. The structural analysis employs a finite element numerical model consisting of the cask, the fuel assemblies, the fuel rods, the guide tubes and the cask’s internal structures that hold the fuel assemblies in position. Appropriate mechanical properties for the cask’s structural components, as well as the elastic-plastic properties typical of high-burnup Zircaloy-4 cladding, are utilized. Emphasis is placed on fuel rods responses at locations where maximum forces would be expected, which include end-plate positions and spacer-grid positions at assembly mid-span. Temporal and spatial variations of the forces acting on the fuel rods are calculated and post-processed to obtain frequency distributions, which statistically represent the total fuel rod population in the cask. The results show that the largest pinch force, (ror-to-rod contact force), is 1700 lb, the maximum axial force is 600 lb, and the largest bending moment is 175 in-lb. Failure analysis of fuel rods using these force quantities, and considering the effects of potential hydrides reorientation on cladding failure resistance, indicates, under conservative assumptions, a factor of safety of least 2 against longitudinal tearing, and no failure is predicted for transverse tearing or rod breakage. Fuel reconfiguration is predicted not to occur, and although partial tearing of guide tubes is possible, it is not enough to impair post-accident assembly retrieval. (author)

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

    International Nuclear Information System (INIS)

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

  8. Spent fuel dry storage technology development: fuel temperature measurements under imposed dry storage conditions (I kW PWR spent fuel assembly)

    International Nuclear Information System (INIS)

    A spent fuel assembly temperature test under imposed dry storage conditions was conducted at the Engine Maintenance Assembly and Disassembly (E-MAD) facility on the Nevada Test Site in support of spent fuel dry storage technology development. This document presents the test data and results obtained from an approximately 1.0 kW decay heat level PWR spent fuel assembly. A spent fuel test apparatus was designed to utilize a representative stainless steel spent fuel canister, a canister lid containing internal temperature instrumentation to measure fuel cladding temperatures, and a carbon steel liner that encloses the canister and lid. Electrical heaters along the liner length, on the lid, and below the canister are used to impose dry storage canister temperature profiles. Temperature instrumentation is provided on the liner and canister. The liner and canister are supported by a test stand in one of the large hot cells (West Process Cell) inside E-MAD. Fuel temperature measurements have been performed using imposed canister temperature profiles from the electrically heated and spent fuel drywell tests being conducted at E-MAD as well as for four constant canister temperature profiles, each with a vacuum, helium and air backfill. Computer models have been utilized in conjunction with the test to predict the thermal response of the fuel cladding. Computer predictions are presented, and they show good agreement with the test data

  9. Neutronic study on seed-blanket type reduced-moderation water reactor fuel assembly

    International Nuclear Information System (INIS)

    Parametric studies have been done for a PWR-type reduced-moderation water reactor (RMWR) with seed-blanket fuel assemblies to achieve a high conversion ratio, a negative void reactivity coefficient and a high burnup by using MOX, metal (Pu+U+Zr) or T-MOX (PuO2+ThO2) fuels. From the result of the assembly burnup calculation, it has been seen that 50% to 60% of seed in a seed-blanket (MOX-UO2) assembly has higher conversion ratio compared to the other combinations of seeds and blankets. And the recommended number of seed-blanket layers is 20, in which the number of seed layers is 15 (S15) and that of blanket layers is 5 (B5). It was found that the conversion ratio of a seed-blanket assembly decreases, when seed and blanket are arranged so as to look like a flower shape (Hanagara). By the optimization of different parameters, the S15B5 fuel assembly with the height of seed of 1,000x2 mm, internal blanket of 150 mm and axial blanket of 400x2 mm is recommended for a high conversion ratio. In this assembly, the gap of seed fuel rod is 1.0 mm and that of blanket fuel rod is 0.4 mm. In the S15B5 assembly, the conversion ratio is 1.0 and the average burnup in (seed + internal blanket + outer blanket) region is 38 GWd/t. The cycle length of the core is 16.5 effective full power in month (EFPM) by 6 batches refuelling scheme and the enrichment of fissile Pu is 14.6 wt%. The void coefficient is +22 pcm/%void, though, it is expected that the void coefficient will be negative if the radial neutron leakage is taken into account in the calculation. It is also possible to use the S15B5 fuel assembly as a high burnup reactor to achieve 45 GWd/t in (seed + internal blanket + outer blanket) region, but, it is necessary to decrease the height of seed to 500x2 mm to improve the void coefficient. In this reactor, the conversion ratio is 0.97 and void coefficient is +21 pcm/%void. The fuel temperature coefficient is negative for both of the cases. It is possible to improve the conversion

  10. Transport of fresh MOX fuel assemblies for the Monju initial core

    International Nuclear Information System (INIS)

    Transport of fresh MOX fuel assemblies for the prototype FBR MONJU initial core started in July 1992 and ended in March 1994. As many as 205 fresh MOX fuel assemblies for an inner core, 91 assemblies for an outer core and 5 assemblies for testing) were transported in nine transport missions. The packaging for fuel assemblies, which has shielding and shock absorbing material inside, meets IAEA regulatory requirements for Type B(U) packaging including hypothetical accident conditions such as the 9 m drop test, fire test, etc. Moreover, this package design feature such advanced technologies as high performance neutron shielding material and an automatic hold-down mechanism for the fuel assemblies. Every effort was made to carry out safe transport in conjunction with the cooperation of every competent organisation. This effort includes establishment of the transport control centre, communication training, and accompanying of the radiation monitoring expert. No transport accident occurred during the transport and all the transport missions were successfully completed on schedule. (Author)

  11. Assaying Used Nuclear Fuel Assemblies Using Lead Slowing-Down Spectroscopy and Singular Value Decomposition

    International Nuclear Information System (INIS)

    This study investigates the use of a Lead Slowing-Down Spectrometer (LSDS) for the direct and independent measurement of fissile isotopes in light-water nuclear reactor fuel assemblies. The current study applies MCNPX, a Monte Carlo radiation transport code, to simulate the measurement of the assay of the used nuclear fuel assemblies in the LSDS. An empirical model has been developed based on the calibration of the LSDS to responses generated from the simulated assay of six well-characterized fuel assemblies. The effects of self-shielding are taken into account by using empirical basis vectors calculated from the singular value decomposition (SVD) of a matrix containing the self-shielding functions from the assay of assemblies in the calibration set. The performance of the empirical algorithm was tested on version 1 of the Next-Generation Safeguards Initiative (NGSI) used fuel library consisting of 64 assemblies, as well as a set of 27 diversion assemblies, both of which were developed by Los Alamos National Laboratory. The potential for direct and independent assay of the sum of the masses of Pu-239 and Pu-241 to within 2%, on average, has been demonstrated

  12. Investigation of a Shock Absorber for Safeguard of Fuel Assemblies Failure

    International Nuclear Information System (INIS)

    The Ignalina NPP has two reactors. The Unit 1 was shut down, therefore the special equipment was designed for transportation of the fuel from Unit 1 to Unit 2. The fuel-loaded basket can drop during transportation. The special shock absorber was designed in order to avoid failure of fuel assemblies during transportation. In case of drop of fuel loaded basket, the failure of fuel assemblies can occur. This shock absorber was studied by scaled experiments at Lithuanian Energy Institute. Static and dynamic investigations of shock absorber are presented in this paper, including dependency of axial force versus axial compression. The finite element codes BRIGADE/Plus and ABAQUS/Explicit were used for analysis. Static simulation was used to optimize the dimensions of shock absorber. Dynamic analysis shows that shock absorber is capable to withstand the dynamic load for successful force suppression function in case of an accident. (authors)

  13. Nondestructive measurements with a WWER-440 fuel assembly model using neutron and gamma sources

    International Nuclear Information System (INIS)

    Laboratory measurements were carried out in order to give a better insight to the behaviour of neutron and gamma radiation in spent nuclear fuel. A WWER-440 fuel assembly model containing fresh 3.6 % enriched fuel was used together with a line-shaped gamma source or a neutron point source. The contribution of single fuel rods to the measured gamma signal was measured at different energies in the geometry used in real spent fuel measurements. The origin of the measured neutrons was detected by employing the neutron source. The so-called end effects of the assembly were examined, as well. Both a horizontal, using a fork detector, and a vertical detector geometry was studied. In fork detector measurements the neutrons were detected by means of cadmium-wrapped and bare fission chambers with boron concentrations of 2100 ppm and 4000 ppm in addition to pure water

  14. Dynamic analytical and experimental research of shock absorber to safeguard the nuclear fuel assemblies

    International Nuclear Information System (INIS)

    Highlights: • Plastical deformation of the shock absorber. • Dynamic testing of the scaled shock absorber. • Dynamic simulation of the shock absorber using finite element method. • Strain-rate evaluation in dynamic analysis. • Variation of displacement, acceleration and velocity during dynamic impact. -- Abstract: The Ignalina Nuclear Power Plant (NPP) has two RBMK-1500 graphite moderated boiling water multi-channel reactors. The Ignalina NPP Unit 1 was shut down at the end of 2004 while Unit 2 has been in operation for over 5 years. After shutdown at the Unit 1 remained spent fuel assemblies with low burn-up depth. In order to reuse these assemblies in the reactor of Unit 2 a special set of equipment was developed. One of the most important items of this set is a container, which is used for the transportation of spent fuel assemblies between the reactors of Unit 1 and Unit 2. A special shock absorber was designed to avoid failure of fuel assemblies in case of hypothetical spent fuel assemblies drop accident during uploading/unloading of spent fuel assemblies to/from container. This shock absorber was examined using scaled experiments and finite element analysis. Static and dynamic investigations of the shock absorber were performed for the estimation and optimization of its geometrical parameters. The objective of this work is the estimation whether the proposed design of shock absorber can fulfil the stopping function of the spent fuel assemblies and is capable to withstand the dynamics load. Experimental testing of scaled shock absorber models and dynamic analytical investigations using the finite element code ABAQUS/Explicit were performed. The simulation model was verified by comparing the experimental and simulation results and it was concluded that the shock absorber is capable to withstand the dynamic load, i.e. successful force suppression function in case of accident

  15. Time-dependent radiation release characteristics of irradiated nuclear fuel assemblies in Korean nuclear power plants

    International Nuclear Information System (INIS)

    The time-dependent radiation release characteristics of nuclear fuel assemblies need to be understood to manage operation safely and protect workers from the radiation released over the operation period of a nuclear reactor, including its overhaul period. This study examines the time-dependent radiation release characteristics of 2 types of PWR nuclear fuel assemblies in Korea according to their burn-up. Two types of nuclear fuel assemblies considered were: a 16 x 16 fuel assembly of Ulchin units 5 and 6 (KSNP type) and a 17 x 17 fuel assembly of Kori units 3 and 4 (Westinghouse type). The 5 different burn-ups (MWD/MTU) were considered: 100, 10,000, 20,000, 30,000, and 40,000. The calculations showed that the increase rate of neutron dose distribution was relatively higher than that of the gamma dose distribution according to burn-up, while the gamma dose distributions were much higher than the neutron dose distributions. (orig.)

  16. LLNL MOX fuel lead assemblies data report for the surplus plutonium disposition environmental impact statement

    International Nuclear Information System (INIS)

    The purpose of this document is to support the US Department of Energy (DOE) Fissile Materials Disposition Program's preparation of the draft surplus plutonium disposition environmental impact statement. This is one of several responses to data call requests for background information on activities associated with the operation of the lead assembly (LA) mixed-oxide (MOX) fuel fabrication facility. The DOE Office of Fissile Materials Disposition (DOE-MD) has developed a dual-path strategy for disposition of surplus weapons-grade plutonium. One of the paths is to disposition surplus plutonium through irradiation of MOX fuel in commercial nuclear reactors. MOX fuel consists of plutonium and uranium oxides (PuO2 and UO2), typically containing 95% or more UO2. DOE-MD requested that the DOE Site Operations Offices nominate DOE sites that meet established minimum requirements that could produce MOX LAs. LLNL has proposed an LA MOX fuel fabrication approach that would be done entirely inside an S and S Category 1 area. This includes receipt and storage of PuO2 powder, fabrication of MOX fuel pellets, assembly of fuel rods and bundles, and shipping of the packaged fuel to a commercial reactor site. Support activities will take place within a Category 1 area. Building 332 will be used to receive and store the bulk PuO2 powder, fabricate MOX fuel pellets, and assemble fuel rods. Building 334 will be used to assemble, store, and ship fuel bundles. Only minor modifications would be required of Building 332. Uncontaminated glove boxes would need to be removed, petition walls would need to be removed, and minor modifications to the ventilation system would be required

  17. LLNL MOX fuel lead assemblies data report for the surplus plutonium disposition environmental impact statement

    Energy Technology Data Exchange (ETDEWEB)

    O`Connor, D.G.; Fisher, S.E.; Holdaway, R. [and others

    1998-08-01

    The purpose of this document is to support the US Department of Energy (DOE) Fissile Materials Disposition Program`s preparation of the draft surplus plutonium disposition environmental impact statement. This is one of several responses to data call requests for background information on activities associated with the operation of the lead assembly (LA) mixed-oxide (MOX) fuel fabrication facility. The DOE Office of Fissile Materials Disposition (DOE-MD) has developed a dual-path strategy for disposition of surplus weapons-grade plutonium. One of the paths is to disposition surplus plutonium through irradiation of MOX fuel in commercial nuclear reactors. MOX fuel consists of plutonium and uranium oxides (PuO{sub 2} and UO{sub 2}), typically containing 95% or more UO{sub 2}. DOE-MD requested that the DOE Site Operations Offices nominate DOE sites that meet established minimum requirements that could produce MOX LAs. LLNL has proposed an LA MOX fuel fabrication approach that would be done entirely inside an S and S Category 1 area. This includes receipt and storage of PuO{sub 2} powder, fabrication of MOX fuel pellets, assembly of fuel rods and bundles, and shipping of the packaged fuel to a commercial reactor site. Support activities will take place within a Category 1 area. Building 332 will be used to receive and store the bulk PuO{sub 2} powder, fabricate MOX fuel pellets, and assemble fuel rods. Building 334 will be used to assemble, store, and ship fuel bundles. Only minor modifications would be required of Building 332. Uncontaminated glove boxes would need to be removed, petition walls would need to be removed, and minor modifications to the ventilation system would be required.

  18. AREVA's fuel assemblies addressing high performance requirements of the worldwide PWR fleet

    International Nuclear Information System (INIS)

    Taking advantage of its presence in the fuel activities since the start of commercial nuclear worldwide operation, AREVA is continuing to support the customers with the priority on reliability, to: >participate in plant operational performance for the in core fuel reliability, the Zero Tolerance for Failure ZTF as a continuous improvement target and the minimisation of manufacturing/quality troubles, >guarantee the supply chain a proven product stability and continuous availability, >support performance improvements with proven design and technology for fuel management updating and cycle cost optimization, >support licensing assessments for fuel assembly and reloads, data/methodologies/services, >meet regulatory challenges regarding new phenomena, addressing emergent performance issues and emerging industry challenges for changing operating regimes. This capacity is based on supplies by AREVA accumulating very large experience both in manufacturing and in plant operation, which is demonstrated by: >manufacturing location in 4 countries including 9 fuel factories in USA, Germany, Belgium and France. Up to now about 120,000 fuel assemblies and 8,000 RCCA have been released to PWR nuclear countries, from AREVA European factories, >irradiation performed or in progress in about half of PWR world wide nuclear plants. Our optimum performances cover rod burn ups of to 82GWD/tU and fuel assemblies successfully operated under various world wide fuel management types. AREVA's experience, which is the largest in the world, has the extensive support of the well known fuel components such as the M5'TM'cladding, the MONOBLOC'TM'guide tube, the HTP'TM' and HMP'TM' structure components and the comprehensive services brought in engineering, irradiation and post irradiation fields. All of AREVA's fuel knowledge is devoted to extend the definition of fuel reliability to cover the whole scope of fuel vendor support. Our Top Reliability and Quality provide customers with continuous

  19. Management of research reactor; dynamic characteristics analysis for reactor structures related with vibration of HANARO fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, Chang Kee; Shim, Joo Sup [Shinwa Technology Information, Seoul (Korea)

    2001-04-01

    The objective of this study is to deduce the dynamic correlation between the fuel assembly and the reactor structure. Dynamic characteristics analyses for reactor structure related with vibration of HANARO fuel assembly have been performed For the dynamic characteristic analysis, the in-air models of the round and hexagonal flow tubes, 18-element and 36-element fuel assemblies, and reactor structure were developed. By calculating the hydrodynamic mass and distributing it on the in-air models, the in-water models of the flow tubes, the fuel assemblies, and the reactor structure were developed. Then, modal analyses for developed in-air and in-water models have been performed. Especially, two 18-element fuel assemblies and three 36-element fuel assemblies were included in the in-water reactor models. For the verification of the modal analysis results, the natural frequencies and the mode shapes of the fuel assembly were compared with those obtained from the experiment. Finally the analysis results of the reactor structure were compared with them performed by AECL Based on the reactor model without PCS piping, the in-water reactor model including the fuel assemblies was developed, and its modal analysis was performed. The analysis results demonstrate that there are no resonance between the fuel assembly and the reactor structures. 26 refs., 419 figs., 85 tabs. (Author)

  20. The optimization of an AP1000 fuel assembly for the transmutation of plutonium and minor actinides

    Science.gov (United States)

    Washington, Jeremy A.

    The average nuclear power plant produces twenty metric tons of used nuclear fuel per year, containing approximately 95 wt% uranium, 1 wt% plutonium, and 4 wt% fission products and transuranic elements. Fast reactors are a preferred option for the transmutation of plutonium and minor actinides; however, an optimistic deployment time of at least 20 years indicates a need for a near-term solution. The goal of this thesis is to examine the potential of light water reactors for plutonium and minor actinides transmutation as a near-term solution. This thesis screens the available nuclear isotope database to identify potential absorbers as coatings on a transmutation fuel in a light water reactor. A spectral shift absorber coating tunes the neutron energy spectrum experienced by the underlying target fuel. Eleven different spectral shift absorbers (B4C, CdO, Dy2O3, Er 2O3, Eu2O3, Gd2O3, HfO2, In2O3, Lu2O3, Sm2O3, and TaC) have been selected for further evaluation. A model developed using the NEWT module of SCALE 6.1 code provided performance data for the burnup of the target fuel rods. Irradiation of the target fuels occurs in a Westinghouse 17x17 XL Robust Fuel Assembly over a 1400 Effective Full Power Days (EFPD) interval. The fuels evaluated in this thesis include PuO2, Pu3Si2, PuN, MOX, PuZrH, PuZrHTh, PuZrO 2, and PuUZrH. MOX (5 wt% PuO2), Pu0.31ZrH 1.6Th1.08, and PuZrO2MgO (8 wt%) are selected for detailed analysis in a multi-pin transmutation assembly. A coupled model optimized the resulting transmutation fuel elements. The optimization considered three stages of fuel assemblies containing target fuel pins. The first stage optimized four target fuel pins adjacent to the central instrumentation channel. The second stage evaluated a variety of assemblies with multiple target fuel pins and the third stage re-optimized target fuel pins in the second-stage assembly. A PuZrO2MgO (8 wt%) target fuel with a coating of Lu 2O3 resulted in the greatest reduction in curium-244

  1. Welding machine and welding process for nuclear fuel assembly structures

    International Nuclear Information System (INIS)

    The welding device comprises a mounting jig which receives the guide tubes and the assembly supporting structures in the desired spatial orientation. It also comprises a welding head which can travel on rails along the length of the guide tubes and has at least a welding spring chuck movable in two axes and rotatable relative to the welding machine; the spring chuck can pass between two adjacent tube rows and takes a tubes where a weld is necessary. The welding spring chuck can apply spot-welding pulses. This is used for the assembly of guide tubes and bundles for water-cooled nuclear reactors

  2. COXPRO-II: a computer program for calculating radiation and conduction heat transfer in irradiated fuel assemblies

    International Nuclear Information System (INIS)

    This report describes the computer program COXPRO-II, which was written for performing thermal analyses of irradiated fuel assemblies in a gaseous environment with no forced cooling. The heat transfer modes within the fuel pin bundle are radiation exchange among fuel pin surfaces and conduction by the stagnant gas. The array of parallel cylindrical fuel pins may be enclosed by a metal wrapper or shroud. Heat is dissipated from the outer surface of the fuel pin assembly by radiation and convection. Both equilateral triangle and square fuel pin arrays can be analyzed. Steady-state and unsteady-state conditions are included. Temperatures predicted by the COXPRO-II code have been validated by comparing them with experimental measurements. Temperature predictions compare favorably to temperature measurements in pressurized water reactor (PWR) and liquid-metal fast breeder reactor (LMFBR) simulated, electrically heated fuel assemblies. Also, temperature comparisons are made on an actual irradiated Fast-Flux Test Facility (FFTF) LMFBR fuel assembly

  3. Development of prediction method of void fraction distribution in fuel assemblies for use in safety analysis

    International Nuclear Information System (INIS)

    The establishment of code system for BWR safety analysis is now in progress at Institute of Nuclear Safety (INS), in order to predict the onset of boiling transition (BT) in nuclear fuel assemblies in any thermal-hydraulic condition without relying on the thermal-hydraulic characteristic data provided by licensee. The prediction method for void fraction distribution across cross section of BWR fuel assemblies has been developed based on multi-dimensional two-fluid model. Lift forces working on bubbles and void diffusion that can not be handled with one-dimensional analysis were considered. Comparisons between calculated results and experimental data obtained from thermal-hydraulic tests of PWR and BWR mock-up fuel assemblies showed good agreement. Lift force models have been empirical and further studies were needed, but the calculations showed the possibility of applying these models to multi-dimensional gas-liquid two-phase flow analysis. (author)

  4. HEXBU-3D, a three-dimensional PWR-simulator program for hexagonal fuel assemblies

    International Nuclear Information System (INIS)

    HEXBU-3D is a three-dimensional nodal simulator program for PWR reactors. It is designed for a reactor core that consists of hexagonal fuel assemblies and of big follower-type control assemblies. The program solves two-group diffusion equations in homogenized fuel assembly geometry by a sophisticated nodal method. The treatment of feedback effects from xenon-poisoning, fuel temperature, moderator temperature and density and soluble boron concentration are included in the program. The nodal equations are solved by a fast two-level iteration technique and the eigenvalue can be either the effective multiplication factor or the boron concentration of the moderator. Burnup calculations are performed by tabulated sets of burnup-dependent cross sections evaluated by a cell burnup program. HEXBY-3D has been originally programmed in FORTRAN V for the UNIVAC 1108 computer, but there is also another version which is operable on the CDC CYBER 170 computer. (author)

  5. Use of high energy gamma emission tomography for partial defect verification of spent fuel assemblies

    International Nuclear Information System (INIS)

    The possibility to use passive gamma emission tomography for revealing non-destructively the rod structure of spent BWR fuel assemblies has been studied in cooperation with the Finnish Support Programme to the IAEA Safeguards (task FIN A98) and the Technical University of Budabest in Hungary. The ultimate goal is to develop partial verification methods for verification of spent nuclear fuel. The task included experimental measurements of irradiated BWR assemblies using underwater measurement techniques together with computer analysis of the measured data as well as computer simulation of tomographic measurements. The results obtained show that rod-level partial defect verification of spent LWR fuel assemblies is feasible using computed gamma emission tomography. This report describes the results of this project. (orig.). (7 refs., 29 figs., 2 tabs.)

  6. Design and performance verification of fuel assembly and steam generator simulators for SMART reactor

    International Nuclear Information System (INIS)

    The SMART reactor has been developed at KAERI, for the generation of electric power and also for seawater desalination. In order to verify the performance of the SMART design with respect to flow and pressure distribution, an experimental test facility named SCOP has been developed. For the purpose of preserving the flow distribution characteristics, SCOP is linearly reduced with a scaling ratio of 1/5. A CFD analysis was carried out to draw basic design parameters of the venturi tube and the perforated plates in a fuel assembly simulator. A CALIP, which is a flow and pressure drop calibration test facility, has been constructed to evaluate the pressure drop characteristic of fuel assembly and steam generator simulators. This paper shows the results of the actual performance verification and evaluation of fuel assembly and steam generator simulator, were evaluated using a CALIP. (author)

  7. Vibration characteristic analysis for HANARO fuel assembly and flow tube submerged in the water

    International Nuclear Information System (INIS)

    The vibration characteristics of HANARO fuel assembly and flow tube that submerged in the water have been investigated. For this purpose, the finite element models of the in-water fuel assemblies and flow tubes were developed. Then, modal analysis of the developed finite element models were performed by utilizing the ANSYS program. The analysis results show that the fundamental vibration modes of the in-water 18-element and 36-element fuel assemblies are lateral bending modes, and its natural frequencies are found to be 16.1Hz and 16.5Hz, respectively. For the verification of the developed finite element models, modal analysis results were compared with those obtained from the modal test. These results demonstrate that the natural frequencies of the first mode obtained from finite element analysis agree well with those of the modal test and the estimation of the hydrodynamic mass is appropriate

  8. Methodology for local verification of flow regimes in fuel assemblies charts

    International Nuclear Information System (INIS)

    The best estimate thermal hydraulic codes describe adequately two-phase flows in nuclear energy facilities if there is proper system of closed relations. It could be obtained from the reliable information on structure forms of two-phase flows, its boundaries and reliable regime charts. In the paper the methodology of automatic recognition of the boundaries of the main types of two phase flows for rod fuel assemblies is presented. The methodology is based on definition of thermal hydraulic parameters distribution in experimental fuel assembly. The measurements were carried out using ASD signals of acoustic noise. In the paper data on two-phase flow regimes boundaries recognition especially low boundaries of bubble flow are summarized for experimental fuel assembly. The methodology of flow regimes charts applied to recognition of upper boundaries of boiling crisis regime was verificated. The satisfactory coincidence with experimental results have been shown. (author)

  9. Fuel assembly stress and deflection analysis for loss-of-coolant accident and seismic excitation

    International Nuclear Information System (INIS)

    Babcock and Wilcox has evaluated the capability of the fuel assemblies to withstand the effects of a loss-of-coolant accident (LOCA) blowdown, the operational basis earthquake (OBE) and design basis earthquake (DBE), and the simultaneous occurrence of the DBE and LOCA. This method of analysis is applicable to all of B and W's nuclear steam system contracts that specify the skirt-supported pressure vessel. Loads during the saturated and subcooled phases of blowdown following a loss-of-coolant accident were calculated. The maximum loads on the fuel assemblies were found to be below allowable limits, and the maximum deflections of the fuel assemblies were found to be less than those that could prevent the insertion of control rods or the flow of coolant through the core. (U.S.)

  10. Experience in generalizing the data on the rod fuel assembly burnout by the method of cells

    International Nuclear Information System (INIS)

    The comparative analysis of results of calculations on the rod fuel assembly burnout by the method of cells is conducted. The method of cells is based on determining local parameters of coolant flow in the assembly cross section. The channel cross section is conditionally subdivided into elementary cells communicating with each other. For such system of interacting cells the equations of thermohydraulics are solved. Local values of enthalpy and mass rate obtained are used for fuel burnout calculation. The analysis performed has shown that generalizing the empirical data for a given set of experiments on the basis of local parameters of flow in the cells already nowadays assures in most cases accuracy of burnout conditions prediction which is not less than the traditional dependences based on one-dimensional description of the flow. The conclusion is drawn on the prospects of using the method of cells for calculating the burnout in rod fuel assemblies and necessity of its subsequent development

  11. Integrated Three-Voltage-Booster DC-DC Converter to Achieve High Voltage Gain with Leakage-Energy Recycling for PV or Fuel-Cell Power Systems

    Directory of Open Access Journals (Sweden)

    Chih-Lung Shen

    2015-09-01

    Full Text Available In this paper, an integrated three-voltage-booster DC-DC (direct current to direct current converter is proposed to achieve high voltage gain for renewable-energy generation systems. The proposed converter integrates three voltage-boosters into one power stage, which is composed of an active switch, a coupled-inductor, five diodes, and five capacitors. As compared with conventional high step-up converters, it has a lower component count. In addition, the features of leakage-energy recycling and switching loss reduction can be accomplished for conversion efficiency improvement. While the active switch is turned off, the converter can inherently clamp the voltage across power switch and suppress voltage spikes. Moreover, the reverse-recovery currents of all diodes can be alleviated by leakage inductance. A 200 W prototype operating at 100 kHz switching frequency with 36 V input and 400 V output is implemented to verify the theoretical analysis and to demonstrate the feasibility of the proposed high step-up DC-DC converter.

  12. Neutronic optimization in high conversion Th-233U fuel assembly with simulated annealing

    International Nuclear Information System (INIS)

    This paper reports on fuel design optimization of a PWR operating in a self sustainable Th-233U fuel cycle. Monte Carlo simulated annealing method was used in order to identify the fuel assembly configuration with the most attractive breeding performance. In previous studies, it was shown that breeding may be achieved by employing heterogeneous Seed-Blanket fuel geometry. The arrangement of seed and blanket pins within the assemblies may be determined by varying the designed parameters based on basic reactor physics phenomena which affect breeding. However, the amount of free parameters may still prove to be prohibitively large in order to systematically explore the design space for optimal solution. Therefore, the Monte Carlo annealing algorithm for neutronic optimization is applied in order to identify the most favorable design. The objective of simulated annealing optimization is to find a set of design parameters, which maximizes some given performance function (such as relative period of net breeding) under specified constraints (such as fuel cycle length). The first objective of the study was to demonstrate that the simulated annealing optimization algorithm will lead to the same fuel pins arrangement as was obtained in the previous studies which used only basic physics phenomena as guidance for optimization. In the second part of this work, the simulated annealing method was used to optimize fuel pins arrangement in much larger fuel assembly, where the basic physics intuition does not yield clearly optimal configuration. The simulated annealing method was found to be very efficient in selecting the optimal design in both cases. In the future, this method will be used for optimization of fuel assembly design with larger number of free parameters in order to determine the most favorable trade-off between the breeding performance and core average power density. (authors)

  13. LANL MOX fuel lead assemblies data report for the surplus plutonium disposition environmental impact statement

    International Nuclear Information System (INIS)

    The purpose of this document is to support the US Department of Energy (DOE) Fissile Materials Disposition Program's preparation of the draft surplus plutonium disposition environmental impact statement. This is one of several responses to data call requests for background information on activities associated with the operation of the lead assembly (LA) mixed-oxide (MOX) fuel fabrication facility. LANL has proposed an LA MOX fuel fabrication approach that would be done entirely inside an S and S Category 1 area. This includes receipt and storage of PuO2 powder, fabrication of MOX fuel pellets, assembly of fuel rods and bundles, and shipping of the packaged fuel to a commercial reactor site. Support activities will take place within both Category 1 and 2 areas. Technical Area (TA) 55/Plutonium Facility 4 will be used to store the bulk PuO2 powder, fabricate MOX fuel pellets, assemble rods, and store fuel bundles. Bundles will be assembled at a separate facility, several of which have been identified as suitable for that activity. The Chemistry and Metallurgy Research Building (at TA-3) will be used for analytical chemistry support. Waste operations will be conducted in TA-50 and TA-54. Only very minor modifications will be needed to accommodate the LA program. These modifications consist mostly of minor equipment upgrades. A commercial reactor operator has not been identified for the LA irradiation. Postirradiation examination (PIE) of the irradiated fuel will take place at either Oak Ridge National Laboratory or ANL-W. The only modifications required at either PIE site would be to accommodate full-length irradiated fuel rods. Results from this program are critical to the overall plutonium distribution schedule

  14. LANL MOX fuel lead assemblies data report for the surplus plutonium disposition environmental impact statement

    Energy Technology Data Exchange (ETDEWEB)

    Fisher, S.E.; Holdaway, R.; Ludwig, S.B. [and others

    1998-08-01

    The purpose of this document is to support the US Department of Energy (DOE) Fissile Materials Disposition Program`s preparation of the draft surplus plutonium disposition environmental impact statement. This is one of several responses to data call requests for background information on activities associated with the operation of the lead assembly (LA) mixed-oxide (MOX) fuel fabrication facility. LANL has proposed an LA MOX fuel fabrication approach that would be done entirely inside an S and S Category 1 area. This includes receipt and storage of PuO{sub 2} powder, fabrication of MOX fuel pellets, assembly of fuel rods and bundles, and shipping of the packaged fuel to a commercial reactor site. Support activities will take place within both Category 1 and 2 areas. Technical Area (TA) 55/Plutonium Facility 4 will be used to store the bulk PuO{sub 2} powder, fabricate MOX fuel pellets, assemble rods, and store fuel bundles. Bundles will be assembled at a separate facility, several of which have been identified as suitable for that activity. The Chemistry and Metallurgy Research Building (at TA-3) will be used for analytical chemistry support. Waste operations will be conducted in TA-50 and TA-54. Only very minor modifications will be needed to accommodate the LA program. These modifications consist mostly of minor equipment upgrades. A commercial reactor operator has not been identified for the LA irradiation. Postirradiation examination (PIE) of the irradiated fuel will take place at either Oak Ridge National Laboratory or ANL-W. The only modifications required at either PIE site would be to accommodate full-length irradiated fuel rods. Results from this program are critical to the overall plutonium distribution schedule.

  15. Analysis of Advanced Fuel Assemblies and Core Designs for the Current and Next Generations of LWRs

    International Nuclear Information System (INIS)

    The objective of the project is to design and analyze advanced fuel assemblies for use in current and future light water reactors and to assess their ability to reduce the inventory of transuranic elements, while preserving operational safety. The reprocessing of spent nuclear fuel can delay or avoid the need for a second geological repository in the US. Current light water reactor fuel assembly designs under investigation could reduce the plutonium inventory of reprocessed fuel. Nevertheless, these designs are not effective in stabilizing or reducing the inventory of minor actinides. In the course of this project, we developed and analyzed advanced fuel assembly designs with improved thermal transmutation capability regarding transuranic elements and especially minor actinides. These designs will be intended for use in thermal spectrum (e.g., current and future fleet of light water reactors in the US). We investigated various fuel types, namely high burn-up advanced mixed oxides and inert matrix fuels, in various geometrical designs that are compliant with the core internals of current and future light water reactors. Neutronic/thermal hydraulic effects were included. Transmutation efficiency and safety parameters were used to rank and down-select the various designs.

  16. Study of the neutronic behavior of a fuel assembly with gadolinium of a reactor HPLWR

    International Nuclear Information System (INIS)

    This work presents a neutronic study of a square assembly design of double line of fuel rods, with moderator box to center of the arrangement, for the nuclear reactor cooled with supercritical water, High Performance Light Water Reactor (HPLWR). For the fuel analyses of the reactor HPLWR the neutronic code Helios-2 was used, settling down as the first study on fuel under conditions of supercritical water that has been simulated with this code. The analyzed variables, essentials in the neutronic design of any reactor, were the infinite neutrons multiplication factor (k∞) and the maximum power peaking factor (PPFmax), as well as the reactivity coefficients by the fuel temperature. The k∞ and PPFmax values were obtained under conditions in cold (293.6 K) and in hot (to 880.8 K). The tests were realized for a reference fuel assembly design, with 40 fuel rods with enrichments of 4 and 5% of U-235, and considering different concentrations of consumable poison (gadolinium - Gd2O3) in some rods of the same assembly. The obtained results show values k∞ and PPFmax minors to the present in the conventional light water reactors. Moreover, the reactivity coefficients by fuel temperature were verified with the purpose of satisfying the safety conditions required in the nuclear reactors. (Author)

  17. Analysis of Advanced Fuel Assemblies and Core Designs for the Current and Next Generations of LWRs

    Energy Technology Data Exchange (ETDEWEB)

    Ragusa, Jean; Vierow, Karen

    2011-09-01

    The objective of the project is to design and analyze advanced fuel assemblies for use in current and future light water reactors and to assess their ability to reduce the inventory of transuranic elements, while preserving operational safety. The reprocessing of spent nuclear fuel can delay or avoid the need for a second geological repository in the US. Current light water reactor fuel assembly designs under investigation could reduce the plutonium inventory of reprocessed fuel. Nevertheless, these designs are not effective in stabilizing or reducing the inventory of minor actinides. In the course of this project, we developed and analyzed advanced fuel assembly designs with improved thermal transmutation capability regarding transuranic elements and especially minor actinides. These designs will be intended for use in thermal spectrum (e.g., current and future fleet of light water reactors in the US). We investigated various fuel types, namely high burn-up advanced mixed oxides and inert matrix fuels, in various geometrical designs that are compliant with the core internals of current and future light water reactors. Neutronic/thermal hydraulic effects were included. Transmutation efficiency and safety parameters were used to rank and down-select the various designs.

  18. Development of mechanical test techniques for structural components of irradiated PWR fuel assembly

    International Nuclear Information System (INIS)

    An increase of fuel burnup and duration of fuel life remains one of the main methods for a nuclear power engineering enhancement. Properties of structural materials providing corrosion resistance, mechanical strength, and dimensional instability of the components of a fuel assembly (FA) are of great importance for fuel operational reliability in such fuel life cycles. Generally, PWR fuel assemblies consist of a top nozzle, spacer grid, bottom nozzle, and guide/instrumentation tubes. The top and bottom nozzle are fixed to the guide tubes using a screw or bulge method. The spacer grid fixed to the guide/instrumentation tubes using a spot weld or bulge method. To understand the in-reactor performance of PWR FA, several devices and test techniques have been developed for mechanical property tests. Among the structural components of PWR FA, a spacer grid, a hold down spring of a top nozzle and a connecting part of FA were considered. Experimental works were carried out for the unirradiated and irradiated components of advanced nuclear fuel assemblies for KSNPs and Westinghouse type PWRs at IMEF (Irradiated Materials Examination Facility) at KAERI. The developed techniques were verified through a hot cell tests. (author)

  19. Development of mechanical test techniques on the irradiated grid elements in PWR fuel assembly

    International Nuclear Information System (INIS)

    The mechanical properties of the parts of a nuclear fuel assembly are degraded during the operation of the reactor, through the mechanism of irradiation damage. The properties changes of the parts of the fuel assembly should be quantitatively estimated to ensure the safety of the fuel assembly and rod during the operation. The test techniques developed in this paper are used to produce the irradiation data of the grid 1x1 cell springs, the grid 1x1 cell, the spring on one face of the 1x1 cell, the inner/outer strip of the grid and the welded part. The specimens were irradiated in the CT test hole of HANARO of a 30 MW thermal output at 300degC. From the spring tests of mid grid 1x1 cell and grid plate, the irradiation effects can be examined. The irradiation effects on the irradiation growth also were occurred. The buckling load of mid grid 1x1 cell does not change with a neutron irradiation. From the tensile tests, the strengths increased but the elongations decreased due to an irradiation. The tensile test and microstructure examination of the spot and fillet welded parts are performed for the evaluation of an irradiation effects. Through these tests of components, the essential data on the fuel assembly design could be obtained. These results will be used to update the irradiation behavior databases, to improve the performance of fuel assembly, and to predict the service life of the fuel assembly in a reactor. (author)

  20. Process for recycling components of a PEM fuel cell membrane electrode assembly

    Science.gov (United States)

    Shore, Lawrence

    2012-02-28

    The membrane electrode assembly (MEA) of a PEM fuel cell can be recycled by contacting the MEA with a lower alkyl alcohol solvent which separates the membrane from the anode and cathode layers of the assembly. The resulting solution containing both the polymer membrane and supported noble metal catalysts can be heated under mild conditions to disperse the polymer membrane as particles and the supported noble metal catalysts and polymer membrane particles separated by known filtration means.

  1. Acceptance of failed SNF [spent nuclear fuel] assemblies by the Federal Waste Management System

    International Nuclear Information System (INIS)

    This report is one of a series of eight prepared by E. R. Johnson Associates, Inc. (JAI) under ORNL's contract with DOE's OCRWM Systems Integration Program and in support of the Annual Capacity Report (ACR) Issue Resolution Process. The report topics relate specifically to the list of high priority technical waste acceptance issues developed jointly by DOE and a utility-working group. JAI performed various analyses and studies on each topic to serve as starting points for further discussion and analysis leading eventually to finalizing the process by which DOE will accept spent fuel and waste into its waste management system. The eight reports are concerned with the conditions under which spent fuel and high level waste will be accepted in the following categories: failed fuel; consolidated fuel and associated structural parts; non-fuel-assembly hardware; fuel in metal storage casks; fuel in multi-element sealed canisters; inspection and testing requirements for wastes; canister criteria; spent fuel selection for delivery; and defense and commercial high-level waste packages. This document discusses acceptance of failed spent fuel assemblies by the Federal Waste Management System. 18 refs., 7 figs., 25 tabs

  2. Study on an Extraction Method for a Fuel Rod Image and a Visualization of the Color Information in a Sectional Image of a Spent Fuel Assembly

    International Nuclear Information System (INIS)

    Image processing methods for an extraction of a nuclear fuel rod image and visualization methods of the RGB color data were studied with a sectional image of spent fuel assembly. The fuel rod images could be extracted by using a histogram analysis, an edge detection and RGB rotor data. In these results, a size of the spent fuel assembly could be measured by using a histogram analysis method and a shape of the spent fuel rod could be observed by using an edge detection method. Finally, a various analyses were established for status of the spent fuel assembly by realized various 3D images for the color data in an image of a spent fuel assembly

  3. Contemporary and prospective fuel cycles for VVER-440 based on new assemblies with higher Uranium capacity and higher average fuel enrichment

    International Nuclear Information System (INIS)

    RRC Kurchatov Institute has performed an extensive cycle of calculations intended to validate the opportunities of improving different fuel cycles for VVER-440 reactors. Works were performed to upgrade and improve VVER-440 fuel cycles on the basis of second generation fuel assemblies allowing core thermal power to be up rated to 107%-108% of its nominal value (1375 MW), while maintaining the same fuel operation lifetime. Currently intensive work is underway to develop fuel cycles based on second-generation assemblies with higher fuel capacity and average fuel enrichment per assembly increased up to 4.87% of U-235. Fuel capacity of second-generation assemblies was increased by means of eliminated central apertures of fuel pellets, and pellet diameter extended due to reduced fuel cladding thickness. This paper intends to summarize the results of works performed in the field of VVER-440 fuel cycle modernization, and to present yet unemployed opportunities and prospects of further improvement of VVER-440 neutronic and operating parameters by means of additional optimization of fuel assembly designs and fuel element arrangements applied. (author)

  4. Thermal-Hydraulic Design of Mixed Transition Core for FCM Fuel Assembly

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Hyuk; Seo, K. W; Kim, S. J.; Park, J. P; Hwang, D. H. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Lee, W. J [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    A fully ceramic micro-encapsulated(FCM) fuel based on the dispersed particle fuel concept was considered on the one of accident tolerant fuel(ATF). A mixed core is established using the FCM fuel in the existing LWR core where UO2 fuel pellet was loaded. In order to demonstrate the thermal-hydraulic compatibility of the FCM fuel in the existing LWR core, pin by pin analysis is performed for transition period from mixed core with FCM fuel and UO2 fuel to the FCM fuel only. Parallelized MATRA code using MPI is developed for pin by pin calculation. Pin by pin analysis on mixed transition core for FCM fuel and reference UO2 fuel was performed on a quarter core with 13310 subchannels in assistance with the parallel algorithm with MPI with 20 cores. The thermal margin for the pin by pin model was evaluated by employing a quarter-core power distribution data provided by MASTER code that is nodal code. The pin by pin model shows the feasibility of mixed transition core of FCM fuel assembly based on the MDNBR results.

  5. Dynamics of nuclear fuel assemblies in vertical flow channels: computer modelling and associated studies

    International Nuclear Information System (INIS)

    A computer model, designed to predict the dynamic behaviour of nuclear fuel assemblies in axial flow, is described in this report. The numerical methods used to construct and solve the matrix equations of motion in the model are discussed together with an outline of the method used to interpret the fuel assembly stability data. The mathematics developed for forced response calculations are described in detail. Certain structural and hydrodynamic modelling parameters must be determined by experiment. These parameters are identified and the methods used for their evaluation are briefly described. Examples of typical applications of the dynamic model are presented towards the end of the report. (author)

  6. Study on seismic response characteristics of reactor vessel internals and fuel assembly for OBE elimination

    International Nuclear Information System (INIS)

    To resolve a general argument about OBE elimination for the future nuclear power plant design, seismic responses of reactor vessel internals and fuel assembly for Ulchin nuclear power plant units 3 and 4 in Korea are investigated as an example. Dynamic analyses of the coupled internals and core are performed for the seismic excitations using the reactor vessel motions. By investigating the response relations between OBE and SSE and their response characteristics, the critical components for OBE loading are addressed. Also the fuel assembly responses are calculated using the core plate motions and their behavior is found to be insignificant for OBE elimination. (author)

  7. Sipping equipment for leak testing of fuel assemblies in VVER-440 reactors

    International Nuclear Information System (INIS)

    Sipping equipment for the Soviet-type VVER-440 pressurized water reactors was developed on the basis of the proven in-core sipping technique used for boiling water reactor fuel assemblies. The main components of the system are the sipping hood with seven test positions, the control panel for system operation and sample collection, and the manifold connection line. During testing the upper ends of the hexagonal fuel assemblies are lifted into the air-filled sipping hood to interrupt the coolant flow by means of pneumatically actuated grippers. The first equipment of this kind has been in use in the nuclear plant Jaslovske-Bohunice, Czechoslovakia, since 1986. (orig.)

  8. Three-dimensional flow field analysis of the standard fuel assembly for China advanced research reactor

    International Nuclear Information System (INIS)

    Numerical simulation of the flow field of the standard fuel assembly in China Advanced Research Reactor is carried out by using computational fluid dynamics software CFX4.4 and CFX5.5. The flow distribution and pressure difference of different coolant channels in the standard fuel assembly at rated operating condition are reached. Based on the computational pressure drop results of different flow rate, the resistance characteristic curve is given and compared with experimental results. The two results are in good agreement. (authors)

  9. A method for survey of underwater non-contact deformation of fuel assemblies

    International Nuclear Information System (INIS)

    The survey of underwater non-contact deformation is based on the combined method of the video image-sensor and calibration. After the photography of the targets by five image-sensors at the same time, the images are processed by the image processing system, to get the edge image of the fuel assembly, and then transferred to the video digital signal. The results are analog displayed on the computer, and thus the bending and its future trend of the fuel assembly can be directly displayed. The unknown dimension can be obtained by calibration.(authors)

  10. TVS-A type fuel assembly designed for the VVER-1000 nuclear reactor. Operational experience and fuel preparation for the Temelin NPP

    International Nuclear Information System (INIS)

    The following topics are treated: Introductory description of the assets of the assembly; TVS-A design for the VVER-1000 reactor; Thermal hydraulic properties of the assembly; Results of practical use of the assembly; Fuel cycles with TVS-A assemblies; TVS-A design improvement, TVSA-ALFA design; and TVSA-T fuel assembly designed for the Temelin NPP. Conclusions are as follows: (i) TVS-A assemblies have been used with success at 17 units of the VVER-1000 type. (ii) A 4-year fuel cycle with the replacement of 42 assemblies has been implemented. Currently, transition to an efficient 5-year cycle with the replacement of 36 assemblies is under way. An 18-month fuel cycle (3 x 1.5 years) is also planned. Twenty-one fuel assemblies were in use for 5-7 years, and a burnup of 62 MWd/kgU in a fuel assembly and 72 MWd/kgU in a fuel pin was achieved, the operating period being 2075 effective days. (iii) The new TVSA-ALFA design features a higher uranium content, lower amount of material, low hydraulic resistance and improved thermal properties. The prospects for this new design are supported by 4-year experience with TVSA assembly modifications at the Kalinin NPP. (iv) The TVSA-T fuel assembly design intended for the Temelin NPP is based on the TVSA-ALFA design. While satisfying current requirements for operational flexibility and economy, this design will be compatible with the fuel which is currently in use. (P.A.)

  11. Inspection experience with RA-3 spent nuclear fuel assemblies at CNEA's central storage facility

    International Nuclear Information System (INIS)

    Aluminum-based spent nuclear fuel from Argentina's RA-3 research reactor is to be shipped to the Savannah River Site near Aiken, South Carolina, USA. The spent nuclear fuel contains highly enriched uranium of U.S. origin and is being returned under the US Department of Energy's Foreign Research Reactor/Domestic Research Reactor (FRR/DRR) Receipt Program. An intensive inspection of 207 stored fuel assemblies was conducted to assess shipping cask containment limitations and assembly handling considerations. The inspection was performed with video equipment designed for remote operation, high portability, easy setup and usage. Fuel assemblies were raised from their vertical storage tubes, inspected by remote video, and then returned to their original storage tube or transferred to an alternate location. The inspections were made with three simultaneous video systems, each with dedicated viewing, digital recording, and tele-operated control from a shielded location. All 207 fuel assemblies were safely and successfully inspected in fifteen working days. Total dose to personnel was about one-half of anticipated dose. (author)

  12. Evaluation on Hydrodynamic Masses of Tubular Fuel Assemblies of an Advanced HANARO Reactor (AHR)

    International Nuclear Information System (INIS)

    An Advanced Research Reactor(ARR) is being designed by KAERI since 2002. The final goal of the project is to develop a new and unique research reactor model which is superior in safety and economical aspects. Because many structures of a reactor are surrounded by fluid, we should take fluid effect into consideration when dynamic behavior of a submerged structure is analyzed. In this work, the hydrodynamic mass approach is chosen to analyze structural behavior in fluid and its basic theory is reviewed. We had developed a software of hydrodynamic mass evaluation in the last study. Convergence characteristic of the software was checked with respect to the number of elements. The software was verified to give a reliable result although elements were not so fine. Based on the results, the hydrodynamic mass was estimated to analyze dynamic behavior of the standard and the reduced fuel assembly submerged in water. Despite of occurrence of fluid-induced vibration or earthquake loading, a tubular fuel assembly as well as a fuel channel must keep their structural integrity. Through analysis of the submerged behavior, probable problems have to be identified and their solutions should be reflected on the improved design. Therefore for the first step, hydrodynamic masses were estimated for characteristic sections of the standard and the reduced fuel assemblies. In near future, we will apply the hydrodynamic masses to the dynamic analysis of the submerged fuel assembly

  13. Design analysis of a thorium fueled reactor with seed-blanket assembly configuration

    International Nuclear Information System (INIS)

    Recently, thorium is receiving increasing attention as an important fertile material for the expanding nuclear power programs around the world. The superior nuclear and physical properties of thorium-based fuels could lead to very low fuel cycle cost and make thorium reactors economically attractive. In addition, the use of thorium in reactors would permit more efficient utilization of low cost uranium reserves and reduction of nuclear wastes. In this work, the nuclear characteristics of a new type of thorium fueled reactor (Radkowsky Thorium Reactor) consisting of seed-blanket assemblies are addressed and compared with those of typical assemblies of a PWR (CE type). Also, an assessment on several advantages of thorium fueled reactors is provided. All these results are based on the HELIOS code calculation

  14. Service properties of structural materials of BN-600 reactor fuel assemblies at high damaging doses

    International Nuclear Information System (INIS)

    Based on postradiation investigation of fuel assembly materials of BN-600 reactor a consideration is given to main ways of designing materials which can provide for high burn-up of nuclear fuels in fast reactors. Austenitic steels 08KhN11M3T, 10Kh17N13M2T and ferrite-martensitic steels 1Kh13M213FR, 05Kh12N2M were tested as fuel assembly cans in BN-600 reactor. Austenitic steels EhJ-847, EhP-172, ChS-68 were used for fuel cans. It'is shown that radiation resistance of the steels can be improved by optimization of chemical composition and by enhanced homogeneity of composition, structure and initial mechanical properties. 20 refs.; 4 figs

  15. Criticality safety analysis of IRT-200 storage pool with IRT-4M fuel assemblies

    International Nuclear Information System (INIS)

    In the paper some results of nuclear safety analysis of storage pool with 8-tubes IRT-4M fuel assemblies of the research reactor IRT-200, Sofia, are presented. The calculations have been performed by the modular code system SCALE4.4, which is used worldwide for safety analyses of facilities for transport and storage of spent nuclear fuel. A conservative approach for evaluation of the effective multiplication factor Keff of the storage pool has been applied. The calculations have been carried out for IRT-4M fresh fuel with initial enrichment of 19.75% 235U. The analysis of the obtained results shows that the technological equipment and the storage conditions assure nuclear safety during the storage of IRT-4M spent fuel assemblies in accordance with the Bulgarian norm and standards, which require Keff < 0.95. (author)

  16. Destruction of plutonium using non-uranium fuels in pressurized water reactor peripheral assemblies

    International Nuclear Information System (INIS)

    This thesis examines and confirms the feasibility of using non-uranium fuel in a pressurized water reactor (PWR) radial blanket to eliminate plutonium of both weapons and civilian origin. In the equilibrium cycle, the periphery of the PWR is loaded with alternating fresh and once burned non-uranium fuel assemblies, with the interior of the core comprised of conventional three batch UO2 assemblies. Plutonium throughput is such that there is no net plutonium production: production in the interior is offset by destruction in the periphery. Using this approach a 50 MT WGPu inventory could be eliminated in approximately 400 reactor years of operation. Assuming all other existing constraints were removed, the 72 operating US PWRs could disposition 50 MT of WGPu in 5.6 years. Use of a low fissile loading plutonium-erbium inert-oxide-matrix composition in the peripheral assemblies essentially destroys 100% of the 239Pu and ≥90% totalPu over two 18 month fuel cycles. Core radial power peaking, reactivity vs EFPD profiles and core average reactivity coefficients were found to be comparable to standard PWR values. Hence, minimal impact on reload licensing is anticipated. Examination of potential candidate fuel matrices based on the existing experience base and thermo-physical properties resulted in the recommendation of three inert fuel matrix compositions for further study: zirconia, alumina and TRISO particle fuels. Objective metrics for quantifying the inherent proliferation resistance of plutonium host waste and fuel forms are proposed and were applied to compare the proposed spent WGPu non-uranium fuel to spent WGPu MOX fuels and WGPu borosilicate glass logs. The elimination disposition option spent non-uranium fuel product was found to present significantly greater barriers to proliferation than other plutonium disposal products

  17. Destruction of plutonium using non-uranium fuels in pressurized water reactor peripheral assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Chodak, P. III

    1996-05-01

    This thesis examines and confirms the feasibility of using non-uranium fuel in a pressurized water reactor (PWR) radial blanket to eliminate plutonium of both weapons and civilian origin. In the equilibrium cycle, the periphery of the PWR is loaded with alternating fresh and once burned non-uranium fuel assemblies, with the interior of the core comprised of conventional three batch UO{sub 2} assemblies. Plutonium throughput is such that there is no net plutonium production: production in the interior is offset by destruction in the periphery. Using this approach a 50 MT WGPu inventory could be eliminated in approximately 400 reactor years of operation. Assuming all other existing constraints were removed, the 72 operating US PWRs could disposition 50 MT of WGPu in 5.6 years. Use of a low fissile loading plutonium-erbium inert-oxide-matrix composition in the peripheral assemblies essentially destroys 100% of the {sup 239}Pu and {ge}90% {sub total}Pu over two 18 month fuel cycles. Core radial power peaking, reactivity vs EFPD profiles and core average reactivity coefficients were found to be comparable to standard PWR values. Hence, minimal impact on reload licensing is anticipated. Examination of potential candidate fuel matrices based on the existing experience base and thermo-physical properties resulted in the recommendation of three inert fuel matrix compositions for further study: zirconia, alumina and TRISO particle fuels. Objective metrics for quantifying the inherent proliferation resistance of plutonium host waste and fuel forms are proposed and were applied to compare the proposed spent WGPu non-uranium fuel to spent WGPu MOX fuels and WGPu borosilicate glass logs. The elimination disposition option spent non-uranium fuel product was found to present significantly greater barriers to proliferation than other plutonium disposal products.

  18. MELCOR model for an experimental 17x17 spent fuel PWR assembly.

    Energy Technology Data Exchange (ETDEWEB)

    Cardoni, Jeffrey

    2010-11-01

    A MELCOR model has been developed to simulate a pressurized water reactor (PWR) 17 x 17 assembly in a spent fuel pool rack cell undergoing severe accident conditions. To the extent possible, the MELCOR model reflects the actual geometry, materials, and masses present in the experimental arrangement for the Sandia Fuel Project (SFP). The report presents an overview of the SFP experimental arrangement, the MELCOR model specifications, demonstration calculation results, and the input model listing.

  19. Solution High-Energy Burst Assembly (SHEBA) results from subprompt critical experiments with uranyl fluoride fuel

    Energy Technology Data Exchange (ETDEWEB)

    Cappiello, C.C.; Butterfield, K.B.; Sanchez, R.G.; Bounds, J.A.; Kimpland, R.H.; Damjanovich, R.P.; Jaegers, P.J.

    1997-08-01

    Experiments were performed to measure a variety of parameters for SHEBA: behavior of the facility during transient and steady-state operation; characteristics of the SHEBA fuel; delayed-critical solution height vs solution temperature; initial reactor period and reactivity vs solution height; calibration of power level vs reactor power instrumentation readings; flux profile in SHEBA; radiation levels and neutron spectra outside the assembly for code verification and criticality alarm and dosimetry purposes; and effect on reactivity of voids in the fuel.

  20. On-line system for monitoring of boiling in nuclear reactor fuel assemblies

    International Nuclear Information System (INIS)

    An important goal of nuclear reactor instrumentation is the continuous monitoring of the state of the reactor and the detection of deviations from the normal behaviour at an early stage. Early detection of anomalies enables one to make the necessary steps in order to prevent further damage of nuclear fuel. In the present paper, an on-line core monitoring system is described by means of which boiling anomaly in nuclear reactor fuel assemblies can be detected. (author). 9 refs, 7 figs

  1. Solution High-Energy Burst Assembly (SHEBA) results from subprompt critical experiments with uranyl fluoride fuel

    International Nuclear Information System (INIS)

    Experiments were performed to measure a variety of parameters for SHEBA: behavior of the facility during transient and steady-state operation; characteristics of the SHEBA fuel; delayed-critical solution height vs solution temperature; initial reactor period and reactivity vs solution height; calibration of power level vs reactor power instrumentation readings; flux profile in SHEBA; radiation levels and neutron spectra outside the assembly for code verification and criticality alarm and dosimetry purposes; and effect on reactivity of voids in the fuel

  2. Melt-dilute treatment of spent nuclear fuel assemblies from research and test reactors

    International Nuclear Information System (INIS)

    The Savannah River Site is the U.S. Department of Energy's preferred site for return and treatment of all aluminum-base, spent, research and test reactor fuel assemblies. There are over 20,000 spent fuel assemblies now stored in different countries around the world, and by 2035 many will be returned to SRS for treatment and interim storage, in preparation for disposal in a geologic repository. The early fuel assemblies for research and test reactors were made using aluminum clad plates that were fabricated from highly enriched (93%) uranium-aluminum alloy. Later, powder metallurgical fabrication methods were developed to produce plate fuels with higher uranium contents using either uranium aluminide, uranium oxide or uranium silicide powders mixed with aluminum. Silicide fuel elements generally are fabricated with low enriched uranium containing less than 20% 2'35U. Following irradiation, the spent fuel assemblies are discharged from the reactor, and most assemblies have been stored in underwater pools, some since the early 1950's. A number of disposition options including direct/co-disposal and melt-dilute treatment were evaluated recently. The melt-dilute technique was identified as the preferred method for treatment of aluminum-base spent fuel. The technique consists of melting the spent fuel assembly and adding depleted uranium to the melt for isotopic dilution to 2'35U. Aluminum is added, if necessary, to produce a predetermined alloy composition. Additionally, neutron poisons may be added to the melt where they form solid solution phases or compounds with uranium and/or aluminum. Lowering the enrichment reduces both criticality and proliferation concerns for storage. Consolidation by melting also reduces the number of storage canisters. Laboratory and small-scale process demonstration using irradiated fuel is underway. Tests of the off gas absorption system have been initiated using both surrogate and irradiated RERTR mini fuel plates. An experimental L

  3. Development of a Hi per PWR Fuel assembly for OPR1000 and APR1400

    International Nuclear Information System (INIS)

    KEPCO Nuclear Fuel Co., Ltd. (KNF) launched a new advanced high performance fuel development project in September of 2005, and the HIPER (High Performance with Efficiency and Reliability) fuel was successfully developed in 2010, accomplishing a series of out-of-pile tests. The HIPER Lead Test Assemblies (LTAs) were fabricated in early 2011, and LTA in-reactor verification testing was started in the mid of 2011. The LTAs are scheduled to undergo in-reactor verification testing over four years approximately, and then HIPER fuel will be supplied commercially after obtaining the commercial supply license. In this paper, the development objectives, main design features, including PWR fuel technology development strategy in Korea, out-of-pile test outlines and in-pile test plan for HIPER fuel are described

  4. Nuclear fuel assembly spacer and loop spring with enhanced flexibility

    International Nuclear Information System (INIS)

    This patent describes a fuel bundle having a surrounding channel, and upper tie plat, a lower tie plate, fuel rods and a pari of water rods, wherein the fuel rods are arranged in a matrix and positioned in parallel upstanding relation to one another between the upper and lower tie plates, the pair of water rods are arranged side-by-side and positioned centrally of the fuel bundle such that the water rod pair occupies seven matrix positions centrally of the matrix, the pair including a first water rod occupying a first three positions in the matrix and part of a forth position, the second water rod of the water rod pair occupying a second three positions in the matrix and part of the fourth position, a system for the securing of the paired water rods comprising in combination: first and second U-shaped spacers, each spacer having a portion that extends between the water rods, each portion bearing against each water rod to maintain the water rods in spaced relationships; and means for biasing the water rods against the U-shaped spacers

  5. Report of lower endplug welding, and testing and inspecting result for MONJU 1th reload core fuel assembly

    International Nuclear Information System (INIS)

    The procedure and result of lower endplugwelding, Test and Inspection and Shipment of the 1th reload core fuel assembly (80 Fuel Assemblies) for the fast breeder reactor MONJU are reported, which had been examined and inspected in Tamatsukuri Branch, Material Insurance Office, Quality Assurance Section, Technical Administration Division, Plutonium Fuel Center (before: Inspection Section, Plutonium Fuel Division), from June 1994 to January 1996. The number of cladding tubes welded to the endplug were totally 13,804: 7,418 for Core - Inside of 43 fuel Assemblies and 6,836 for Core-Outside of 37 fuel Assemblies. 13,794 of them, 7,414 Core-Inside and 6,379 Core-Outside, were approved by the test and sent to Plutonium Fuel Center. 10 of them weren't approved mainly because of default welding. Disapproval rating was 0.07%. (author)

  6. A Unique Hybrid Propulsion System Design for Large Space Boosters

    Science.gov (United States)

    Rodgers, Frederick C.

    1990-01-01

    A study was made of the application of hybrid rocket propulsion technology to large space boosters. Safety, reliability, cost, and performance comprised the evaluation criteria, in order of relative importance, for this study. The effort considered the so called classic hybrid design approach versus a novel approach which utilizes a fuel-rich gas generator for the fuel source. Other trades included various fuel/oxidizer combinations, pressure-fed versus pump fed oxidizer delivery systems, and reusable versus expandable booster systems. Following this initial trade study, a point design was generated. A gas generated-type fuel grain with pump fed liquid oxygen comprised the basis of this point design. This design study provided a mechanism for considering the means of implementing the gas generator approach for further defining details of the design. Subsequently, a system trade study was performed which determined the sensitivity of the design to various design parameters and predicted optimum values for these same parameters. The study concluded that a gas generator hybrid booster design offers enhanced safety and reliability over current of proposed solid booster designs while providing equal or greater performance levels. These improvements can be accomplished at considerably lower cost than for the liquid booster designs of equivalent capability.

  7. Method and apparatus for measuring subchannel voids in a light water reactor test fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Shiraishi, L.M.; Wilhelmson, D.A.; Matzner, B.

    1992-03-24

    This patent describes a nuclear fuel bundle test assembly. It comprises: upstanding vertical rods, the rods having an outside diameter the same as the outside diameter of fuel rods within a nuclear fuel bundle; a lower support plate for supporting the vertical rods and providing current to the lower end of the vertical upstanding rods, the lower support plate exposing the lower open ends of the vertical upstanding rods to enable the entrance and exit of testing probes to the interior of the vertical upstanding rods; an upper support plate for maintaining the rods in vertical upstanding relation and providing current to the upper end of the vertical upstanding rods; a power source connected across the upper and lower support plates for supplying power to the test assembly to produce heat at the individual upstanding rods; a test chamber surrounding the vertical upstanding rods for permitting inflow of water at the bottom of the rods and the outflow of water and steam at the upper end of the vertical upstanding rods to enable the test assembly to emulate a nuclear fuel bundle, the improvement to the nuclear fuel bundle test assembly comprising: first and second probes for vertical excursion interior of first and second upstanding vertical rods; a gamma ray source mounted to one of the probes, the radioactive source emitting particles attenuated by passage through water; a Geiger-Muller detector mounted to the other of the probes, the Geiger-Muller detector for insertion to the electrically heated tube within a nuclear fuel bundle test assembly, the detector comprising: a Geiger-Muller counter; a conducting and shielding material surrounding the Geiger-Muller counter, the material having a window therein for permitting collimated radiation to be received to the counter.

  8. Assessment of dry storage performance of spent LWR fuel assemblies with increasing burnup

    International Nuclear Information System (INIS)

    To assess the extended storage performance of spent LWR-fuel, the available experience can be collated into 3 storage modes: mode I: fast decrease rate of temperature between maximum of licensed dry storage temperature and 300 deg. C; mode II: medium decrease rate of the fuel rod dry storage temperature between 300 deg. C and 200 deg. C; mode III: slow to negligible decrease rate of fuel rod dry storage temperature for temperatures less than 200 deg. C. Mode I is typical for early interim storage, mode III covers extremely long term storage which is encountered presumably for nearly all dry storage extensions to be considered. Mode II dry storage is characterised by the fact that all creep deformations of the spent fuel cladding can already be regarded as terminated as well as the corrosive attack of the cladding. Reviewing the fission product behaviour under dry storage conditions it can be pointed out that the fission products generated in the UO2-fuel under in service conditions are practically immobile in the UO2-fuel lattice during storage. Consequently all fission product driven defect mechanisms like stress corrosion cracking (SCC), uniform fuel rod internal fission product corrosion of the cladding, localised fuel rod internal fission product corrosion of the cladding, will not take place. The leading defect mechanism for spent fuel rod in dry storage - also for fuel rod with increased burn-up - remains creep due to the hoop strain resulting from the fuel rod internal fission gas pressure. Limiting the creep to its primary and secondary stages prevents fuel rod degradation. Post-pile creep of fuel rod cladding can be described conservatively by the creep of unirradiated cladding. The allowable uniform strain of the cladding in its typical post-pile condition preventing tertiary creep under dry spent fuel storage conditions is 1-2%. Dry storage performance prediction of fuel assemblies with a burn-up ≤ 65 GWd/tHM was calculated based on the fuel assemblies

  9. Assembly of laboratory line for nuclear fuel reprocessing

    International Nuclear Information System (INIS)

    The dismantling of a laboratory line for spent fuel reprocessing after the termination of the research programme and the procedures for hot and semi-hot cell decontamination are described. The equipment was mostly disassembled in smaller parts which were then decontaminated by wiping them with cotton wool soaked in detergent and citric acid, varnished with two-component epoxi varnish, wrapped into multiple polyethylene foils, sealed in PVC bags and thus ready for transport. (B.S.)

  10. Pulsation characteristics of boiling water cooled reactor two fuel assembly model

    International Nuclear Information System (INIS)

    The results of experimental studies into the pulsation characteristics of the natural circulation circuit model for the boiling water cooled reactor are given. Influence of nonidentity of fuel assembly power on stability of coolant flow rate was investigated. The methods for avoiding the whole circuit and interassembly hydrodynamic instabilities are suggested

  11. Electric power generation by a submersible microbial fuel cell equipped with a membrane electrode assembly

    DEFF Research Database (Denmark)

    Min, Booki; Poulsen, Finn Willy; Thygesen, Anders;

    2012-01-01

    Membrane electrode assemblies (MEAs) were incorporated into the cathode chamber of a submersible microbial fuel cell (SMFC). A close contact of the electrodes could produce high power output from SMFC in which anode and cathode electrodes were connected in parallel. In polarization test, the...

  12. The procedure of computational evaluation of the margin to CHF in new generation WWER fuel assemblies

    International Nuclear Information System (INIS)

    A modified and upgraded empirical procedure of the Institute for Physics and Power Engineering (SSC RF-IPPE) has been presented, applicable to the grids-enhancers (EG) of different types located at random over the length of fuel assembly (FA), which allows its application for optimizing the FA and EG designs. (author)

  13. Calculate Some Characteristic Parameters Of VVER-1000's Fuel Assembly By MCNP4C2 Code

    International Nuclear Information System (INIS)

    This report presents the descriptions of parameters characteristics of the LEU and MOX Fuel Assemblies of VVER-1000 reactor, and calculation results such as infinite neutron multiplication factor kinf, two groups energies constants, neutron flux distribution by using Monte Carlo code MCNP. (author)

  14. Fuel cell system including a unit for electrical isolation of a fuel cell stack from a manifold assembly and method therefor

    Science.gov (United States)

    Kelley; Dana A. , Farooque; Mohammad , Davis; Keith

    2007-10-02

    A fuel cell system with improved electrical isolation having a fuel cell stack with a positive potential end and a negative potential, a manifold for use in coupling gases to and from a face of the fuel cell stack, an electrical isolating assembly for electrically isolating the manifold from the stack, and a unit for adjusting an electrical potential of the manifold such as to impede the flow of electrolyte from the stack across the isolating assembly.

  15. Visual inspection system for radioactive fuel assemblies using fiberoptics

    International Nuclear Information System (INIS)

    A system is described for the visual inspection of a radioactive assembly of tubes in an underwater environment, comprising an elongated fiberoptic image guide for remotely transmitting an image of the tubes to be visually inspected, a light source for emitting light, an elongated fiberoptic light guide for transmitting the light emitted from the light source to the tubes. The distal end of the image guide is parallel and adjacent to one side of the light guide so that the distal ends of the image guide and light guide present an elongated cross section that facilitates the insertion of the distal ends in the spaces between the tubes, a first receiving means for remotely displaying the image conducted by the image guide, a second receiving means for remotely displaying the location of the distal ends of the adjacent image and light guides to facilitate the positioning thereof within the assembly, a positioning means for remotely positioning the distal ends of the image and light guide, and means for mechanically linking the second receiving means with the positioning means so that when the image and light guides are moved, the second receiving means is moved the same amount

  16. Nuclear fuel assembly with upholding structure and anti-flight device

    International Nuclear Information System (INIS)

    The fuel assembly is made by two distinct substructures which can move vertically relatively to each other. The first substructure is made of the top end of the assembly joined by some of the tie bars to a mobile plate. These tie bars are solidly attached to the grids that the fuel pins in position. The second substructure is made of the other tie bars and a casing containing springs. These tie bars are free to move vertically through the top end of the assembly and the grids. The first substructure is held against the upper plate of the core by the hydraulic pressure of the coolant and the second substructure is held against the bottom core support plate by the springs

  17. Nuclear fuel assembly with a shock absorber system especially for seismic shocks

    International Nuclear Information System (INIS)

    Hydraulic system for absorbing impacts imparted to fuel assemblies. Internal buffers that brace the fuel assemblies so as to restrain their longitudinal displacement, rest against mobile spring buffers. Some of these spring buffers have plungers sliding in hollow tubular guide uprights provided with longitudinal slots. The end of each upright is closed by a plate fitted with an orifice. When an earthquake forces the plunger and spring buffer assemblies to move, the water under pressure in the upright guide tubes stop this movement. This water, which gushes from the holes in the plates, enables the displacement to take place at a controlled rate at which the forces applied are absorbed in complete safety. The plungers gradually close up the slots in the guide uprights, thereby progressively reducing the section through which the water inside the guide upright can flow out. The resistance increases progressively and protects the structure of the reactor core

  18. Experimental investigation of a representative PWR nuclear fuel assembly spacer grid

    Energy Technology Data Exchange (ETDEWEB)

    Castro, Higor Fabiano Pereira de; Mesquita, Amir Zacarias; Navarro, Moyses A.; Mattos, Joao Roberto Loureiro de; Santos, Andre A. Campagnole dos, E-mail: higorfabiano@hotmail.com, E-mail: amir@cdtn.br, E-mail: moysesnavarro@yahoo.com.br, E-mail: jrmattos@cdtn.br, E-mail: aacs@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2015-07-01

    The spacer grids are important structures present in nuclear fuel assembly from Pressurized Water Reactors (PWR). They play an important structural role and also assist in heat removal through the assembly by promoting increased turbulence of the flow. However the presence of spacer grids in the fuel assembly causes a localized pressure drop. In this paper we present experimental results of the water flow velocity profiles for five heights from a spacer grid present in a 5 x 5 rod bundle. These velocity profiles were obtained using a LDV (Laser Doppler Velocimetry). The tests were conducted for Reynolds numbers ranging from 1.8 x 10{sup 4} to 5.4 x 10{sup 4}. This experimental research was carried out in thermo-hydraulics laboratory of Nuclear Technology Development Center - CDTN. This experimental research also assists in CFD - Computational Fluid Dynamics numerical analysis process which is also developed at CDTN. (author)

  19. Intermediate review on the transportation of spent fuel assemblies

    International Nuclear Information System (INIS)

    The transportation of spent fuel from the Swiss nuclear power plants to the reprocessing facilities in France and England was interrupted in May 1998 because of contamination that occurred. These measures were presented in the March 1999 statement made by the Office for the Safety of Nuclear Plants (HSK). The transport of spent fuel has been once more permitted and carried out under new conditions since August 1999. In its interim report of October 2000, HSK analyses and evaluates the experience gained since the resumption of transports. For each measure required, it compares the advantages and drawbacks and makes decisions on the maintenance or reduction of the measures to be taken. Between August 1999 and July 2000, 12 spent fuel transports were carried out between the Swiss nuclear power plants and the COGEMA reprocessing facility in France (7 from Goesgen, 4 from Beznau and 1 from Leibstadt). Neither noticeable disagreement with nor exceeding of contamination limits were noted during those 12 transports. This satisfactory result demonstrates that the measures required to be taken are effective. HSK expected from the measures a reduction of the frequency of exceeding contamination limits to less than 5% and also a marked reduction in their frequency. The present results correspond to this expectation; however, the statistical basis is not yet sufficient to be able to draw definitive conclusions. Nevertheless it is noticed that the situation in France, where similar measures have been taken, was very clearly improved. The frequency of exceeding contamination limits was reduced to 2% during the first semester of the year 2000, while it amounted to more than 30% before April 1998. It is the comprehensiveness of the measures required by HSK which allows the avoidance of contamination. The analysis shows that just a small number of measures only contribute insignificantly to the goal sought after. Therefore, two measures will be suppressed (packing of the empty

  20. Numerical analyses of flow distributions in nuclear fuel assemblies affected by grid deformations

    International Nuclear Information System (INIS)

    Highlights: • Deformed spacer grid of a fuel assembly restricts coolant flow. • CFD analyses are conducted to assess flow redistribution and recovery. • Flow field is analyzed for normal operation, blowdown and reflood phases. • Forty-five times hydraulic diameter is required to recover 95% of flow rate. - Abstract: In the event of a safety shutdown earthquake (SSE) in a nuclear power plant, the spacer grid of the fuel assembly will be deformed as a result of the vibrations. If the flow area in a subchannel is reduced due to the grid deformation, the coolant flow will be restricted and consequently a loss of flow occurs in the affected fuel assembly during the accident. In this study, computational fluid dynamics (CFD) analyses are conducted in order to assess the flow redistribution and flow recovery in fuel assemblies. The real geometries of an outer grid and mixing vane are used in the simulation, and the region including the inner grid is modeled as a porous media zone. The resistance coefficients of the porous media model are determined using CFD analyses. The Reynolds-averaged Navier–Stokes equation with a non-linear turbulence model was used to solve the three-dimensional anisotropic turbulence flow in the rod bundles during normal operation, blowdown, and reflood phases following a loss-of-coolant accident (LOCA). In these analyses, it is assumed that forty percent of the flow area is blocked by grid deformations. The results demonstrate that a downstream distance of 45 times the hydraulic diameter is required for the coolant flow to recover to 95% of the original flow rate in the affected fuel assembly