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Sample records for reactor fuel assembly

  1. Nuclear reactor fuel assembly

    International Nuclear Information System (INIS)

    Sasaki, Y.; Tashima, J.

    1975-01-01

    A description is given of nuclear reactor fuel assemblies arranged in the form of a lattice wherein there is attached to the interface of one of two adjacent fuel assemblies a plate spring having a concave portion curved toward said interface and to the interface of the other fuel assembly a plate spring having a convex portion curved away from said interface

  2. Nuclear reactor fuel assembly

    International Nuclear Information System (INIS)

    Sakurai, Shungo; Ogiya, Shunsuke.

    1990-01-01

    In a fuel assembly, if the entire fuels comprise mixed oxide fuels, reactivity change in cold temperature-power operation is increased to worsen the reactor shutdown margin. The reactor shutdown margin has been improved by increasing the burnable poison concentration thereby reducing the reactivity of the fuel assembly. However, since unburnt poisons are present at the completion of the reactor operation, the reactivity can not be utilized effectively to bring about economical disadvantage. In view of the above, the reactivity change between lower temperature-power operations is reduced by providing a non-boiling range with more than 9.1% of cross sectional area at the inside of a channel at the central portion of the fuel assembly. As a result, the amount of the unburnt burnable poisons is decreased, the economy of fuel assembly is improved and the reactor shutdown margin can be increase. (N.H.)

  3. Nuclear reactor fuel assembly

    International Nuclear Information System (INIS)

    Marmonier, Pierre; Mesnage, Bernard; Nervi, J.C.

    1975-01-01

    This invention refers to fuel assemblies for a liquid metal cooled fast neutron reactor. Each assembly is composed of a hollow vertical casing, of regular polygonal section, containing a bundle of clad pins filled with a fissile or fertile substance. The casing is open at its upper end and has a cylindrical foot at its lower end for positioning the assembly in a housing provided in the horizontal diagrid, on which the core assembly rests. A set of flat bars located on the external surface of the casing enables it to be correctly orientated in its housing among the other core assemblies [fr

  4. Nuclear reactor fuel assembly

    International Nuclear Information System (INIS)

    Vikhorev, Yu.V.; Biryukov, G.I.; Kirilyuk, N.A.; Lobanov, V.N.

    1977-01-01

    A fuel assembly is proposed for nuclear reactors allowing remote replacement of control rod bundles or their shifting from one assembly to another, i.e., their multipurpose use. This leads to a significant increase in fuel assembly usability. In the fuel assembly the control rod bundle is placed in guide tube channels to which baffles are attached for fuel element spacing. The remote handling of control rods is provided by a hollow cylinder with openings in its lower bottom through which the control rods pass. All control rods in a bundle are mounted to a cross beam which in turn is mounted in the cylinder and is designed for grasping the whole rod bundle by a remotely controlled telescopic mechanism in bundle replacement or shifting. (Z.M.)

  5. Nuclear reactor fuel assembly

    International Nuclear Information System (INIS)

    1975-01-01

    A description is given of a nuclear reactor fuel assembly comprising a cluster of fuel elements supported by transversal grids so that their axes are parallel to and at a distance from each other, in order to establish interstices for the axial flow of a coolant. At least one of the interstices is occupied by an axial duct reserved for an auxiliary cooling fluid and is fitted with side holes through which the auxiliary cooling fluid is sprayed into the cluster. Deflectors extend as from a transversal grid in a position opposite the holes to deflect the cooling fluid jet towards those parts of the fuel elements that are not accessible to the auxiliary coolant. This assembly is intended for reactors cooled by light or heavy water [fr

  6. Reactor and fuel assembly

    International Nuclear Information System (INIS)

    Ishii, Yoshihiko; Bessho, Yasunori; Sano, Hiroki; Yokomizo, Osamu; Yamashita, Jun-ichi.

    1990-01-01

    The present invention realizes an effective spectral operation by applying an optimum pressure loss coefficient while taking the characteristics of a lower tie plate into consideration. That is, the pressure loss coefficient of the lower tie plate is optimized by varying the cross sectional area of a fuel assembly flow channel in the lower tie plate or varying the surface roughness of a coolant flow channel in the lower tie plate. Since there is a pressure loss coefficient to optimize the moderator density over a flow rate change region, the effect of spectral shift rods can be improved by setting the optimum pressure loss coefficient of the lower tie plate. According to the present invention, existent fuel assemblies can easily be changed successively to fuel assemblies having spectral shift rods of a great spectral shift effect by using existent reactor facilities as they are. (I.S.)

  7. Fuel assembly and reactor core

    International Nuclear Information System (INIS)

    Aoyama, Motoo; Koyama, Jun-ichi; Uchikawa, Sadao; Bessho, Yasunori; Nakajima, Akiyoshi; Maruyama, Hiromi; Ozawa, Michihiro; Nakamura, Mitsuya.

    1990-01-01

    The present invention concerns fuel assemblies charged in a BWR type reactor and the reactor core. The fuel assembly comprises fuel rods containing burnable poisons and fuel rods not containing burnable poisons. Both of the highest and the lowest gadolinia concentrations of the fuel rods containing gadolinia as burnable poisons are present in the lower region of the fuel assembly. This can increase the spectral shift effect without increasing the maximum linear power density. (I.N.)

  8. Reactor fuel assembly

    International Nuclear Information System (INIS)

    Anthony, A.J.; Groves, M.D.

    1980-01-01

    A nuclear reactor fuel assembly having a lower end fitting and actuating means interacting therewith for holding the assembly down on the core support stand against the upward flow of coolant. Locking means for interacting with projections on the support stand are carried by the lower end fitting and are actuated by the movement of an actuating rod operated from above the top of the assembly. In one embodiment of the invention the downward movement of the actuating rod forces a latched spring to move outward into locking engagement with a shoulder on the support stand projections. In another embodiment, the actuating rod is rotated to effect the locking between the end fitting and the projection. (author)

  9. Fuel assembly in a reactor

    International Nuclear Information System (INIS)

    Saito, Shozo; Kawahara, Akira.

    1975-01-01

    Object: To provide a fuel assembly in a reactor which can effectively prevent damage of the clad tube caused by mutual interference between pellets and the clad tube. Structure: A clad tube for a fuel element, which is located in the outer peripheral portion, among the fuel elements constituting fuel assemblies arranged in assembled and lattice fashion within a channel box, is increased in thickness by reducing the inside diameter thereof to be smaller than that of fuel elements internally located, thereby preventing damage of the clad tube resulting from rapid rise in output produced when control rods are removed. (Kamimura, M.)

  10. Fuel assembly and reactor core

    International Nuclear Information System (INIS)

    Yuchi, Yoko; Aoyama, Motoo; Haikawa, Katsumasa; Yamanaka, Akihiro; Koyama, Jun-ichi.

    1996-01-01

    In a fuel assembly of a BWR type reactor, a region substantially containing burnable poison is divided into an upper region and a lower region having different average concentrations of burnable poison along a transverse cross section perpendicular to the axial direction. The ratio of burnable poison contents of both regions is determined to not more than 80%, and the average concentration of the burnable poison in the lower region is determined to not less than 9% by weight. An infinite multiplication factor at an initial stage of the burning of the fuel assembly is controlled effectively by the burnable poisons. Namely, the ratio of the axial power can be controlled by the distribution of the enrichment degree of uranium fuels and the distribution of the burnable poison concentration in the axial direction. Since the average enrichment degree of the reactor core has to be increased in order to provide an initially loaded reactor core at high burnup degree. Distortion of the power distribution in the axial direction of the reactor core to which fuel assemblies at high enrichment degree are loaded is flattened to improve thermal margin, to extend continuous operation period and increase a burnup degree upon take-out thereby improving fuel economy without worsening the reactor core characteristics of the initially loaded reactor core. (N.H.)

  11. Reactor fuel assembly fastening

    International Nuclear Information System (INIS)

    Formanek, F.J.; Schukei, G.E.

    1980-01-01

    A nuclear fuel assembly is described, adapted to be locked into first mating surfaces on a core support stand, comprising a lower end fitting having posts for resting on the stand; elongated hook members pivotally connected at one end to the lower end fitting and having a second mating surface at the other end to engage the first mating surfaces; actuating means located between the posts on the lower end fitting and being vertically movable relative to the end fitting; and rigid links pivotally attached at one end to the hook members intermediate the connection of the hook members to the end fitting and the second mating surface and pivotally attached at the other end to the actuating means, the link having a length between the pivoted connections such that the second mating surface on the hook members locks into engagement with the first mating surfaces on the stand as the links approach the horizontal. (author)

  12. Fuel assemblies for nuclear reactor

    International Nuclear Information System (INIS)

    Nishi, Akihito.

    1987-01-01

    Purpose: To control power-up rate at the initial burning stage of new fuel assemblies due to fuel exchange in a pressure tube type power reactor. Constitution: Burnable poisons are disposed to a most portion of fuel pellets in a fuel assembly to such a low concentration as the burn-up rate changes with time at the initial stage of the burning. The most portion means substantially more than one-half part of the pellets and gadolinia is used as burn-up poisons to be dispersed and the concentration is set to less than about 0.2 %. Upon elapse of about 15 days after the charging, the burnable poisons are eliminated and the infinite multiplication factors are about at 1.2 to attain a predetermined power state. Since the power-up rate of the nuclear reactor fuel assembly is about 0.1 % power/hour and the power-up rate of the fuel assembly around the exchanged channel is lower than that, it can be lowered sufficiently than the limit for the power-up rate practiced upon reactor start-up thereby enabling to replace fuels during power operation. (Horiuchi, T.)

  13. Fuel assemblies for nuclear reactors

    International Nuclear Information System (INIS)

    Jabsen, F.S.

    1979-01-01

    In a nuclear fuel assembly, hollow guide posts protrude into a fuel assembly and fitting grill from a biased spring pad with a plunger that moves with the spring pad plugging one end of each of the guide posts. A plate on the end fitting grill that has a hole for fluid discharge partially plugs the other end of the guide post. Pressurized water coolant that fills the guide post volume acts as a shock absorber and should the reactor core receive a major seismic or other shock, the fuel assembly is compelled to move towards a pad depending from a transversely disposed support grid. The pad bears against the spring pad and the plunger progressively blocks the orifices provided by slots in the guide posts thus gradually absorbing the applied shock. After the orifice has been completely blocked, controlled fluid discharge continues through a hole coil spring cooperating in the attenuation of the shock. (author)

  14. Simulated nuclear reactor fuel assembly

    International Nuclear Information System (INIS)

    Berta, V.T.

    1993-01-01

    An apparatus for electrically simulating a nuclear reactor fuel assembly. It includes a heater assembly having a top end and a bottom end and a plurality of concentric heater tubes having electrical circuitry connected to a power source, and radially spaced from each other. An outer target tube and an inner target tube is concentric with the heater tubes and with each other, and the outer target tube surrounds and is radially spaced from the heater tubes. The inner target tube is surrounded by and radially spaced from the heater tubes and outer target tube. The top of the assembly is generally open to allow for the electrical power connection to the heater tubes, and the bottom of the assembly includes means for completing the electrical circuitry in the heater tubes to provide electrical resistance heating to simulate the power profile in a nuclear reactor. The embedded conductor elements in each heater tube is split into two halves for a substantial portion of its length and provided with electrical isolation such that each half of the conductor is joined at one end and is not joined at the other end

  15. Fuel assembly for a nuclear reactor

    International Nuclear Information System (INIS)

    Gjertsen, R.K.; Tower, S.N.; Huckestein, E.A.

    1982-01-01

    A fuel assembly for a nuclear reactor comprises a 5x5 array of guide tubes in a generally 20x20 array of fuel elements, the guide tubes being arranged to accommodate either control rods or water displacer rods. The fuel assembly has top and bottom Inconel (Registered Trade Mark) grids and intermediate Zircaloy grids in engagement with the guide tubes and supporting the fuel elements and guide tubes while allowing flow of reactor coolant through the assembly. (author)

  16. Fuel assemblies for nuclear reactors

    International Nuclear Information System (INIS)

    Leclercg, J.

    1985-01-01

    Improvements to guide tubes for the fuel assemblies of light water nuclear reactors, said assemblies being immersed in operation in the cooling water of the core of such a reactor, the guide tubes being of the type made from zircaloy and fixed at their two ends respectively to an upper end part and a lower end part made from stainless steel or Irconel and which incorporate devices for braking the fall of the control rods which they house during the rapid shutdown of the reactor, wherein the said braking devices are constituted by means for restricting the diameter of the guide tubes comprising for each guide tube a zircaloy inner sleeve spot welded to the said guide tube and whose internal diameter permits the passage, with a calibrated clearance, of the corresponding control rod, the sleeve being distributed over the lower portion of each guide tube and associated with orifices made in the actual guide tubes to produce the progressive hydraulic absorption of the end of the fall of the control rods

  17. Fuel assemblies for use in nuclear reactors

    International Nuclear Information System (INIS)

    Schluderberg, D.C.

    1981-01-01

    A fuel assembly for use in pressurized water cooled nuclear fast breeder reactors is described in which moderator to fuel ratios, conducive to a high Pu-U-D 2 O reactor breeding ratio, are obtained whilst at the same time ensuring accurate spacing of fuel pins without the parasitic losses associated with the use of spacer grids. (U.K.)

  18. Fuel assembly for a nuclear reactor

    International Nuclear Information System (INIS)

    Gjertsen, R.K.

    1982-01-01

    A fuel assembly in a nuclear reactor comprises a locking mechanism that is capable of locking the fuel assembly to the core plate of a nuclear reactor to prevent inadvertent movement of the fuel assembly. The locking mechanism comprises a ratchet mechanism 108 that allows the fuel assembly to be easily locked to the core plate but prevents unlocking except when the ratchet is disengaged. The ratchet mechanism is coupled to the locking mechanism by a rotatable guide tube for a control rod or water displacer rod. (author)

  19. Nuclear reactor fuel assembly grid

    International Nuclear Information System (INIS)

    Alder, J.L.; Kmonk, S.; Racki, F.R.

    1981-01-01

    A grid for a nuclear reactor fuel assembly which includes intersecting straps arranged to form a structure of egg crate configuration. The cells defined by the intersecting straps are adapted to contain axially extending fuel rods, each of which occupy one cell, while each control rod guide tube or thimble occupies the space of four cells. To effect attachment of each guide thimble to the grid, a short intermediate sleeve is brazed to the strap walls and the guide thimble is then inserted therein and mechanically secured to the sleeve walls. Each sleeve preferably, although not necessarily, is equipped with circumferentially spaced openings useful in adjusting dimples and springs in adjacent cells. To accurately orient each sleeve in position in the grid, the ends of straps extending in one direction project through transversely extending straps and terminate in the wall of the guide sleeve. Other straps positioned at right angles thereto terminate in that portion of the wall of a strap which lies next to a wall of the sleeve

  20. Fuel assembly and reactor core

    International Nuclear Information System (INIS)

    Moriwaki, Masanao; Aoyama, Motoo; Masumi, Ryoji; Ishibashi, Yoko.

    1995-01-01

    A fuel assembly comprises a plurality of fuel rods filled with nuclear fuels, a plurality of burnable poison-incorporated fuel rods and a spectral shift-type water rod. As the burnable poison for the burnable poison-incorporated fuel rod, a plurality of burnable poison elements each having a different neutron absorption cross section are used. A burnable poison element such as boron having a relatively small neutron absorbing cross section is disposed more in the upper half region than the lower half region of the burnable poison-incorporated fuel rods. In addition, a burnable poison element such as gadolinium having a relatively large neutron absorbing cross section is disposed more in the lower half-region than the upper half region thereof. This can flatten the power distribution in the vertical direction of the fuel assembly and the power distribution in the horizontal direction at the final stage of the operation cycle. (I.N.)

  1. Nuclear reactor fuel sub-assemblies

    International Nuclear Information System (INIS)

    Dodd, J.A.

    1981-01-01

    An improved fuel sub-assembly for a liquid metal cooled fast breeder reactor, is described, in which fatigue damage due to buffeting by cross-current flows is reduced and protection is provided against damage by contact with other reactor structures during loading and unloading of the sub-assembly. (U.K.)

  2. Nuclear reactor fuel sub-assemblies

    International Nuclear Information System (INIS)

    Ford, J.; Bishop, J.F.W.

    1981-01-01

    An improved fuel sub-assembly for liquid metal cooled fast breeder nuclear reactors is described which facilitates dismantling operations for reprocessing purposes. The method of dismantling is described. (U.K.)

  3. Fuel assembly and reactor core

    International Nuclear Information System (INIS)

    Ishibashi, Yoko; Aoyama, Motoo; Oyama, Jun-ichi; Masumi, Ryoji; Soneda, Hideo.

    1994-01-01

    A fuel assembly comprises a plurality of fuel rods filled with nuclear fuels, a plurality of burnable poison rods incorporated with burnable poisons, and water rods which can vary the height in the tube depending on the coolant flow rate flown into the assembly. The amount of entire burnable poisons of the burnable poison-containing rods in adjacent with the water rods is smaller than the amount of entire burnable poisons in the burnable poison containing rods not in adjacent with the water rods. Then the average concentration of burnable poisons in the axial upper half region is made smaller than the average concentration of the burnable poisons at the axial lower half region. Further, a burnable poison concentration at the upper half region of at least one of burnable poison-containing rods in adjacent with the water rods is made lower than the burnable poison concentration in the lower half region. Since this can fasten the combustion of the burnable poisons, a fuel assembly having good fuel economy can be attained. (I.N.)

  4. Fuel assembly for nuclear reactor

    International Nuclear Information System (INIS)

    Yamanaka, Akihiro; Haikawa, Katsumasa; Haraguchi, Yuko; Nakamura, Mitsuya; Aoyama, Motoo; Koyama, Jun-ichi.

    1996-01-01

    In a BWR type fuel assembly comprising first fuel rods filled with nuclear fission products and second fuel rods filled with burnable poisons and nuclear fission products, the concentration of the burnable poisons mixed to a portion of the second fuel rods is controlled so that it is reduced at the upper portion and increased at the lower portion in the axial direction. In addition, a product of the difference of an average concentration of burnable poisons between the upper portion and the lower portion and the number of fuel rods is determined to higher than a first set value determined corresponding to the limit value of a maximum linear power density. The sum of the difference of the average concentration of the burnable poisons between the upper portion and the lower portion of the second fuel rod and the number of the second fuel rods is determined to lower than a second set value determined corresponding to a required value of a surplus reactivity. If the number of the fuel rods mixed with the burnable poisons is increased, the infinite multiplication factor at an initial stage of the burning is lowered and, if the concentration of the mixed burnable poisons is increased, the time of exhaustion of the burnable poisons is delayed. As a result, the maximum value of the infinite multiplication factor is suppressed thereby enabling to control surplus reactivity. (N.H.)

  5. Fuel assemblies for use in nuclear reactors

    International Nuclear Information System (INIS)

    Mochida, Takaaki.

    1987-01-01

    Purpose: To increase the plutonium utilization amount and improve the uranium-saving effect in the fuel assemblies of PWR type reactor using mixed uranium-plutonium oxides. Constitution: MOX fuel rods comprising mixed plutonium-uranium oxides are disposed to the outer circumference of a fuel assembly and uranium fuel rods only composed of uranium oxides are disposed to the central portion thereof. In such a fuel assembly, since the uranium fuel rods are present at the periphery of the control rod, the control rod worth is the same as that of the uranium fuel assembly in the prior art. Further, since about 25 % of the entire fuel rods is composed of the MOX fuel rods, the plutonium utilization amount is increased. Further, since the MOX fuel rods at low enrichment degree are present at the outer circumferential portion, mismatching at the boundary to the adjacent MOX fuel assembly is reduced and the problem of local power peaking increase in the MOX fuel assembly is neither present. (Kamimura, M.)

  6. Nuclear reactor fuel assembly spacer grids

    International Nuclear Information System (INIS)

    Jabsen, F.S.

    1977-01-01

    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.)

  7. Nuclear reactor seismic fuel assembly grid

    International Nuclear Information System (INIS)

    Anthony, A.J.

    1977-01-01

    The strength of a nuclear reactor fuel assembly is enhanced by increasing the crush strength of the zircaloy spacer grids which locate and support the fuel elements in the fuel assembly. Increased resistance to deformation as a result of laterally directed forces is achieved by increasing the section modulus of the perimeter strip through bending the upper and lower edges thereof inwardly. The perimeter strip is further rigidized by forming, in the central portion thereof, dimples which extend inwardly with respect to the fuel assembly. The integrity of the spacer grid may also be enhanced by providing back-up arches for some or all of the integral fuel element locating springs and the strength of the fuel assembly may be further enhanced by providing, intermediate its ends, a steel seismic grid. 13 claims, 6 figures

  8. Fuel assembly and reactor core

    International Nuclear Information System (INIS)

    Masumi, Ryoji; Aoyama, Motoo; Koyama, Jun-ichi; Ishibashi, Yoko; Mochida, Takaaki; Soneda, Hideo.

    1994-01-01

    In a fuel assembly having moderator rods, an axial average value of a ratio between the total of the lateral cross sectional area of a portion to be filled with moderators and the total of the lateral cross sectional area of fuel pellets is determined as greater than 0.4, a lateral cross sectional area of a portion to be filled with moderators per one moderator rod is determined as from 14 to 50cm 2 and the ratio between the total of the lateral cross sectional area of moderators and a total of the lateral cross sectional area of fuel pellets in a horizontal cross section is determined as from 2.7 to 3.4. Since the axial average value for lateral cross sectional area of a portion to be filled with moderators/lateral cross sectional area of fuel pellets is determined as ≥ 0.4, the lateral cross sectional area of moderators of moderator rods is increased, the lateral cross sectional area of a gap water region is decreased to reduce the value of local power peaking coefficient, so that thermal margin is ensured. At least one of the moderating rods is formed as a double-walled water rod tube to enhance an effect of spectral shift by flow rate control, reduce an uranium enrichment degree, and conduct operation without inserting control rods. (N.H.)

  9. Nuclear fuel assembly for fast neutron reactors

    International Nuclear Information System (INIS)

    Ilyunin, V.G.; Murogov, V.M.; Troyanov, M.F.; Rinejskij, A.A.; Ustinov, G.G.; Shmelev, A.N.

    1982-01-01

    The fuel assembly of a fast reactor consists of fuel elements comprising sections with fissionable and breeding material and tubes with hollows designed for entrapping gaseous fission products. Tubes joining up to the said sections are divided in a middle and a peripheral group such that at least one of the tube groups is placed in the space behind the coolant inlet ports. The configuration above allows reducing internal overpressure in the fuel assembly, thus reducing the volume of necessary structural elements in the core. (J.B.)

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

    International Nuclear Information System (INIS)

    Kobayashi, Kaoru.

    1998-01-01

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

  11. Equations of macrotransport in reactor fuel assemblies

    International Nuclear Information System (INIS)

    Sorokin, A.P.; Zhukov, A.V.; Kornienko, Yu.N.; Ushakov, P.A.

    1986-01-01

    The rigorous statement of equations of macrotransport is obtained. These equations are bases for channel-by-channel methods of thermohydraulic calculations of reactor fuel assemblies within the scope of the model of discontinuous multiphase coolant flow (including chemical reactions); they also describe a wide range of problems on thermo-physical reactor fuel assembly justification. It has been carried out by smoothing equations of mass, momentum and enthalpy transfer in cross section of each phase of the elementary fuel assembly subchannel. The equation for cross section flows is obtaind by smoothing the equation of momentum transfer on the interphase. Interaction of phases on the channel boundary is described using the Stanton number. The conclusion is performed using the generalized equation of substance transfer. The statement of channel-by-channel method without the scope of homogeneous flow model is given

  12. Fuel assembly for FBR type reactor

    International Nuclear Information System (INIS)

    Hayashi, Hideyuki.

    1995-01-01

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

  13. Fuel assemblies for BWR type reactors

    International Nuclear Information System (INIS)

    Ishizuka, Takao.

    1981-01-01

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

  14. Reconstitutable fuel assembly for a nuclear reactor

    International Nuclear Information System (INIS)

    Ferlan, S.J.; Kmonk, S.; Schallenberger, J.M.

    1982-01-01

    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

  15. Nuclear reactor fuel element assemblies

    International Nuclear Information System (INIS)

    Krawiec, D.M.; Bevilacqua, F.

    1974-01-01

    The fuel elements of each fuel element group are separated from each other by means of a multitude of thin, intersecting plates in the from of grid strips. Flow deflectors near the surface of the fuel elements are used in order to make the coolant flow more turbulent. They are designed as vanes and arranged at a distance on the grid strips. Each deflector vane has two arms stretching in opposite directions, each one into a neighbouring channel. In outward direction, the deflector vanes are converging. The strips with the vanes can be put on the supporting grid of the fuel elements. The vane structure can be reinforced by providing distortions in the strip material near the vanes. (DG) [de

  16. Fuel assembly for BWR type reactor

    International Nuclear Information System (INIS)

    Kato, Shigeru.

    1993-01-01

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

  17. Nuclear reactor fuel element assemblies

    International Nuclear Information System (INIS)

    Raven, L.F.

    1975-01-01

    A spacer grid for a nuclear fuel element comprises a plurality of cojointed cylindrical ferrules adapted to receive a nuclear fuel pin. Each ferrule has a pair of circumferentially spaced rigid stop members extending inside the ferrule and a spring locating member attached to the ferrule and also extending from the ferrule wall inwardly thereof at such a circumferential spacing relative to the rigid stop members that the line of action of the spring locating member passes in opposition to and between the rigid stop members which lie in the same diametric plane. At least some of the cylindrical ferrules have one rim shaped to promote turbulence in fluid flowing through the grid. (Official Gazette)

  18. Fuel assembly and nuclear reactor core

    International Nuclear Information System (INIS)

    Masumi, Ryoji; Aoyama, Motoo; Yamashita, Jun-ichi.

    1995-01-01

    The present invention concerns a fuel assembly and a nuclear reactor core capable of improving a transmutation rate of transuranium elements while improving a residual rate of fission products. In a reactor core of a BWR type reactor to which fuel rods with transuranium elements (TRU) enriched are loaded, the enrichment degree of transuranium elements occupying in fuel materials is determined not less than 2wt%, as well as a ratio of number of atoms between hydrogen and fuel heavy metals in an average reactor core under usual operation state (H/HM) is determined not more than 3 times. In addition, a ratio of the volumes between coolant regions and fuel material regions is determined not more than 2 times. A T ratio (TRU/Pu) is lowered as the TRU enrichment degree is higher and the H/HM ratio is lower. In order to reduce the T ratio not more than 1, the TRU enrichment degree is determined as not less than 2wt%, and the H/HM ratio is determined to not more than 3 times. Accordingly, since the H/HM ratio is reduced to not more than 1, and TRU is transmuted while recycling it with plutonium, the transmutation ratio of transuranium elements can be improved while improving the residual rate of fission products. (N.H.)

  19. MOX fuel assembly and reactor core

    International Nuclear Information System (INIS)

    Shimada, Hidemitsu; Koyama, Jun-ichi; Aoyama, Motoo

    1998-01-01

    The MOX fuel assembly of the present invention is of a c-lattice type loaded to a BWR type reactor. 74 MOX fuel rods filled with mixed oxides of uranium and plutonium and two water rods disposed to a space equal to that for 7 MOX fuel rods are arranged in 9 x 9 matrix. MOX fuel rods having the lowest enrichment degree are disposed to four corners of the 9 x 9 matrix. The enrichment degree means a ratio of the weight of fission products based on the total weight of fuels. Two MOX fuel rods having the same enrichment degree are arranged in each direction so as to be continuous from the MOX fuel rods at four corners in the direction of the same row and different column and same column and the different row. In addition, among the outermost circumferential portion of the 9 x 9 matrix, MOX fuel rods having a lower enrichment degree next to the MOX fuel rods having the lowest enrichment degree are arranged, each by three to a portion where MOX fuel rods having the lowest enrichment degree are not disposed. (I.N.)

  20. Paired replacement fuel assemblies for BWR-type reactor

    International Nuclear Information System (INIS)

    Oguchi, Kazushige.

    1997-01-01

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

  1. Reactor fuel element and fuel assembly

    International Nuclear Information System (INIS)

    Okada, Seiji; Ishida, Tsuyoshi; Ikeda, Atsuko.

    1997-01-01

    A mixture of fission products and burnable poisons is disposed at least to a portion between MOX pellets to form a burnable poison-incorporated fuel element without mixing burnable poisons to the MOX pellets. Alternatively, a mixture of materials other than the fission products and burnable poisons is formed into disks, a fuel lamination portion is divided into at least to two regions, and the ratio of number of the disks of the mixture relative to the volume of the region is increased toward the lower portion of the fuel lamination portion. With such a constitution, the axial power distribution of fuels can be made flat easily. Alternatively, the thickness of the disk of the mixture is increased toward the lower region of the fuel lamination portion to flatten the axial power distribution of the fuels in the same manner easily. The time and the cost required for the manufacture are reduced, and MOX fuels filled with burnable poisons with easy maintenance and control can be realized. (N.H.)

  2. Storage arrangement for nuclear reactor fuel assemblies

    International Nuclear Information System (INIS)

    Wade, E.E.

    1977-01-01

    Said invention is intended for providing an arrangement of spent fuel assembly storage inside which the space is efficiently used without accumulating a critical mass. The storage is provided for long fuel assemblies having along their longitudinal axis an active part containing the fuel and an inactive part empty of fuel. Said storage arrangement comprises a framework constituting some long-shaped cells designed so as each of them can receive a fuel assembly. Means of axial positioning of said assembly in a cell make it possible to support the fuel assemblies inside the framework according to a spacing ratio, along the cell axis, such as the active part of an assembly is adjacent to the inactive part of the adjacent assemblies [fr

  3. Nuclear reactor fuel element sub-assemblies

    International Nuclear Information System (INIS)

    Hill, G.D.; Trevalion, P.A.

    1977-01-01

    A fuel element sub-assembly for a liquid metal cooled fast reactor is described. It comprises a bundle of fuel pins enclosed by a tubular wrapper having a lower end journal for plugging into an upper aperture in a core supporting structure and a spike bar with an articulated bush for engaging a lower aperture in the core supporting structure. The articulated bush is retained on a spherical end portion of the spike bar by a pair of parallel retaining pins arranged transversely and disposed one each side of the spike bar. The pins are tubular and collapsible at a predetermined loading to enable the spherical end portion to pass between them. The articulated bush has an internal groove for engagement by a lifting grab, this groove being formed in a bore for receiving the spherical end portion of the spike bar. The construction lessens liability to rattling of the fuel element sub-assemblies and aids removal for replacement. (U.K.)

  4. VVANTAGE 6 - an advanced fuel assembly design for VVER reactors

    International Nuclear Information System (INIS)

    Doshi, P.K.; DeMario, E.E.; Knott, R.P.

    1993-01-01

    Over the last 25 years, Westinghouse fuel assemblies for pressurized water reactors (PWR's) have undergone significant changes to the current VANTAGE 5. VANTAGE 5 PWR fuel includes features such as removable top nozzles, debris filter bottom nozzles, low-pressure-drop zircaloy grids, zircaloy intermediate flow mixing grids, optimized fuel rods, in-fuel burnable absorbers, and increased burnup capability to region average values of 48000 MWD/MTU. These features have now been adopted to the VVER reactors. Westinghouse has completed conceptual designs for an advanced fuel assembly and other core components for VVER-1000 reactors known as VANTAGE 6. This report describes the VVANTAGE 6 fuel assembly design

  5. Grid structure for nuclear reactor fuel assembly

    International Nuclear Information System (INIS)

    Wachter, W.J.; Akey, J.G.

    1975-01-01

    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

  6. Fuel assembly for a nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ferrari, H M; Miller, D L; Tong, L S

    1973-09-06

    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.

  7. Fuel assembly transfer and storage system for nuclear reactors

    International Nuclear Information System (INIS)

    Allain, Albert; Thomas, Claude.

    1981-01-01

    Transfer and storage system on a site comprising several reactors and at least one building housing the installations common to all these reactors. The system includes: transfer and storage modules for the fuel assemblies comprising a containment capable of containing several assemblies carried on a transport vehicle, a set of tracks for the modules between the reactors and the common installations, handling facilities associated with each reactor for moving the irradiated assemblies from the reactor to a transfer module placed in loading position on a track serving the reactor and conversely to move the new assemblies from the transfer module to the reactor, and at least one handling facility located in the common installation building for loading the modules with new assemblies [fr

  8. About fuel assemblies optimization in research reactor

    International Nuclear Information System (INIS)

    Malers, Yu.P.

    1992-01-01

    Ealier was considered an algorithm for optimization of fuel assembly arrangement in a research reator. The alggorithm was based on an analytical relation between distributions of energy release and fuel concentration and on the method of succesive linearization and partially integral-number programming. In the paper are solved the problems, appeared as a result of realization of the used approach and required more correct formulation of the algorithm and introduction in it some variations

  9. A partial grid for a nuclear reactor fuel assembly

    International Nuclear Information System (INIS)

    Demario, E.E.

    1985-01-01

    The invention relates to a nuclear-reactor fuel assembly including fuel-rod supporting transverse grids. The fuel assembly includes at least one additional transverse grid which is disposed between two fuel-rod supporting grids and consists of at least one partial grid structure extending across only a portion of the fuel assembly and having fuel rods and control-rod guide thimbles of only said portion extending therethrough. The partial grid structure includes means for providing lateral support of the fuel rods and/or means for laterally deflecting coolant flow, and it is formed of inter-leaved inner straps and border straps, the interleaved inner straps preferably being of substantially smaller height than the border straps to reduce the amount of material capable of parasitically absorbing neutrons. The additional transverse grid may comprise several partial grid structures associated with different groups of fuel rods of the fuel assembly

  10. Nuclear reactor fuel assembly spacer grid

    International Nuclear Information System (INIS)

    Jabsen, F.S.

    1977-01-01

    A spacer grid for a nuclear fuel assembly is comprised 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, disposed 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. (Auth.)

  11. Removable fuel assembly for nuclear reactor

    International Nuclear Information System (INIS)

    Dubief, J.M.; Bonnamour, M.

    1984-01-01

    To facilitate the replacement of one or more fuel rods, taking into account the fact the operations are remote operations and under several meters of water, the following invention is presented. The fuel assembly is composed of a bundle of canned fuel pencils maintened on a structure which includes ends linked by spacer tubes. These tubes are fixed to one end in such a manner they are removable. For this, the plug of each tube has a plane stop surface on the end part and a conic coupling and guiding plug cooperating with a truncated bearing of the end part. Flat parts made on the cone allow to stop the tube rotating [fr

  12. Fuel assemblies for FBR type reactor

    International Nuclear Information System (INIS)

    Ikeda, Kiyoshi.

    1981-01-01

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

  13. Fuel assembly for a nuclear reactor

    International Nuclear Information System (INIS)

    Ferrari, H.M.; Miller, D.L.; Tong, L.S.

    1975-01-01

    A description is given of a fuel assembly including multiple open channel grids for holding fuel rods and control rod guide thimbles in predetermined fixed relationship with each other. Metallic straps are interwoven to form a grid or egg crate configuration having openings which receive the fuel rods and guide thimbles. To properly support and cool the fuel rods near the grid-fuel rod interface, springs and dimples on the grid straps project into each opening, the dimples being oriented in a direction to permit flow of coolant upwardly therethrough. To minimize turbulence in coolant flow, the leading edge of each grid strap is provided with cutout sections which form scallops effective in channeling coolant in a uniform flow path through the network of grid openings

  14. Support a nuclear fuel assembly in a reactor

    International Nuclear Information System (INIS)

    Leclercq, J.

    1985-01-01

    The device has to maintain the assemblies with regard to a horizontal plate of the core. The assemblies, having the same section, are arranged side by side in a regular polygonal lattice and each asssembly is, either equipped with at least two zones to receive the rods which are vertically inserted and maintained during the reactor operation, or beside an assembly which is equipped. The device has two sets comprising each one at least one deformable locking element and a rigid element which raches with it, one fixed to the fuel assembly and the other fixed to a horizontal plate attached to the reactor core, positioned so that inserting a fuel rod into an emplacement in the fuel assembly deforms the bolt transversally to lock it with the rigid piece. The invention can be applied to water moderated reactors [fr

  15. Fuel assembly for BWR type reactor

    International Nuclear Information System (INIS)

    Ueda, Makoto

    1990-01-01

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

  16. Nuclear reactor, fuel assembly and neutron measuring system

    International Nuclear Information System (INIS)

    Chaki, Masao; Murase, Michio; Zukeran, Atsushi; Moriya, Kimiaki

    1998-01-01

    The present invention provides a BWR type reactor improved with the efficiency of used fuels and fuel economy by increasing a rated power and reducing exchange fuels. Namely, in a BWR type reactor at present, a thermal limit value is determined by conducting nuclear calculation of the reactor core based on data of reactor flow rate measurement and data of neutron flux measurement. However, since the neutron calculation of the reactor core is based on fuel assemblies while the points for the neutron measurement are present at the outside of the fuel assemblies, errors are caused. A margin including the errors has been used as a thermal limit value during operation. In the present invention, neutron fluxes in the fuel assembly as a base of the nuclear calculation can be measured by the same number of neutron detector tubes, but the number of the measuring points is increased to four times. With such procedures, errors caused by the difference of the neutron calculation and values at neutron measuring points can be reduced. As a result, a margin of the thermal limit value is reduced to increase the degree of freedom of reactor operation. Then, the economical property of the reactor operation can be improved. (N.H.)

  17. Fuel assembly for pressure loss variable PWR type reactor

    International Nuclear Information System (INIS)

    Yoshikuni, Masaaki.

    1993-01-01

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

  18. Simulation model of dynamical behaviour of reactor fuel assemblies

    International Nuclear Information System (INIS)

    Planchard, J.

    1994-01-01

    This report briefly describes the homogenized dynamical equations of a tube bundle placed in a perfect irrotational fluid, on case of small displacements. This approach can be used to study the mechanical behaviour of fuel assemblies of PWR reactor submitted to earthquake or depressurization blow-down. The numerical calculations require to define the added mass matrix of the fuel assemblies, for which the principle of computation is presented. (author). 14 refs., 4 figs

  19. Spacer grid for a nuclear reactor fuel assembly

    International Nuclear Information System (INIS)

    Jabsen, F.S.

    1978-01-01

    The spacer grid consists of pairs of plates forming rectangular cells and enclosing the cylindrical fuel assemblies. They have got rigid as well as elastic projections extending into the cells and holding the fuel assemblies. Additional pairs of plates are arranged in about the center of the grid of plates. They have got only elastic projections extending on both sides of the plates into one cell each. This spacer grid may be used for reactor cores with and without fuel channels. By the combination of spring-elastic and rigid projections there is obtained a reinforced outer tie. Hydraulic pressure losses, parasitic neutron capture, and hot spots are essentially reduced. (DG) [de

  20. Drum of storing fuel assemblies of nuclear reactor

    International Nuclear Information System (INIS)

    Artemiev, L.N.; Batjukov, V.I.; Fadeev, A.I.

    1979-01-01

    The proposed drum for storing fuel assemblies of a nuclear reactor comprises a holder rotatable around its axis and provided with tubular sockets arranged in concentric rows along the circumference of the holder so that the axis of at least one socket of each row intersects the trajectory described by the grip of the recharging mechanism in the course of its movement. The proposed drum design makes it possible to facilitate and speed up the process of recharging fuel asemblies

  1. Device for transferring fast nuclear reactor fuel assemblies

    International Nuclear Information System (INIS)

    Plagnard, Andre.

    1982-01-01

    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 [fr

  2. Hydraulic shock damper for fuel assemblies of nuclear reactors

    International Nuclear Information System (INIS)

    Jabson, F.S.

    1978-01-01

    A typical embodiment of this invention provides a hydraulic mechanism for alleviating the effect of seismic forces and other stresses that are applied to a fuel assembly in a nuclear reactor. Illustratively, hollow guide posts potrude into a fuel assembly end fitting grid from biased spring pads. Plungers that move with the spring pads plug one end of each of the respective guide posts. Plates on the end fitting grid that have individual holes for fluid discharge partially plug the other ends of the respective guide posts, thereby providing a hydraulic means for absorbing the longitudinal component of seismic shocks and other anticipated forces. (Auth.)

  3. Light water reactors fuel assembly mechanical design and evaluation

    International Nuclear Information System (INIS)

    Anon.

    1981-01-01

    This standard establishes a procedure for performing an evaluation of the mechanical design of fuel assemblies for light water-cooled commercial power reactors. It does not address the various aspects of neutronic or thermalhydraulic performance except where these factors impose loads or constraints on the mechanical design of the fuel assemblies. This standard also includes a set of specific requirements for design, various potential performance problems and criteria aimed specifically at averting them. This standard replaces ANSI/ANS-57.5-1978

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

    International Nuclear Information System (INIS)

    Rogers, J.W.

    1988-01-01

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

  5. Fuel assembly for use in BWR type reactor

    International Nuclear Information System (INIS)

    Inaba, Yuzo.

    1988-01-01

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

  6. Comparison of fuel assemblies in lead cooled fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Perez, A.; Sanchez, H.; Aguilar, L.; Espinosa P, G., E-mail: alejandria.peval@gmail.com [Universidad Autonoma Metropolitana, Unidad Iztapalapa, San Rafael Atlixco No. 186, Col. Vicentina, 09340 Ciudad de Mexico (Mexico)

    2016-09-15

    This paper presents a comparison of the thermal-fluid processes in the core, fuel heat transfer, and thermal power between two fuel assemblies: square and hexagonal, in a lead-cooled fast reactor (Lfr). A multi-physics reduced order model for the analysis of Lfr single channel is developed in this work. The work focused on a coupling between process of neutron kinetic, fuel heat transfer process and thermal-fluid, in a single channel. The thermal power is obtained from neutron point kinetics model, considering a non-uniform power distribution. The analysis of the processes of thermal-fluid considers thermal expansion effects. The transient heat transfer in fuel is carried out in an annular geometry, and one-dimensional in radial direction for each axial node. The results presented in comparing these assemblies consider the temperature field in the fuel, in the thermal fluid and under steady state, and transient conditions. Transients consider flow of coolant and inlet temperature of coolant. The mathematical model of Lfr considers three main modules: the heat transfer in the annular fuel, the power generation with feedback effects on neutronic, and the thermal-fluid in the single channel. The modeling of nuclear reactors in general, the coupling is crucial by the feedback between the neutron processes with fuel heat transfer, and thermo-fluid, where is very common the numerical instabilities, after all it has to refine the model to achieve the design data. In this work is considered as a reference the ELSY reactor for the heat transfer analysis in the fuel and pure lead properties for analyzing the thermal-fluid. The results found shows that the hexagonal array has highest temperature in the fuel, respect to square array. (Author)

  7. Comparison of fuel assemblies in lead cooled fast reactors

    International Nuclear Information System (INIS)

    Perez, A.; Sanchez, H.; Aguilar, L.; Espinosa P, G.

    2016-09-01

    This paper presents a comparison of the thermal-fluid processes in the core, fuel heat transfer, and thermal power between two fuel assemblies: square and hexagonal, in a lead-cooled fast reactor (Lfr). A multi-physics reduced order model for the analysis of Lfr single channel is developed in this work. The work focused on a coupling between process of neutron kinetic, fuel heat transfer process and thermal-fluid, in a single channel. The thermal power is obtained from neutron point kinetics model, considering a non-uniform power distribution. The analysis of the processes of thermal-fluid considers thermal expansion effects. The transient heat transfer in fuel is carried out in an annular geometry, and one-dimensional in radial direction for each axial node. The results presented in comparing these assemblies consider the temperature field in the fuel, in the thermal fluid and under steady state, and transient conditions. Transients consider flow of coolant and inlet temperature of coolant. The mathematical model of Lfr considers three main modules: the heat transfer in the annular fuel, the power generation with feedback effects on neutronic, and the thermal-fluid in the single channel. The modeling of nuclear reactors in general, the coupling is crucial by the feedback between the neutron processes with fuel heat transfer, and thermo-fluid, where is very common the numerical instabilities, after all it has to refine the model to achieve the design data. In this work is considered as a reference the ELSY reactor for the heat transfer analysis in the fuel and pure lead properties for analyzing the thermal-fluid. The results found shows that the hexagonal array has highest temperature in the fuel, respect to square array. (Author)

  8. Fuel assemblies

    International Nuclear Information System (INIS)

    Nakatsuka, Masafumi.

    1979-01-01

    Purpose: To prevent scattering of gaseous fission products released from fuel assemblies stored in an fbr type reactor. Constitution; A cap provided with means capable of storing gas is adapted to amount to the assembly handling head, for example, by way of threading in a storage rack of spent fuel assemblies consisting of a bottom plate, a top plate and an assembly support mechanism. By previously eliminating the gas inside of the assembly and the cap in the storage rack, gaseous fission products upon loading, if released from fuel rods during storage, are stored in the cap and do not scatter in the storage rack. (Horiuchi, T.)

  9. Fuel assembly for gas-cooled nuclear reactors

    International Nuclear Information System (INIS)

    Yellowlees, J.M.

    1976-01-01

    A fuel assembly is described for gas-cooled nuclear reactor which consists of a wrapper tube within which are positioned a number of spaced apart beds in a stack, with each bed containing spherical coated particles of fuel; each of the beds has a perforated top and bottom plate; gaseous coolant passes successively through each of the beds; through each of the beds also passes a bypass tube; part of the gas travels through the bed and part passes through the bypass tube; the gas coolant which passes through both the bed and the bypass tube mixes in the space on the outlet side of the bed before entering the next bed

  10. Handling apparatus for a nuclear reactor fuel assembly

    International Nuclear Information System (INIS)

    Shallenberger, J.M.; Hornak, L.P.; Desmarchais, W.E.

    1978-01-01

    An apparatus is disclosed for handling radioactive fuel assembly during transfer operations. The radioactive fuel assembly is drawn up into a shielding sleeve which substantially reduces the level of radioactivity immediately surrounding the sleeve thereby permitting direct access by operating personnel. The lifting assembly which draws the fuel assembly up within the shielding sleeve is mounted to and forms an integral part of the handling apparatus. The shielding sleeve accompanies the fuel assembly during all of the transfer operations

  11. Gripping means for fuel assemblies of nuclear reactor

    International Nuclear Information System (INIS)

    Batjukov, V.I.; Fadeev, A.I.; Shkhian, T.G.; Vjugov, O.N.

    1980-01-01

    The proposed gripping means for fuel assemblies of a nuclear reactor comprises a housing, whereupon there is movably mounted a slider provided with longitudinally extending slots to receive gripping jaws whose tails are pivotably secured to the housing of the gripping means. On one side, the end faces of the longitudinally extending slots are slanted with respect to the longitudinal axis of the gripping means and come in contact with the teeth of the gripping jaws provided on the end which is opposite to the tail, whereby the jaws open as the slider and housing of the gripping means moves relative to each other so that the teeth are received in an internal groove provided in the head of the fuel assembly

  12. Physical characteristics of GE [General Electric] BWR [boiling-water reactor] fuel assemblies

    International Nuclear Information System (INIS)

    Moore, R.S.; Notz, K.J.

    1989-06-01

    The physical characteristics of fuel assemblies manufactured by the General Electric Company for boiling-water reactors are classified and described. The classification into assembly types is based on the GE reactor product line, the Characteristics Data Base (CDB) assembly class, and the GE fuel design. Thirty production assembly types are identified. Detailed physical data are presented for each assembly type in an appendix. Descriptions of special (nonstandard) fuels are also reported. 52 refs., 1 fig., 6 tabs

  13. Fuel assembly

    International Nuclear Information System (INIS)

    Ueda, Makoto; Ogiya, Shunsuke.

    1989-01-01

    For improving the economy of a BWR type reactor by making the operation cycle longer, the fuel enrichment degree has to be increased further. However, this makes the subcriticality shallower in the upper portion of the reactor core, to bring about a possibility that the reactor shutdown becomes impossible. In the present invention, a portion of fuel rod is constituted as partial length fuel rods (P-fuel rods) in which the entire stack length in the effective portion is made shorter by reducing the concentration of fissionable materials in the axial portion. A plurality of moderator rods are disposed at least on one diagonal line of a fuel assembly and P-fuel rods are arranged at a position put between the moderator rods. This makes it possible to reactor shutdown and makes the axial power distribution satisfactory even if the fuel enrichment degree is increased. (T.M.)

  14. Simulating the temperature noise in fast reactor fuel assemblies

    International Nuclear Information System (INIS)

    Kebadze, B.V.; Pykhtina, T.V.; Tarasko, M.Z.

    1987-01-01

    Characteristics of temperature noise at various modes of coolant flow in fast reactor fuel assemblies (FA) and for different points of sensor installation are investigated. Stationary mode of coolant flow and mode with a partial overlapping of FA through cross section, resulting in local temperature increase and sodium boiling, are considered. Numerical simulation permits to evaluate time characteristicsof temperature noise and to formulate requirements for dynamic characteristics of the sensors, and also to clarify the dependence of coolant distribution parameters on the sensor location and peculiarities of stationary temperature profile

  15. Apparatus for securing structural tubes in nuclear reactor fuel assemblies

    International Nuclear Information System (INIS)

    Kerry, J.S.

    1987-01-01

    This patent describes a nuclear reactor fuel assembly having a structural tube with a predetermined inside diameter, a generally cylindrical insert of an axial length substantially smaller than the axial length of the structural tube and having a generally cylindrical passageway of a predetermined diameter smaller than the predetermined inside diameter for providing an effectively reduced inside diameter for the structural tube. The insert comprises: means, having an outside diameter approximately equal to the predetermined inside diameter, for coaxially centering the insert within the structural tube; forming lobes, operable when expanded to locally deform against the structural tube thereby locking the insert within the structural tube

  16. Fuel assemblies for use in BWR type reactors

    International Nuclear Information System (INIS)

    Hirukawa, Koji.

    1987-01-01

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

  17. Spacing grid intended for nuclear reactor fuel assemblies

    International Nuclear Information System (INIS)

    Patterson, J.F.; Flora, B.S.

    1977-01-01

    This invention concerns a new improved type of spacing grid that can be used in nuclear reactor fuel assemblies. Under the invention a spacing grid is provided, preferably of the bimetallic type. This grid includes a set of flexible inconel strips positioned by structural 'zircalloy' fittings, having relatively low neutron absorption characteristics in comparison with systems where the flexible strips are welded in position, or where the spring forms an integral part of the structure. The openings for the fuel elements which are defined by the structural fittings intercrossing are fitted internally with bosses which work in conjunction with a spring directed downwards as from the flexible strip so as to position the individual fuel rods in their respective openings inside the grid structure. These flexible strips are arranged in rows extending in directions which depend on the particular design of the fuel asembly and which contain flexible components so distributed that the loads of the individual springs tend to equalize each other mutually. The reaction load exerting itself on the supporting structure is reduced to the minimum, and this results in a lesser distortion in the reactor and an equalisation of the spring loads [fr

  18. Burnable poison rod for a nuclear reactor fuel assembly

    International Nuclear Information System (INIS)

    Funk, C.E.; Oneufer, A.S.

    1984-01-01

    A burnable poison rod for use in a nuclear reactor fuel assembly which includes concentrically disposed rods having an annular space therebetween which extends the full length of the rods. The inner rod is hollow to permit circulation of coolant therethrough. Annular burnable poison pellets are positioned in the annular space which is closed at both ends by plugs. A spring clip is located in the plenum space above the pellet stack in the rods. The spring clip is of cylindrical configuration having a gap in the material which provides two ends adapted to be squeezed toward each other. A cross section of the clip shows that its ends contain alternating flat and round edges, the round edges conforming to the outer rod inner surface to provide a retentive force which is releasably applied to the pellet stack as it grows during operation in a reactor

  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. Fuel assembly

    International Nuclear Information System (INIS)

    Yokota, Tokunobu.

    1990-01-01

    A fuel assembly used in a FBR type nuclear reactor comprises a plurality of fuel rods and a moderator guide member (water rod). A moderator exit opening/closing mechanism is formed at the upper portion of the moderator guide member for opening and closing a moderator exit. In the initial fuel charging operation cycle to the reactor, the moderator exit is closed by the moderator exit opening/closing mechanism. Then, voids are accumulated at the inner upper portion of the moderator guide member to harden spectrum and a great amount of plutonium is generated and accumulated in the fuel assembly. Further, in the fuel re-charging operation cycle, the moderator guide member is used having the moderator exit opened. In this case, voids are discharged from the moderator guide member to decrease the ratio, and the plutonium accumulated in the initial charging operation cycle is burnt. In this way, the fuel economy can be improved. (I.N.)

  1. Tests of the RBMK-1500 reactor fuel assemblies in the Leningrad reactor

    International Nuclear Information System (INIS)

    Aden, V.C.; Varovin, I.A.; Vorontsov, B.A.

    1981-01-01

    Test of fuel assemblies of the RBMK-1500 reactor is conducted in the reactor of the Leningrad NPP unit 2 for proving the calculational values of critical power of the RBMK-1500 reactor fuel assemblies adopted in design. The experiment presupposes the maximal approximation of the fuel assembly operation parameters to the calculational critical parameters without bringing into the mode of heat transfer crisis. The experiments are carried out at 500, 850 and 900 MW(el) of the reactor. The maximal channel power made up 472 kW at 20.5 t/h coolant flow rate and 49% mass steam content at the outlet of the channel. It was concluded that there was supply up to the heat transfer crisis in all the investigated modes. Data of temperature measurings of the fuel element cans, readings of the devices of the failure control system of the fuel element cans and external inspection of the assemblies after the tests testify to it [ru

  2. Fuel assembly

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  3. Fuel assembly

    International Nuclear Information System (INIS)

    Gjertsen, R.K.; Bassler, E.A.; Huckestein, E.A.; Salton, R.B.; Tower, S.N.

    1988-01-01

    A fuel assembly adapted for use with a pressurized water nuclear reactor having capabilities for fluid moderator spectral shift control is described comprising: parallel arranged elongated nuclear fuel elements; means for providing for axial support of the fuel elements and for arranging the fuel elements in a spaced array; thimbles interspersed among the fuel elements adapted for insertion of a rod control cluster therewithin; means for structurally joining the fuel elements and the guide thimbles; fluid moderator control means for providing a volume of low neutron absorbing fluid within the fuel assembly and for removing a substantially equivalent volume of reactor coolant water therefrom, a first flow manifold at one end of the fuel assembly sealingly connected to a first end of the moderator control tubes whereby the first ends are commonly flow connected; and a second flow manifold, having an inlet passage and an outlet passage therein, sealingly connected to a second end of the moderator control tubes at a second end of the fuel assembly

  4. Damage of fuel assembly premature changing in a power reactor

    International Nuclear Information System (INIS)

    Rudik, A.P.

    1987-01-01

    Material balance, including energy recovery and nuclear fuel flow rate, under conditions of premature FA extraction from power reactor is considered. It is shown that in cases when before and after FA extraction reactor operates not under optimal conditions damage of FA premature changing is proportional to the first degree of fuel incomplete burning. If normal operating conditions of reactor or its operation after FA changing is optimal, the damage is proportional to the square of fuel incomplete burning

  5. Improvements in or relating to gripping means for handling nuclear reactor fuel assemblies

    International Nuclear Information System (INIS)

    Batjukov, V.I.; Vjugov, O.N.; Fadeev, A.I.; Shkhian, T.G.

    1980-01-01

    A gripping means for handling fuel assemblies, the heads of which are internally recessed to receive gripping jaws, forms part of a reactor refuelling machine and is telescopically accommodated within a manipulator tube of the machine. A through hole is provided to allow cooling medium to be passed through the fuel assemblies to remove afterheat when the gripping means is used to transfer assemblies from a reactor core to spent fuel storage sockets. (author)

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

    International Nuclear Information System (INIS)

    1977-01-01

    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. Fuel assembly

    International Nuclear Information System (INIS)

    Kawai, Mitsuo.

    1988-01-01

    Purpose: To reduce the corrosion rate and suppress the increase of radioactive corrosion products in reactor water of nuclear fuel assemblies for use in BWR type reactors having spacer springs made of nickel based deposition reinforced type alloys. Constitution: Spacer rings made of nickel based deposition reinforced type alloy are incorporated and used as fuel assemblies after applying treatment of dipping and maintaining at high temperature water followed by heating in steams. Since this can remove the nickel leaching into reactor water at the initial stage, Co-58 as the radioactive corrosion products in the reactor water can be reduced, and the operation at in-service inspection or repairement can be facilitated to improve the working efficiency of the nuclear power plant. The dipping time is desirably more than 10 hours and more desirably more than 30 hours. (Horiuchi, T. )

  8. TEMP-M program for thermal-hydraulic calculation of fast reactor fuel assemblies

    International Nuclear Information System (INIS)

    Bogoslovskaya, C.P.; Sorokin, A.P.; Tikhomirov, B.B.; Titov, P.A.; Ushakov, P.A.

    1983-01-01

    TEMP-M program (Fortran, BESM-6 computer) for thermal-hydraulic calculation of fast reactor fuel assemblies is described. Results of calculation of temperature field in a 127 fuel element assembly of BN-600, reactor accomplished according to TEMP-N program are considered as an example. Algorithm, realized in the program, enables to calculate the distributions of coolant heating, fuel element temperature (over perimeter and length) and assembly shell temperature. The distribution of coolant heating in assembly channels is determined from a solution of the balance equation system which accounts for interchannel exchange, nonadiabatic conditions on the assembly shell. The TEMP-M program gives necessary information for calculation of strength, seviceability of fast reactor core elements, serves an effective instrument for calculations when projecting reactor cores and analyzing thermal-hydraulic characteristics of operating reactor fuel assemblies

  9. Grapples for manipulating end fittings for nuclear reactor fuel assemblies

    International Nuclear Information System (INIS)

    Jabsen, F.S.

    1982-01-01

    A nuclear fuel assembly includes control rod guide tubes the upper ends of which protrude beyond a spider and are locked in place by means of a cellular lattice seated in grooves in the outer surfaces of the sleeves. A grapple is provided for disengaging the structure comprising lattice, spider, springs and a grill from the end of the fuel assembly to enable these components to be removed in an assembly state and subsequently replaced after inspection and repair. (author)

  10. Transfer hood for handling fuel assemblies in nuclear reactors and especially fast reactors

    International Nuclear Information System (INIS)

    Aubert, M.; Merland, D.; Renaux, C.

    1975-01-01

    A description is given of a hood for transferring fuel assemblies between two or more separate guide ramps inclined to the vertical at the same angle of slope and extending respectively through a first passage into the reactor vessel and through a second passage into a fuel-assembly storage chamber. The hood comprises at least one shielded duct joined to a flanged rotating portion which is so arranged that the open lower end of the shielded duct is positioned in the line of extension of one guide ramp and then the other as a result of displacement of the rotating portion

  11. Fuel assembly

    International Nuclear Information System (INIS)

    Abe, Hideaki; Sakai, Takao; Ishida, Tomio; Yokota, Norikatsu.

    1992-01-01

    The lower ends of a plurality of plate-like shape memory alloys are secured at the periphery of the upper inside of the handling head of a fuel assembly. As the shape memory alloy, a Cu-Zn alloy, a Ti-Pd alloy or a Fe-Ni alloy is used. When high temperature coolants flow out to the handling head, the shape memory alloy deforms by warping to the outer side more greatly toward the upper portion thereof with the temperature increase of the coolants. As the result, the shape of the flow channel of the coolants is changed so as to enlarge at the exit of the upper end of the fuel assembly. Then, the pressure loss of the coolants in the fuel assembly is decreased by the enlargement. Accordingly, the flow rate of the coolants in the fuel assembly is increased to lower the temperature of the coolants. Further, high temperature coolants and low temperature coolants are mixed sufficiently just above the fuel assembly. This can suppress the temperature fluctuation of the mixed coolants in the upper portion of the reactor core, thereby enabling to decrease a fatigue and failures of the structural components in the upper portion of the reactor core. (I.N.)

  12. Inspection and repair apparatus for a nuclear reactor fuel assembly

    International Nuclear Information System (INIS)

    Shallenberger, J.M.; Hornak, L.P.; Desmarchais, W.E.

    1975-01-01

    An apparatus is disclosed for inspecting and repairing a radioactive fuel assembly. The radioactive fuel assembly is positioned within a shielding sleeve which substantially reduces the level of radioactivity immediately surrounding the sleeve thereby permitting direct access by operating personnel. In one embodiment, a rotatable collar is mounted to the sleeve at a midlength location. An access port, an inspection port and an instrument port are included with the collar so that operating personnel may directly inspect the fuel assembly and effectuate any necessary repairs

  13. MOCA, Criticality of VVER Reactor Hexagonal Fuel Assemblies

    International Nuclear Information System (INIS)

    KYNCL, Jan

    1994-01-01

    1 - Description of program or function: Criticality problem in neutron transport for hexagonal fuel assembly in VVER nuclear reactor. The assembly is assumed to be either arranged in an infinite hexagonal array or placed in vacuum. The problem is solved in three- dimensional geometry, using standard energy group formalism and assuming that effective scattering cross sections are presented as Legendre polynomial expansions. The code evaluates ten different physical quantities, e.g. multiplication factor, neutron flux per energy group and spatial zone, integrated over angle and power in any zone of the assembly. 2 - Method of solution: Monte Carlo method of successive generations is applied. Computation proceeds according to an analog random process. The code is organized into three blocks: In the first block, the input data are converted to quantities for use in the Monte Carlo calculation. An initial neutron distribution is calculated, which corresponds to a fission spectrum uniform in spatial and angular variables. The main calculations are carried out in the second block (subroutine PROC2). This block is subdivided into geometrical and physical parts. Neutron tracks in individual zones and groups as well as probabilities for the formation of secondary neutrons are calculated. In the third block (subroutine PROC3), the results are evaluated statistically. Effective multiplication coefficients, the neutron flux per group and zone, and respective errors are computed. These quantities serve as a basis for the evaluation of other quantities. The results are either printed or stored for future evaluations. 3 - Restrictions on the complexity of the problem: In the PC version of the program, the maximum number of neutrons is 1000, the maximum number of energy groups is 4, and the maximum number of material compositions is 15. Angular expansion of scattering cross sections is allowed up to P10. These restrictions can easily be removed by increasing input parameters and

  14. Nuclear reactor fuel assemblies and end fitting grid structures therefor

    International Nuclear Information System (INIS)

    Jabsen, F.S.

    1978-01-01

    An improved end fitting grid structure is described for nuclear fuel assemblies which overcomes the need for load-bearing control rod guide tubes and the expensive special fittings that these tubes required. (UK)

  15. Fuel assembly for light-water cooled nuclear reactors

    International Nuclear Information System (INIS)

    Leroux, J.C.; Burfin, P.

    1995-01-01

    In order to make easier the replacement of damaged fuel rods, a fuel assembly has been designed with a cluster of parallel fuel rods maintained in guide tubes with braces and sockets fixed on each tube ends; at least one of the fixing sockets of each tube is dismountable as well as an adapter plate on the socket, in order to lock or un-lock the guide tubes from the sockets. 11 fig

  16. Fuel assembly

    International Nuclear Information System (INIS)

    Bando, Masaru.

    1993-01-01

    As neutron irradiation progresses on a fuel assembly of an FBR type reactor, a strong force is exerted to cause ruptures if the arrangement of fuel elements is not displaced, whereas the fuel elements may be brought into direct contact with each other not by way of spacers to cause burning damages if the arrangement is displaced. In the present invention, the circumference of fuel elements arranged in a normal triangle lattice is surrounded by a wrapper tube having a hexagonal cross section, wire spacers are wound therearound, and deformable spacers are distributed to optional positions for fuel elements in the wrapper tube. Interaction between the fuel elements caused by irradiation is effectively absorbed, thereby enabling to delay the occurrence of the rupture and burning damages of the elements. (N.H.)

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

    International Nuclear Information System (INIS)

    Delafosse, Jacques.

    1976-01-01

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

  18. Fuel assembly

    International Nuclear Information System (INIS)

    Kurihara, Kunitoshi; Azekura, Kazuo.

    1992-01-01

    In a reactor core of a heavy water moderated light water cooled pressure tube type reactor, no sufficient effects have been obtained for the transfer width to a negative side of void reactivity change in a region of a great void coefficient. Then, a moderation region divided into upper and lower two regions is disposed at the central portion of a fuel assembly. Coolants flown into the lower region can be discharged to the cooling region from an opening disposed at the upper end portion of the lower region. Light water flows from the lower region of the moderator region to the cooling region of the reactor core upper portion, to lower the void coefficient. As a result, the reactivity performance at low void coefficient, i.e., a void reaction rate is transferred to the negative side. Thus, this flattens the power distribution in the fuel assembly, increases the thermal margin and enables rapid operaiton and control of the reactor core, as well as contributes to the increase of fuel burnup ratio and reduction of the fuel cycle cost. (N.H.)

  19. Preliminary neutronics calculation of fusion-fission hybrid reactor breeding spent fuel assembly

    International Nuclear Information System (INIS)

    Ma Xubo; Chen Yixue; Gao Bin

    2013-01-01

    The possibility of using the fusion-fission hybrid reactor breeding spent fuel in PWR was preliminarily studied in this paper. According to the fusion-fission hybrid reactor breeding spent fuel characteristics, PWR assembly including fusion-fission hybrid reactor breeding spent fuel was designed. The parameters such as fuel temperature coefficient, moderator temperature coefficient and their variation were investigated. Results show that the neutron properties of uranium-based assembly and hybrid reactor breeding spent fuel assembly are similar. The design of this paper has a smaller uniformity coefficient of power at the same fissile isotope mass percentage. The results will provide technical support for the future fusion-fission hybrid reactor and PWR combined with cycle system. (authors)

  20. Subchannel analysis of sodium-cooled reactor fuel assemblies with annular fuel pins

    International Nuclear Information System (INIS)

    Memmott, Matthew; Buongiorno, Jacopo; Hejzlar, Pavel

    2009-01-01

    Using a RELAP5-3D subchannel analysis model, the thermal-hydraulic behavior of sodium-cooled fuel assemblies with internally and externally cooled annular fuel rods was investigated, in an effort to enhance the economic performance of sodium-fast reactors by increasing the core power density, decreasing the core pressure drop, and extending the fuel discharge burnup. Both metal and oxide fuels at high and low conversion ratios (CR=0.25 and CR=1.00) were investigated. The externally and internally cooled annular fuel design is most beneficial when applied to the low CR core, as clad temperatures are reduced by up to 62.3degC for the oxide fuel, and up to 18.5degC for the metal fuel. This could result in a power uprates of up to ∼44% for the oxide fuel, and up to ∼43% for the metal fuel. The use of duct ribs was explored to flatten the temperature distribution at the core outlet. Subchannel analyses revealed that no fuel melting would occur in the case of complete blockage of the hot interior-annular channel for both metal and oxide fuels. Also, clad damage would not occur for the metal fuel if the power uprate is 38% or less, but would indeed occur for the oxide fuel. (author)

  1. Improvements in and relating to drums for storing fuel assemblies of nuclear reactors

    International Nuclear Information System (INIS)

    Artemiev, L.N.; Batjukov, V.I.; Fadeev, A.I.

    1977-01-01

    A specification is given for an improved design of drums used for storing nuclear reactor fuel assemblies. It is claimed that this new design reduces wear, increases reliability and enables the recharging operation to be speeded up. (UK)

  2. Device for handling fuel assemblies within a nuclear reactor core

    International Nuclear Information System (INIS)

    Dupuy, G.

    1975-01-01

    A device is described which comprises two arms having synchronized movements, the tubular portions of the two arms being rigidly fixed to each other by means of a sliding connection and capable of being endowed by means of a differential drive system with movements for producing the same effort but applied in the opposite direction in order that the lateral reaction on the grab of the first arm on a fuel assembly should be equal and opposite to the lateral reaction of the cylindrical guide tube of the second arm on the same fuel assembly

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  4. 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.

  5. Fuel assembly

    International Nuclear Information System (INIS)

    Nakajima, Akiyoshi; Bessho, Yasunori; Aoyama, Motoo; Koyama, Jun-ichi; Hirakawa, Hiromasa; Yamashita, Jun-ichi; Hayashi, Tatsuo

    1998-01-01

    In a fuel assembly of a BWR type reactor in which a water rod of a large diameter is disposed at the central portion, the cross sectional area perpendicular to the axial direction comprises a region a of a fuel rod group facing to a wide gap water region to which a control rod is inserted, a region b of a fuel rod group disposed on the side of the wide gap water region other than the region a, a region d of a fuel rod group facing to a narrow gap water region and a region c of a fuel rod group disposed on the side of the narrow gap water region other than the region d. When comparing an amount of fission products contained in the four regions relative to that in the entire regions and average enrichment degrees of fuel rods for the four regions, the relative amount and the average enrichment degree of the fuel rod group of the region a is minimized, and the relative amount and the average enrichment degree of the fuel rod group in the region b is maximized. Then, reactor shut down margin during cold operation can be improved while flattening the power in the cross section perpendicular to the axial direction. (N.H.)

  6. Fuel assembly

    International Nuclear Information System (INIS)

    Ueda, Sei; Ando, Ryohei; Mitsutake, Toru.

    1995-01-01

    The present invention concerns a fuel assembly suitable to a BWR-type reactor and improved especially with the nuclear characteristic, heat performance, hydraulic performance, dismantling or assembling performance and economical property. A part of poison rods are formed as a large-diameter/multi-region poison rods having a larger diameter than a fuel rod. A large number of fuel rods are disposed surrounding a large diameter water rod and a group of the large-diameter/multi-region poison rods in adjacent with the water rod. The large-diameter water rod has a burnable poison at the tube wall portion. At least a portion of the large-diameter poison rods has a coolant circulation portion allowing coolants to circulate therethrough. Since the large-diameter poison rods are disposed at a position of high neutron fluxes, a large neutron multiplication factor suppression effect can be provided, thereby enabling to reduce the number of burnable poison rods relative to fuels. As a result, power peaking in the fuel assembly is moderated and a greater amount of plutonium can be loaded. In addition the flow of cooling water which tends to gather around the large diameter water rod can be controlled to improve cooling performance of fuels. (N.H.)

  7. Testing plutonium fuel assembly production for fast-neutron reactors

    International Nuclear Information System (INIS)

    Nougues, B.; Benhamou, A.; Bertothy, G.; Lepetit, H.

    1975-01-01

    The main characteristics of plutonium fuel elements for fast breeder reactors justify specific test procedures and special techniques. The specific tests relating to the Pu content consist of Pu enrichment and distribution tests, determination of the O/M ratio and external contamination tests. The specific tests performed on fuel configuration are: testing of sintered pellet diameter, testing of pin welding and checking of internal assmbly [fr

  8. Neutron collar calibration for assay of LWR [light-water reactor] fuel assemblies

    International Nuclear Information System (INIS)

    Menlove, H.O.; Pieper, J.E.

    1987-03-01

    The neutron-coincidence collar is used for the verification of the uranium content in light-water reactor fuel assemblies. An AmLi neutron source is used to give an active interrogation of the fuel assembly to measure the 235 U content, and the 238 U content is verified from a passive neutron-coincidence measurement. This report gives the collar calibration data of pressurized-water reactor and boiling-water reactor fuel assemblies. Calibration curves and correction factors are presented for neutron absorbers (burnable poisons) and different fuel assembly sizes. The data were collected at Exxon Nuclear, Franco-Belge de Fabrication de Combustibles, ASEA-Atom, and other nuclear fuel fabrication facilities

  9. Device for refueling a nuclear reactor having a core comprising a plurality of fuel assemblies

    International Nuclear Information System (INIS)

    Van Santen, A.; Elofsson, K.

    1975-01-01

    A nuclear reactor formed of fuel assemblies each including a plurality of parallel fuel rods arranged in a predetermined fuel rod lattice, which rods are freely extractable and insertable at one end of the fuel assembly, is refueled by extracting from one of the fuel assemblies a number of fuel rods substantially less than the total number of fuel rods and replacing these by inserting new fuel rods into the vacated positions. The removal and return of the rods is produced by a tool having a plurality of gripping members capable of engaging shoulders beneath heads formed on the upper ends of the fuel rods. This may be accomplished by providing a tool having a number of gripping members attached to the tool body corresponding to the lattice positions of the fuel rods to be extracted, having gripping members which can be pushed together to grip beneath shoulders on the upper ends of the fuel rods. (Official Gazette)

  10. On-line system for monitoring of boiling in nuclear reactor fuel assemblies

    International Nuclear Information System (INIS)

    Tuerkcan, E.; Kozma, R.; Verhoef, J.P.; Nabeshima, K.

    1996-01-01

    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

  11. Fissile fuel assembly for a sub-moderated nuclear reactor

    International Nuclear Information System (INIS)

    Millot, J.P.; Dejeux, Pol.; Alibran, Patrice.

    1983-01-01

    Each of the core assemblies is composed of a prismatic case made of a neutron absorbing material, inside which very long rods containing the fissile material are arranged parallel to the height of the case and according to a regular network in the straight sections of the case. At least one piece in a fertile material exposed to the neutrons emitted by the fissile material of the assembly is arranged on each one of the side faces of the case. The invention applies in particular to sub-moderated reactors, cooled and moderated by pressurized water [fr

  12. Detailed channel thermal-hydraulic calculation of nuclear reactor fuel assemblies

    International Nuclear Information System (INIS)

    Zhukov, A.V.; Sorokin, A.P.; Ushakov, P.A.; Yur'ev, Yu.S.

    1981-01-01

    The system of equations of mass balance, quantity of motion and energy used in calculation of nuclear reactor fuel assemblies is obtained. The equation system is obtained on the base of integral equations of hydrodynamics interaction in assemblies of smooth fuel elements and fuel elements with wire packing. The calculation results of coolant heating distributions by the fast reactor assembly channels are presented. The analysis of the results obtained shows that interchannel exchange essentially uniforms the coolant heating distribution in the peripheral range of the assembly but it does not remove non-uniformity caused by power distribution non-uniformity in the cross section. Geometry of the peripheral assembly range plays an essential role in the heating distribution. Change of the calculation gap between the peripheral fuel elements and assembly shells can result either in superheating or in subcooling in the peripheral channels relatively to joint internal channels of the assembly. Heat supply to the coolant passing through interassembly gaps decreases temperature in the assembly periphery and results in the increase of temperature non-uniformity by the perimeter of peripheral fuel elements. It is concluded that the applied method of the channel-by-channel calculation is ef-- fective in thermal-physical calculation of nuclear reactor fuel assemblies and it permits to solve a wide range of problems [ru

  13. Fuel assembly leak tightness control on WWER-1000 reactor

    International Nuclear Information System (INIS)

    Ivanova, R.; Gerchev, N.; Mateev, A.

    2001-01-01

    The main index for integrity of the fuel rods cladding is the specific activity value of the primary coolant. This value determines the safe operation of the reactor. The limit for safe operation of WWER-1000 reactor is the value of the total activity of Iodine isotopes in the primary coolant 5.0x10 -3 Ci/l. The paper briefly describes the methodology for performing a fuel tightness test (sipping test) and shows the results from these tests performed during the period 1987 -1999 in units 5 and 6 at the Kozloduy NPP. An additional index related to the safe operation is defined to characterize the fuel cladding integrity Fuel Reliability Index (FRI). The FRI is defined as value of the average activity of 131 I in the primary coolant, corrected with a part of precipitated 235 U migration and fixed to the general permanent purification frequency. Two criteria (quantitative and statistic) are determined to qualify the fuel cladding integrity. The results from sipping tests show good reliability of the fuel irradiated in unit 5 and 6 at the Kozloduy NPP

  14. Packaging and transport case of test fuel assembly irradiated in the Creys-Malville reactor

    International Nuclear Information System (INIS)

    Geffroy, J.; Vivien, J.; Pouard, M.; Dujardin, G.N.; Veron, B.; Michoux, H.

    1986-06-01

    Some irradiated fuel assemblies from the fast neutron Creys Malville reactor will be sent to hot laboratories to follow fuel behavior. These test assemblies will be examined after a limited cooling time and transport is realized at high residual power (about 10kW) and cladding temperature should not rise over 500deg C. The fuel assemblies are not dismantled and transported into sodium. The assembly is placed into a case containing sodium plugged and put into a packaging. Dimensioning, thermal behavior, radiation protection and containment are examined [fr

  15. Comparison of problems and experience of core operation with distorted fuel element assemblies in VVER-1000 and PWR reactors

    International Nuclear Information System (INIS)

    Afanas'ev, A.

    1999-01-01

    The main reactors leading to distortion of fuel element assemblies during reactor operation were studied. A series of actions which compensate this effect was proposed. Criteria of operation limitation in VVER-1000 and PWR reactors are described

  16. Improvements in or relating to nuclear reactor fuel element assemblies

    International Nuclear Information System (INIS)

    Chetter, J.

    1975-01-01

    A description is given of a spacer grid comprising a substantially rigid grid structure formed from intersecting strip members defining cells which are penetrated by fuel pins bearing against rigid stops projecting inside the cells and spring locating members in the form of bow springs which extend longitudinally in the cells of the grid structure to hold the fuel pins against the rigid stops in the cells. The bow spring members of each line of cells extending in one direction across the grid structure have their corresponding ends interconnected by common longitudinal bridging strips to form a ladder spring assembly. (author)

  17. Nuclear reactor fuel assembly with a removably top nozzle

    International Nuclear Information System (INIS)

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

    1985-01-01

    The invention relates to a nuclear fuel assembly having an improved attaching structure for removably mounting the top nozzle of the fuel assembly on the upper end of a control-rod guide thimble. The attaching structure comprises an outer socket defined in a portion of the top nozzle, an inner socket extending from the upper end of the guide thimble and removably received in the outer socket for interlocking engagement therewith, and an elongate locking member adapted to be inserted into the inner socket to maintain said interlocking engagement. Removal of the locking member from the inner socket enables the latter to be withdrawn from the outer socket, thereby enabling the top nozzle to be removed from the guide thimble

  18. Nuclear fuel string assembly

    International Nuclear Information System (INIS)

    Ip, A.K.; Koyanagi, K.; Tarasuk, W.R.

    1976-01-01

    A method of fabricating rodded fuels suitable for use in pressure tube type reactors and in pressure vessel type reactors is described. Fuel rods are secured as an inner and an outer sub-assembly, each rod attached between mounting rings secured to the rod ends. The two sub-assemblies are telescoped together and positioned by spaced thimbles located between them to provide precise positioning while permittng differential axial movement between the sub-assemblies. Such sub-assemblies are particularly suited for mounting as bundle strings. The method provides particular advantages in the assembly of annular-section fuel pins, which includes booster fuel containing enriched fuel material. (LL)

  19. Method for determining the outlet temperature of fuel assemblies unsupplied with thermometer in WWER-440 reactors

    International Nuclear Information System (INIS)

    Miko, S.; Kalya, Z.; Hamvas, I.

    1987-09-01

    The paper outlines a method for the evaluation of the outlet temperatures of fuel assemblies unsupplied with thermometer in WWER-440 reactors. The process is based on interpolation of directly measured assembly temperatures. A quantitative comparison of the errors of described algorithm to those of standard plant-computer interpolation rutine is also presented. (author)

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

    International Nuclear Information System (INIS)

    Barragan M, A.; Martin del Campo M, C.; Francois L, J. L.; Espinosa P, G.

    2012-10-01

    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 (PPF max ), as well as the reactivity coefficients by the fuel temperature. The k∞ and PPF max 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 - Gd 2O3 ) in some rods of the same assembly. The obtained results show values k∞ and PPF max 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)

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

    International Nuclear Information System (INIS)

    Liu Wenjin; Zeng Zhongxiu; Ye Xianhui; Wu Wanjun

    2013-01-01

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

  2. Irradiation performance of experimental fast reactor 'JOYO' MK-1 driver fuel assemblies

    International Nuclear Information System (INIS)

    Itaki, Toshiyuki; Kono, Keiichi; Tachi, Hirokatsu; Yamanouchi, Sadamu; Yuhara, Shunichi; Shibahara, Itaru

    1985-01-01

    The experimental fast reactor ''JOYO'' completed it's breeder core (MK-I) operation in January 1982. The MK-I driver fuel assemblies were removed from the core sequencially in order of burnup increase and have been under postirradiation examination (PIE). The PIE has almost been completed for 30 assemblies including the highest burnup assemblies of 48,000 MWD/MTM. It has been confirmed that all fuel assemblies have exhibited satisfactory performance without detrimental assembly deformation or without any indications of fuel pin breach. The irradiation conditions of the MK-I core were somewhat more moderate than those conditions envisioned for prototypic reactor. However the results of the examination revealed the typical irradiation behavior of LMFBR fuels, although such characteristics were benign as compared with those anticipated in high burnup fuels. Systematic performance data have been accumulated through the fuel fabrication, irradiation and postirradiation examination processes. Based on these data, the MK-I fuel designing and fabrication techniques were totally confirmed. This technical experience and the associated insight into irradiation behavior have established a milestone to the next step of fast reactor fuel development. (author)

  3. Fuel assembly

    International Nuclear Information System (INIS)

    Fushimi, Atsushi; Shimada, Hidemitsu; Aoyama, Motoo; Nakajima, Junjiro

    1998-01-01

    In a fuel assembly for an n x n lattice-like BWR type reactor, n is determined to 9 or greater, and the enrichment degree of plutonium is determined to 4.4% by weight or less. Alternatively, n is determined to 10 or greater, and the enrichment degree of plutonium is determined to 5.2% by weight or less. An average take-out burnup degree is determined to 39GWd/t or less, and the matrix is determined to 9 x 9 or more, or the average take-out burnup degree is determined to 51GWd/t, and the matrix is determined to 10 x 10 or more and the increase of the margin of the maximum power density obtained thereby is utilized for the compensation of the increase of distortion of power distribution due to decrease of the kinds of plutonium enrichment degree, thereby enabling to reduce the kind of the enrichment degree of MOX fuel rods to one. As a result, the manufacturing step for fuel pellets can be simplified to reduce the manufacturing cost for MOX fuel assemblies. (N.H.)

  4. An alternative solution for heavy liquid metal cooled reactors fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Vitale Di Maio, Damiano, E-mail: damiano.vitaledimaio@uniroma1.it [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy); Cretara, Luca; Giannetti, Fabio [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy); Peluso, Vincenzo [“ENEA”, Via Martiri di Monte Sole 4, 40129 Bologna (Italy); Gandini, Augusto [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy); Manni, Fabio [“SRS Engineering Design S.r.l.”, Vicolo delle Palle 25-25/b, 00186 Rome (Italy); Caruso, Gianfranco [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy)

    2014-10-15

    Highlights: • A new fuel assembly locking system for heavy metal cooled reactor is proposed. • Neutronic, mechanical and thermal-hydraulic evaluations of the system behavior have been performed. • A comparison with other solutions has been presented. - Abstract: In the coming future, the electric energy production from nuclear power plants will be provided by both thermal reactors and fast reactors. In order to have a sustainable energy production through fission reactors, fast reactors should provide an increasing contribution to the total electricity production from nuclear power plants. Fast reactors have to achieve economic and technical targets of Generation IV. Among these reactors, Sodium cooled Fast Reactors (SFRs) and Lead cooled Fast Reactors (LFRs) have the greatest possibility to be developed as industrial power plants within few decades. Both SFRs and LFRs require a great R and D effort to overcome some open issues which affect the present designs (e.g. sodium-water reaction for the SFRs, erosion/corrosion for LFRs, etc.). The present paper is mainly focused on LFR fuel assembly (FA) design: issues linked with the high coolant density of lead or lead–bismuth eutectic cooled reactors have been investigated and an innovative solution for the core mechanical design is here proposed and analyzed. The solution, which foresees cylindrical fuel assemblies and exploits the buoyancy force due to the lead high density, allows to simplify the FAs locking system, to reduce their length and could lead to a more uniform neutron flux distribution.

  5. Description and performance characteristics for the neutron Coincidence Collar for the verification of reactor fuel assemblies

    International Nuclear Information System (INIS)

    Menlove, H.O.

    1981-08-01

    An active neutron interrogation method has been developed for the measurement of 235 U content in fresh fuel assemblies. The neutron Coincidence Collar uses neutron interrogation with an AmLi neutron source and coincidence counting the induced fission reaction neutrons from the 235 U. This manual describes the system components, operation, and performance characteristics. Applications of the Coincidence Collar to PWR and BWR types of reactor fuel assemblies are described

  6. Nondestructive determination of burnup and fissile isotope balance in spent fuel assemblies of water cooled reactors

    International Nuclear Information System (INIS)

    Pinel, J.

    1983-03-01

    Two non-destructive methods for measuring fuel assemblies in storage pools have been developed: a gamma fuel scanning method, using the 134 Cs - 137 Cs and 144 Ce gamma rays, and the measurement of the neutron flux emitted by the fuel assembly. For interpreting the measurement, we have used calculated correlations to establish a connection between the measured phenomena and the parameters to be determined. A measurement campaign involving 58 assemblies from the C.N.A. reactor was conducted in the reprocessing plant of LA HAGUE. The results obtained show that the objectives can be achevied within an industrial environment [fr

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

    International Nuclear Information System (INIS)

    Meuschke, R.E.; Satterlee, A.E.

    1978-01-01

    A combined fuel assembly and thimble plug gripper for raising and lowering a fuel assembly into a nuclear reactor core, and for lifting and lowering a thimble plug assembly into the fuel assembly is described. It includes a vertically movable mast housing a mechanism which causes pivotally mounted fingers on the bottom of the mast to be moved into and out of latching engagement with the nozzle of a fuel assembly when the mast is resting on the assembly. The mast includes a second mechanism which supports second fingers pivotally mounted thereon and actuable by a third mechanism into and out of engagement with a thimble plug assembly supporting plugs adapted to be inserted in control rod guide thimbles in the fuel assembly. The second mechanism further includes an arrangement for lowering or raising the plug assembly respectively into or out of the guide thimbles in the fuel assembly. The apparatus includes control and interlock systems which preclude operation of the mechanisms under certain prescribed conditions

  8. Fuel assemblies

    International Nuclear Information System (INIS)

    Yoshioka, Ritsuo.

    1983-01-01

    Purpose: To improve the operation performance of a BWR type reactor by improving the distribution of the uranium enrichment and the incorporation amount of burnable poisons in fuel assemblies. Constitution: The average enrichment of uranium 235 is increased in the upper portion as compared with that in the lower portion, while the incorporation amount of burnable poisons is increased in an upper portion as compared with that in the lower portion. The difference in the incorporation amount of the burnable poisons between the upper and lower portions is attained by charging two kinds of fuel rods; the ones incorporated with the burnable poisons over the entire length and the others incorporated with the burnable poisons only in the upper portions. (Seki, T.)

  9. Calculation of local characteristics of velocity field in turbulent coolant flow in fast reactor fuel assembly

    International Nuclear Information System (INIS)

    Muehlbauer, P.

    1981-08-01

    Experience is described gained with the application of computer code VELASCO in calculating the velocity field in fast reactor fuel assemblies taking into account configuration disturbances due to fuel pin displacement. Theoretical results are compared with the results of experiments conducted by UJV on aerodynamic models HEM-1 (model of the fuel assembly central part) and HEM-2 (model of the fuel assembly peripheral part). The results are reported of calculating the distribution of shear stress in wetted rod surfaces and in the assembly wall (model HEM-2) and the corresponding experimental results are shown. The shear stress distribution in wetted surfaces obtained using the VELASCO code allowed forming an opinion on the code capability of comprising local parameters of turbulent flow through a fuel rod bundle. The applicability was also tested of the code for calculating mean velocities in the individual zones, eg., in elementary cells. (B.S.)

  10. Seismic analysis of fuel and target assemblies at a production reactor

    International Nuclear Information System (INIS)

    Braverman, J.I.; Wang, Y.K.

    1991-01-01

    This paper describes the unique modeling and analysis considerations used to assess the seismic adequacy of the fuel and target assemblies in a production reactor at Savannah River Site. This confirmatory analysis was necessary to provide assurance that the reactor can operate safely during a seismic event and be brought to a safe shutdown condition. The plant which was originally designed in the 1950's required to be assessed to more current seismic criteria. The design of the reactor internals and the magnitude of the structural responses enabled the use of a linear elastic dynamic analysis. A seismic analysis was performed using a finite element model consisting of the fuel and target assemblies, reactor tank, and a portion of the concrete structure supporting the reactor tank. The effects of submergence of the fuel and target assemblies in the water contained within the reactor tank can have a significant effect on their seismic response. Thus, the model included hydrodynamic fluid coupling effects between the assemblies and the reactor tank. Fluid coupling mass terms were based on formulations for solid bodies immersed in incompressible and frictionless fluids. The potential effects of gap conditions were also assessed in this evaluation. 5 refs., 6 figs., 1 tab

  11. Modified fuel assembly design for pressurized water reactors with improved fuel utilization

    International Nuclear Information System (INIS)

    Galperin, A.; Ronen, Y.

    1983-01-01

    A method for reactivity control through variation of the moderator content in the reactor core was proposed. The main idea is to adjust the amount of water in the core from a low value at beginning of cycle to a high value at end of cycle, so as to compensate for fissile material burnup and buildup of fission products. The possible implementation of this idea may be carried out by introducing a number of hollow tubes into the fuel assembly between the fuel rods. Then variation of the moderator content in the core may be managed through a change of the water level in these tubes. cated a potential savings in the fuel cycle requirements and costs. Preliminary steady-state thermal-hydraulic calculations indicate the possibility of implementing the proposed method in the existing pressurized water reactor plants. Feasibility of the proposed design may be finally established after rigorous thermal hydraulics as well as safety analysis calculations. Furthermore, there is need to elaborate the mechanical design of the pressure vessel internals together with cost benefit analysis

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

    International Nuclear Information System (INIS)

    Pond, R.B.; Matos, J.E.

    1996-05-01

    As part of the Department of Energy's spent nuclear fuel acceptance criteria, the mass of uranium and transuranic elements in spent research reactor fuel must be specified. These data are, however, not always known or readily determined. It is the purpose of this report to provide estimates of these data for some of the more common research reactor fuel assembly types. The specific types considered here are MTR, TRIGA and DIDO fuel assemblies. 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 also, are not always known or readily determined. Because of a self-protecting dose rate level of radiation (dose rate greater than 100 ren-x/h at I m in air), it is important to know the dose rate of spent fuel assemblies at all time. Estimates of the photon dose rate for spent MTR, TRIGA and DIDO-type fuel assemblies are given in this report

  13. Preliminary study on the feasibility of ductless fuel assembly for fast reactors

    International Nuclear Information System (INIS)

    Shibahara, Itaru; Enokido, Yuji

    1988-01-01

    Preliminary study on the feasibility of ductless fuel assembly for fast reactors has been conducted. The primary concern is with forecasting the thermal hydraulic characteristics and the heat removal efficiency from the core. The thermal hydraulic analysis revealed the coolant mixing in the core at steady state operating condition was not intensive and the coolant temperature increase was almost proportional to the power of each assembly. The hot spot analysis of the ductless core indicated that the hottest temperature in the core could be comparable with the temperature of the conventional ducted core, even in case the radial power flattening was not actively pursued but with adopting ducted radial blanket assemblies. Under off-normal conditions, the ductless core had improved heat removal capability which was caused by inter-assembly coolant flow. The study has indicated the feasibility of the ductless fuel assembly for fast reactors. The experiments to demonstrate the feasibility will be the next key process for the development. (author)

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

    International Nuclear Information System (INIS)

    Nash, C.A.; Blake, J.E.; Rush, G.C.

    1990-01-01

    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/m 2 ) (1.1E+6 BTU/(ft 2 hr)) 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

  15. Physical characteristics of non-fuel assembly reactor components

    International Nuclear Information System (INIS)

    Hawkes, E.C.

    1994-09-01

    The primary objective of this report is to enhance the utility of the Characteristics Data Base (CDB). This has been accomplished by providing a pictorial representation of the principal non-fuel assembly (NFA) components along with a tabular summary of key information about each type of component. This report is intended for use as an adjunct to the CDB. Toward this end, the report may be used either as a complement to the detailed descriptions in the CDB, or as a stand-alone document that acts as an illustrated abstract of the CDB. Line drawings of major NFA components are included. Data not provided in the CDB are also included. Summary descriptions of each component are given in tabular format

  16. Judgement on the data for fuel assembly outlet temperatures of WWER fuel assemblies in power reactors based on measurements with experimental fuel assemblies

    International Nuclear Information System (INIS)

    Krause, F.

    1986-01-01

    In the period from 1980 to 1985, in the Rheinsberg nuclear power plant experimental fuel assemblies were used on lattices at the periphery of the core. These particular fuel assemblies dispose of an extensive in-core instrumentation with different sensors. Besides this, they are fit out with a device to systematically thottle the coolant flow. The large power gradient present at the core position of the experimental fuel assembly causes a temperature profile along the fuel assemblies which is well provable at the measuring points of the outlet temperature. Along the direction of flow this temperature profile in the coolant degrades only slowly. This effect is to be taken into account when measuring the fuel assembly outlet temperature of WWER fuel assemblies. Besides this, the results of the measurements hinted both at a γ-heating of the temperature measuring points and at tolerances in the calculation of the micro power density distribution. (author)

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

    International Nuclear Information System (INIS)

    Voorbraak, W.P.; Paardekoper, A.; Polle, A.N.; Freudenreich, W.E.

    1993-08-01

    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.)

  18. The conceptual design of the standard and the reduced fuel assemblies for an advanced research reactor

    International Nuclear Information System (INIS)

    Ryu, Jeong Soo; Cho, Yeong Garp; Yoon, Doo Byung; Dan, Ho Jin; Chae, Hee Tack; Park, Cheol

    2005-01-01

    HANARO (Hi-flux Advanced Neutron Application Reactor), is an open-tank-in-pool type research reactor with a thermal power of 30MW. The HANARO has been operating at Korea Atomic Energy Research Institute since 1995. Based on the technical experiences in design and operation for the HANARO, the design of an Advanced Research Reactor (ARR) was launched by KAERI in 2002. The final goal of the project is to develop a new and advanced research reactor model which is superior in safety and economical aspects. This paper summarizes the design improvements of the conceptually designed standard fuel assembly based on the analysis results for the nuclear physics. It includes also the design of the reduced fuel assembly in conjunction with the flow tube as the fuel channel and the guide of the absorber rod. In the near future, the feasibility of the conceptually designed fuel assemblies of the ARR will be verified by investigating the dynamic and the thermal behaviors of the fuel assembly submerged in coolant

  19. Analysis of the equalizing holes resistance in fuel assembly spike for lead-based reactor

    International Nuclear Information System (INIS)

    Zhang, Guangyu; Jin, Ming; Wang, Jianye; Song, Yong

    2017-01-01

    Highlights: • A RELAP5 model for a 10 MWth lead-based reactor was built to study the hydrodynamic characteristics between the equalizing holes in the fuel assembly spike. • Different fuel assembly total blockage scenarios and different resistances for different fuel assemblies were examined. • The inherent safety characteristics of the lead-based reactor was improved by optimizing the configuration of equalizing holes in the fuel assembly spike. - Abstract: To avoid the damage of the fuel rod cladding when a fuel assembly (FA) is totally blocked, a special configuration of the fuel assembly spike was designed with some equalizing holes in the center region which can let the coolant to flow during the totally blockage scenarios of FA. To study the hydrodynamic characteristics between the equalizing holes and an appropriate resistance, a RELAP5 model was developed for a 10 MWth lead-based reactor which used lead-bismuth as coolant. Several FA total blockage and partial core blockage scenarios were selected. The simulation results indicated that when all the FA spike equalizing holes had the same hydraulic resistance, only a narrow range of suitable equalizing holes resistances could be chosen when a FA was blocked. However, in the two or more FA blockage scenarios, there were no appropriate resistances to meet the requirement. In addition, with different FA spike equalizing holes with different resistances, a large range of suitable equalizing hole resistances could be chosen. Especially a series of suitable resistances were selected when the small power FA resistance was 1/2, 1/4, 1/8 of the large one. Under these circumstances, one, two or three FA blockages would not damage the core. These demonstrated that selecting a series of suitable hydraulic resistances for the equalizing holes could improve the safety characteristics of the reactor effectively.

  20. Fuel assembly

    International Nuclear Information System (INIS)

    Hiraiwa, Koji; Ueda, Makoto

    1989-01-01

    In a fuel assembly used for a light water cooled reactor such as a BWR type reactor, a water rod is divided axially into an upper outer tube and a lower outer tube by means of a plug disposed from the lower end of a water rod to a position 1/4 - 1/2 of the entire length for the water rod. Inlet apertures and exit apertures for moderators are respectively perforated for the divided outer tube and upper and lower portions. Further, an upper inner tube with less neutron irradiation growing amount than the outer tube is perforated on the plug in the outer tube, while a lower inner tube with greater neutron irradiation growing amount than the outer tube is suspended from the lower surface of the plug in the outer tube. Then, the opening area for the exit apertures disposed to the upper outer tube and the lower outer tube is controlled depending on the difference of the neutron irradiation growing amount between the upper inner tube and the upper outer tube, and the difference of the neutron irradiation growing amount between the lower inner tube and the lower outer tube. This enables effective spectral shift operation and improve the fuel economy. (T.M.)

  1. Comparison of thermal capabilities of the fuel assemblies for the WWR-M reactor

    International Nuclear Information System (INIS)

    Kirsanov, G.A.; Konoplev, K.A.; Findeisen, A.; Shishkina, Zh.A.

    1989-01-01

    On the basis of measurement results of the WWR-M2, WWR-M3 and WWR-M5 fuel element can temperature in the WWR-M reactor core their thermal capabilities are compared. The use of the WWR-M5 fuel assemblies instead of the WWR-M2 ones in the WWR-M reactor permits to increase specific heat loading by a factor of 2.7. The possibility to increase fuel can temperature up to 110 deg C is confirmed experimentally which corresponds to specific heat loading of 900 kW/l

  2. Buckling resistance calculation of Guide Thimbles for the mechanical design of fuel assembly type PWR under normal reactor operating conditions

    International Nuclear Information System (INIS)

    Cruz, C.B.L.

    1990-01-01

    The calculations demonstrate the fulfillment of one of the mechanical design criteria for the Fuel Assembly Structure under normal reactor operating conditions. The calculations of stresses in the Guide Thimbles are performed with the aid of the program ANSYS. This paper contains program parameters and modelling of a typical Fuel Assembly for a Reactor similar to ANGRA II. (author)

  3. Fuel assembly

    International Nuclear Information System (INIS)

    Wataumi, Kazutoshi; Tajiri, Hiroshi.

    1992-01-01

    In a fuel assembly of a BWR type reactor, a pellet to be loaded comprises an external layer of fissile materials containing burnable poisons and an internal layer of fissile materials not containing burnable poison. For example, there is provided a dual type pellet comprising an external layer made of UO 2 incorporated with Gd 2 O 3 at a predetermined concentration as the burnable poisons and an internal layer made of UO 2 not containing Gd 2 O 3 . The amount of the burnable poisons required for predetermined places is controlled by the thickness of the ring of the external layer. This can dissipate an unnecessary poisoning effect at the final stage of the combustion cycle. Further, since only one or a few kinds of powder mixture of the burnable poisons and the fissile materials is necessary, production and product control can be facilitated. (I.N.)

  4. On numerical simulation of fuel assembly bow in pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Horváth, Ákos, E-mail: akoshorvath@t-online.hu [AREVA, AREVA NP GmbH, Paul-Gossen-Str. 100, 91052 Erlangen (Germany); Budapest University of Technology and Economics, Department of Aircraft and Ships, Stoczek Street 6, Building J, H-1111 Budapest (Hungary); Dressel, Bernd [AREVA, AREVA NP GmbH, Paul-Gossen-Str. 100, 91052 Erlangen (Germany)

    2013-12-15

    Highlights: • Simulation of fuel assembly bow by coupled CFD and finite element method. • Comparison of calculated and experimentally measured bow shapes. • Investigation of boundary condition effect on bow pattern of a fuel assembly row. • Highlighting importance of consideration of fluid–structure interaction. • Assessment of flow redistribution within the fuel assembly row model. - Abstract: Fuel assembly bow in pressurized water reactor cores is largely triggered by lateral hydraulic forces together with creep processes generated by neutron flux. A detailed understanding of the flow induced bow behaviour is, therefore, an important issue. The experimental feedbacks and laboratory tests on fuel assembly bow show that it is characterized to a high degree by fluid–structure interaction (FSI) effects, therefore, consideration of FSI is essential and indispensable in full comprehension of the bow mechanism. In the present study, coupled computational fluid dynamics (CFD) and finite element simulations are introduced, calculating fuel assembly deformation under different conditions as a quasi-stationary phenomenon. The aim has been, on the one hand, to develop such a simplified fuel assembly CFD model, which allows set up of fuel assembly rows without loosing its main hydraulic characteristic; on the other hand, to investigate the bow pattern of a given fuel assembly row under different boundary conditions. The former one has been achieved by comparing bow shapes obtained with different fuel assembly (spacer grid) modelling approaches and mesh resolutions with experimental data. In the second part of the paper a row model containing 7.5 fuel assemblies is introduced, investigating the effect of flow distribution at inlet and outlet boundary regions on fuel assembly bow behaviour. The post processing has been focused on the bow pattern, lateral hydraulic forces, and horizontal flow distribution. The results have revealed importance of consideration of

  5. Fuel assembly

    International Nuclear Information System (INIS)

    Hirukawa, Koji; Sakurada, Koichi.

    1992-01-01

    In a fuel assembly for a BWR type reactor, water rods or water crosses are disposed between fuel rods, and a value with a spring is disposed at the top of the coolant flow channel thereof, which opens a discharge port when pressure is increased to greater than a predetermined value. Further, a control element for the amount of coolant flow rate is inserted retractable to a control element guide tube formed at the lower portion of the water rod or the water cross. When the amount of control elements inserted to the control element guide tube is small and the inflown coolant flow rate is great, the void coefficient at the inside of the water rod is less than 5%. On the other hand, when the control elements are inserted, the flow resistance is increased, so that the void coefficient in the water rod is greater than 80%. When the pressure in the water rod is increased, the valve with the spring is raised to escape water or steams. Then, since the variation range of the change of the void coefficient can be controlled reliably by the amount of the control elements inserted, and nuclear fuel materials can be utilized effectively. (N.H.)

  6. Nuclear reactor fuel assembly with a removable top nozzle

    International Nuclear Information System (INIS)

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

    1986-01-01

    This patent describes a fuel assembly having at least one control rod guide thimble and a top nozzle, the top nozzle including a transversely extending adapter plate. An improved attaching structure is described for removably mounting the top nozzle on the guide thimble comprising: (a) means defining an outer socket in the top nozzle, the outer socket defining means including a passageway extending through the adapter plate and having a first mating element defined in the adapter plate within the passageway; (b) means on an upper end of the guide thimble defining an inner socket, the inner socket defining means including an elongated sleeve having an upper end portion. The upper end portion of the sleeve has a second mating element formed thereon and at least one elongated axial slot defined therein for permitting radial movement of the sleeve upper end portion between a compressed releasing position for removing and inserting the inner socket from and into the outer socket and an expanded locking position for locking the inner and outer sockets together

  7. Transfer hook for nuclear fuel assemblies and nuclear reactor having a such hook

    International Nuclear Information System (INIS)

    Thevenot, L.P.

    1990-01-01

    For removing irradiated nuclear fuel assemblies above the level of the liquid metal in the vessel without loss of cooling, the hook mechanism has a guide tube with two annular cavities and a pump to circulate the reactor cooling fluid which flows out by gravity. A such hook used in a LMFBR reduces the height of the reactor vessel and consequently the initial capital cost [fr

  8. Comparison of the parameters of the IR-8 reactor with different fuel assembly designs with LEU fuel

    International Nuclear Information System (INIS)

    Vatulin, A.; Stetsky, Y.; Dobrikova, I.

    1999-01-01

    The estimation of neutron-physical, heat and hydraulic parameters of the IR-8 research reactor with low enriched uranium (LEU) fuel was performed. Two fuel assembly (FA) designs were reviewed: IRT-4M with the tubular type fuel elements and IRT-MR with the rod type fuel elements. UO 2 -Al dispersion 19.75% enrichment fuel is used in both cases. The results of the calculations were compared with main parameters of the reactor, using the current IRT-3M FA with 90% high enriched uranium (HEU) fuel. The results of these comparisons showed that during the LEU conversion of the reactor the cycle length, excess reactivity and peak power of the IRT-MR type FA are higher than for the IRT-3M type FA and IRT-4M type FA. (author)

  9. Dose rate estimates from irradiated light-water-reactor fuel assemblies in air

    International Nuclear Information System (INIS)

    Lloyd, W.R.; Sheaffer, M.K.; Sutcliffe, W.G.

    1994-01-01

    It is generally considered that irradiated spent fuel is so radioactive (self-protecting) that it can only be moved and processed with specialized equipment and facilities. However, a small, possibly subnational, group acting in secret with no concern for the environment (other than the reduction of signatures) and willing to incur substantial but not lethal radiation doses, could obtain plutonium by stealing and processing irradiated spent fuel that has cooled for several years. In this paper, we estimate the dose rate at various distances and directions from typical pressurized-water reactor (PWR) and boiling-water reactor (BWR) spent-fuel assemblies as a function of cooling time. Our results show that the dose rate is reduced rapidly for the first ten years after exposure in the reactor, and that it is reduced by a factor of ∼10 (from the one year dose rate) after 15 years. Even for fuel that has cooled for 15 years, a lethal dose (LD50) of 450 rem would be received at 1 m from the center of the fuel assembly after several minutes. However, moving from 1 to 5 m reduces the dose rate by over a factor of 10, and moving from 1 to 10 m reduces the dose rate by about a factor of 50. The dose rates 1 m from the top or bottom of the assembly are considerably less (about 10 and 22%, respectively) than 1 m from the center of the assembly, which is the direction of the maximum dose rate

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-10-01

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

  11. Analyses for inserting fresh LEU fuel assemblies instead of fresh HEU fuel assemblies in the Dalat Nuclear Research Reactor in Vietnam

    International Nuclear Information System (INIS)

    Hanan, N. A.; Deen, J.R.; Matos, J.E.

    2005-01-01

    Analyses were performed by the RERTR Program to replace 36 burned HEU (36%) fuel assemblies in the Dalat Nuclear Research Reactor in Vietnam with either 36 fresh fuel assemblies currently on-hand at the reactor or with LEU fuel assemblies to be procured. The study concludes that the current HEU (36%) WWR-M2 fuel assemblies can be replaced with LEU WWR-M2 fuel assemblies that are fully-qualified and have been commercially available since 2001 from the Novosibirsk Chemical Concentrates Plant in Russia. The current reactor configuration using re-shuffled HEU fuel began in June 2004 and is expected to allow normal operation until around August 2006. If 36 HEU assemblies each with 40.2 g 235 U are inserted without fuel shuffling over the next five operating cycles, the core could operate for an additional 10 years until June 2016. Alternatively, inserting 36 LEU fuel assemblies each containing 49.7 g 235 U without fuel shuffling over five operating cycles would allow normal operation for about 14 years from August 2006 until October 2020. The main reason for the longer service life of the LEU fuel is that its 235 U content is higher than the 235 U content needed simply to match the service life of the HEU fuel. Fast neutron fluxes in the experiment regions would be very nearly the same in both the HEU and LEU cores. Thermal neutron fluxes in the experiment regions would be lower by 1-5%, depending on the experiment type and location. (author)

  12. Nuclear fuel assembly

    International Nuclear Information System (INIS)

    Delafosse, Jacques.

    1977-01-01

    This invention relates to a nuclear fuel assembly for a light or heavy water reactor, or for a fast reactor of the kind with a bundle of cladded pins, maintained parallel to each other in a regular network by an assembly of separate supporting grids, fitted with elastic bearing surfaces on these pins [fr

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

    International Nuclear Information System (INIS)

    Yokota, Yoshio; Ko, Kuniaki.

    1995-01-01

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

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

    International Nuclear Information System (INIS)

    Urban, P.; Bender, D.

    2001-01-01

    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)

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

    International Nuclear Information System (INIS)

    Vuckovich, M.; Burkett, J.P.; Sallustio, J.

    1983-01-01

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

    Vuckovich, M; Burkett, J P; Sallustio, J

    1981-11-30

    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.

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

    International Nuclear Information System (INIS)

    Ryu, Jeong Soo; Dan, Ho Jin; Cho, Yeong Garp; Yoon, Doo Byung; Park, Cheol

    2005-05-01

    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

  18. Fuel assembly

    International Nuclear Information System (INIS)

    Nakatsuka, Masafumi; Matsuzuka, Ryuji.

    1976-01-01

    Object: To provide a fuel assembly which can decrease pressure loss of coolant to uniform temperature. Structure: A sectional area of a flow passage in the vicinity of an inner peripheral surface of a wrapper tube is limited over the entire length to prevent the temperature of a fuel element in the outermost peripheral portion from being excessively decreased to thereby flatten temperature distribution. To this end, a plurality of pincture-frame-like sheet metals constituting a spacer for supporting a fuel assembly, which has a plurality of fuel elements planted lengthwise and in given spaced relation within the wrapper tube, is disposed in longitudinal grooves and in stacked fashion to form a substantially honeycomb-like space in cross section. The fuel elements are inserted and supported in the space to form a fuel assembly. (Kamimura, M.)

  19. Fuel assemblies

    International Nuclear Information System (INIS)

    Nagano, Mamoru; Yoshioka, Ritsuo

    1983-01-01

    Purpose: To effectively utilize nuclear fuels by increasing the reactivity of a fuel assembly and reduce the concentration at the central region thereof upon completion of the burning. Constitution: A fuel assembly is bisected into a central region and a peripheral region by disposing an inner channel box within a channel box. The flow rate of coolants passing through the central region is made greater than that in the peripheral region. The concentration of uranium 235 of the fuel rods in the central region is made higher. In such a structure, since the moderating effect in the central region is improved, the reactivity of the fuel assembly is increased and the uranium concentration in the central region upon completion of the burning can be reduced, fuel economy and effective utilization of uranium can be attained. (Kamimura, M.)

  20. System for determining the local power in the fuel assembly of a nuclear reactor

    International Nuclear Information System (INIS)

    Rolstad, Erik; Korpas, T.-H.; Leyse, R.H.; Smith, R.D.

    1979-01-01

    System for determining the local power in the fuel assembly of a nuclear reactor which includes a rod conducting the heat and electricity, along which axial areas act as a gamma radiation thermometer. Each area includes a thermal bridge, a cold source and a pair of junctions acting as thermocouples so placed that they measure the temperature difference between the thermal bridge and the cold source. The power created by the fuel assembly near each area acting as gamma thermometer is found from this difference in temperature [fr

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

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  2. Fuel assembly

    International Nuclear Information System (INIS)

    Fujibayashi, Toru.

    1970-01-01

    Herein disclosed is a fuel assembly in which a fuel rod bundle is easily detachable by rotating a fuel rod fastener rotatably mounted to the upper surface of an upper tie-plate supporting a fuel bundle therebelow. A locking portion at the leading end of each fuel rod protrudes through the upper tie-plate and is engaged with or separated from the tie-plate by the rotation of the fastener. The removal of a desired fuel rod can therefore be remotely accomplished without the necessity of handling pawls, locking washers and nuts. (Owens, K.J.)

  3. Fuel assembly

    International Nuclear Information System (INIS)

    Yamazaki, Hajime.

    1995-01-01

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

  4. Device for supporting a fuel pin cluster within a nuclear reactor fuel assembly wrapper

    International Nuclear Information System (INIS)

    Marmonier, P.; Mesnage, B.; Teulon, J.; Vayra, J.; Venobre, H.

    1976-01-01

    A supporting member for an array of parallel rails each carrying one row of slidably mounted pins of a fuel cluster is placed coaxially at the lower end of a vertical fuel assembly wrapper. Each parallel rail is provided at each end with a downward extension and terminal lug which engages in a lateral groove formed in the periphery of the supporting member in order to lock and maintain the rails and the fuel pins in uniformly spaced relation within the fuel assembly wrapper. 10 claims, 8 figures

  5. Fuel assembly

    International Nuclear Information System (INIS)

    Watanabe, Shoichi; Hirano, Yasushi.

    1998-01-01

    A one-half or more of entire fuel rods in a fuel assembly comprises MOX fuel rods containing less than 1wt% of burnable poisons, and at least a portion of the burnable poisons comprises gadolinium. Then, surplus reactivity at an initial stage of operation cycle is controlled to eliminate burnable poisons remained unburnt at a final stage, as well as increase thermal reactivity. In addition, the content of fission plutonium is determined to greater than the content of uranium 235, and fuel rods at corner portions are made not to incorporate burnable poisons. Fuel rods not containing burnable poisons are disposed at positions in adjacent with fuel rods facing to a water rod at one or two directions. Local power at radial center of the fuel assembly is increased to flatten the distortion of radial power distribution. (N.H.)

  6. On the mixing model for calculating the temperature fields in nuclear reactor fuel assemblies

    International Nuclear Information System (INIS)

    Mikhin, V.I.; Zhukov, A.V.

    1985-01-01

    One of the alternatives of the mixing model applied for calculating temperature fields in nuclear reactor fuel assemblies,including the fuel assemblies with nonequilibrium energy-release in fuel element cross section, is consistently described. The equations for both constant and variable values of coolant density and heat capacity are obtained. The mixing model is based on a set of mass, heat and longitudinal momentum balance equations. This set is closed by the ratios connecting the unknown values for gaps between fuel elements with the averaged values for neighbouring channels. The ratios to close momentum and heat balance equations, explaining, in particular, the nonequivalent heat and mass, momentum and mass transfer coefficients, are suggested. The balance equations with variable coolant density and heat capacity are reduced to the form coinciding with those of the similar equations with constant values of these parameters. Application of one of the main ratios of the mixing model relating the coolant transverse overflow in the gaps between fuel elements to the averaged coolant rates (flow rates) in the neighbouring channels is mainly limited by the coolant stabilized flow in the fuel assemblies with regular symmetrical arrangement of elements. Mass transfer coefficients for these elements are experimentally determined. The ratio in the paper is also applicable for calculation of fuel assembly temperature fields with a small relative shift of elements

  7. Nuclear fuel assembly

    International Nuclear Information System (INIS)

    Anthony, A.J.

    1980-01-01

    A bimetallic spacer means is cooperatively associated with a nuclear fuel assembly and operative to resist the occurrence of in-reactor bowing of the nuclear fuel assembly. The bimetallic spacer means in one embodiment of the invention includes a space grid formed, at least principally, of zircaloy to the external surface of which are attached a plurality of stainless steel strips. In another embodiment the strips are attached to fuel pins. In each of the embodiments, the stainless steel strips during power production expand outwardly to a greater extent than do the members to which the stainless steel strips are attached, thereby forming stiff springs which abut against like bimetallic spacer means with which the other nuclear fuel assemblies are provided in a given nuclear reactor core to thus prevent the occurrence of in-reactor bowing of the nuclear fuel assemblies. (author)

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

    International Nuclear Information System (INIS)

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

    1982-01-01

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

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  10. Destruction of plutonium using non-uranium fuels in pressurized water reactor peripheral assemblies

    International Nuclear Information System (INIS)

    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 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 239 Pu and ≥90% 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

  11. Destruction of plutonium using non-uranium fuels in pressurized water reactor peripheral assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Chodak, III, Paul [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States)

    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 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% {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.

  12. Study of a fuel assembly for the nuclear reactor of IV generation cooled with supercritical water

    International Nuclear Information System (INIS)

    Barragan M, A.; Martin del Campo M, C.; Francois L, J. L.; Espinosa P, G.

    2011-11-01

    In this work a neutron study is presented about a square assembly design of double line of fuel rods, with moderator box to the center of the arrangement, for a nuclear reactor cooled with supercritical water (SCWR). The SCWR reactor was chosen by the characteristics of its design, mainly because is based in light water reactors (PWR and BWR), and the operational experience that has of them allow to use models and similar programs to simulate the fuel and the nucleus of this type of reactors. To develop the necessary models and to carry out the design and analysis of the SCWR reactor, the neutron codes MCNPX and Helios were used. The reason of using both codes, is because the code MCNPX used thoroughly in the neutron simulation of these reactors, it has been our reference code to analyze the results obtained with the Helios code which results are more efficient because its calculation times are minors. In the nucleus design the same parameters for both codes were considered. The results show that the design with Helios is a viable option to simulate these reactors since their values of the neutrons multiplication factor are very similar to those obtained with MCNPX. On the other hand, it could be corroborated that the CASMO-4 code is inadequate to simulate the fuel to the temperature conditions and water pressure in the SCWR. (Author)

  13. Fuel assemblies for use in FBR type reactors

    International Nuclear Information System (INIS)

    Nakahigashi, Shigeo; Terasawa, Michitaka.

    1984-01-01

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

  14. An analysis of fast reactor fuel assembly performance taking into account their mechanical interaction in the core and refuelling line capabilities

    International Nuclear Information System (INIS)

    Buksha, Yu.K.; Zabudko, L.M.; Kravchenko, I.N.; Matveenko, L.V.; Meshkov, M.N.

    1984-01-01

    An approach to assessment of fast reactor fuel assembly performance has been considered. A concept of passive restraint of fuel assemblies in a reactor adopted in the USSR is described. Some methods for calculating the interassembly interactions during operation are briefly outlined, some calculated results are presented. A problem of fuel assembly performance during refuelling taking into account the refuelling line capabilities is considered. Some results from fuel assemblies operation experience in the BN-600 reactor are given. (author)

  15. Fuel assemblies

    International Nuclear Information System (INIS)

    Echigoya, Hironori; Nomata, Terumitsu.

    1983-01-01

    Purpose: To render the axial distribution relatively flat. Constitution: First nuclear element comprises a fuel can made of zircalloy i.e., the metal with less neutron absorption, which is filled with a plurality of UO 2 pellets and sealed by using a lower end plug, a plenum spring and an upper end plug by means of welding. Second fuel element is formed by substituting a part of the UO 2 pellets with a water tube which is sealed with water and has a space for allowing the heat expansion. The nuclear fuel assembly is constituted by using the first and second fuel elements together. In such a structure, since water reflects neutrons and decrease their leakage to increase the temperature, reactivity is added at the upper portion of the fuel assembly to thereby flatten the axial power distribution. Accordingly, stable operation is possible only by means of deep control rods while requiring no shallow control rods. (Sekiya, K.)

  16. A thermal-hydraulic test rig for advanced fast reactor fuel assemblies

    International Nuclear Information System (INIS)

    Rapier, A.C.

    1989-03-01

    A new design of fast reactor fuel assemblies has been proposed in which the pins are supported in grids attached to the wrapper by flexible skirts. Coolant mixing is enhanced by the skirts diverting flow into the cluster of pins at each grid. There are insufficient empirical data available for the detailed design of the skirt or for the input to computer calculations of flow and heat transfer. A test rig to provide these data has been designed and built. (author)

  17. Ion irradiation studies of the origins of pressurized water reactor fuel assembly deformation

    International Nuclear Information System (INIS)

    Hengstler-Eger, Rosmarie Martina

    2012-01-01

    The presented thesis studies ion irradiation damage in Zr-based alloys for pressurized water reactors to explain the origins of unexpectedly high fuel assembly growth in some plants. Transmission electron microscopy was used to investigate the effects of temperature, dose, hydrogen content of the alloy and tensile stress. A clear correlation between the stress orientation towards the crystal lattice and the density of the dislocation loops which are responsible for increased growth was found.

  18. Nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    Hindle, E.D.

    1981-01-01

    An array of rods comprising zirconium alloy sheathed nuclear fuel pellets assembled to form a fuel element for a pressurised water reactor is claimed. The helium gas pressure within each rod differs substantially from that of its closest neighbours

  19. Fuel assembly

    International Nuclear Information System (INIS)

    Nomata, Terumitsu.

    1993-01-01

    Among fuel pellets to be loaded to fuel cans of a fuel assembly, fuel pellets having a small thermal power are charged in a region from the end of each of spacers up to about 50mm on the upstream of coolants that flow vertically at the periphery of fuel rods. Coolants at the periphery of fuel rods are heated by the heat generation, to result in voids. However, since cooling effect on the upstream of the spacers is low due to influences of the spacers. Further, since the fuel pellets disposed in the upstream region have small thermal power, a void coefficient is not increased. Even if a thermal power exceeding cooling performance should be generated, there is no worry of causing burnout in the upstream region. Even if burnout should be caused, safety margin and reliability relative to burnout are improved, to increase an allowable thermal power, thereby enabling to improve integrity and reliability of fuel rods and fuel assemblies. (N.H.)

  20. Spacing grids for a fuel pencil bundle in a nuclear reactor assembly

    International Nuclear Information System (INIS)

    Feutrel, Claude.

    1977-01-01

    This invention relates to the lattices forming the spacing of a bundle of clad fuel pencils in a nuclear reactor assembly, particularly in a water cooled or fast reactor, the purpose of such lattices being to maintain these pencils parallel with respect to each other and according to a given lattice arrangement, whilst also providing these pencils with a flexible support according to different successive areas apportioned with their length in order to present them from vibrating under the effect of the circulation of a liquid coolant environment flowing in contact with these pencils [fr

  1. The application of neural networks for optimization of the configuration of fuel assemblies in PWR reactors

    International Nuclear Information System (INIS)

    Sadighi, M.; Setayeshi, S.; Salehi, A.A.

    2002-01-01

    This paper presents a new method to solve the problem of finding the best configuration of fuel assemblies in a PWR (Pressurized Water Reactor) core. Finding an optimum solution requires a huge amount of calculations in classical methods. It has been shown that the application of continuous Hop field neural network accompanied by the Simulated Annealing method to this problem not only reduces the volume of the calculations, but also guarantees finding the best solution. In this study flattening of neutron flux inside the reactor core of Brusher NPP is considered as an objective function. The result shows the optimum core configuration which is in agreement with the pattern proposed by the designer

  2. Experimental studies of flow induced vibrations of the fuel assembly for the PEC reactor

    International Nuclear Information System (INIS)

    Pitimada, D.; Presaghi, M.; Tampone, O.; Cesari, F.

    1977-01-01

    The vibration behaviour of an assembly of seven mock-up fuel bundles of PEC reactor has been investigated. The assembly was excited by a parallel flow of water simulating sodium. The motion of the group (or of a single bundle in the group) has been measured in transverse sections detecting two orthogonal components of displacement. During the experiences the following parameters were varied: bundle foot and pads restraints, flow rate condition, coolant flow outlet conditions at the head of fuel bundles. Experimental data were processed in order to obtain: trajectories of three points of fuel bundle axis, power density spectra of measured vibration amplitudes, correlations between coolant flow rate and vibration amplitude R.M.S. (author)

  3. Artificial neural networks for spatial distribution of fuel assemblies in reload of PWR reactors

    International Nuclear Information System (INIS)

    Oliveira, Edyene; Castro, Victor F.; Velásquez, Carlos E.; Pereira, Claubia

    2017-01-01

    An artificial neural network methodology is being developed in order to find an optimum spatial distribution of the fuel assemblies in a nuclear reactor core during reload. The main bounding parameter of the modelling was the neutron multiplication factor, k ef f . The characteristics of the network are defined by the nuclear parameters: cycle, burnup, enrichment, fuel type, and average power peak of each element. These parameters were obtained by the ORNL nuclear code package SCALE6.0. As for the artificial neural network, the ANN Feedforward Multi L ayer P erceptron with various layers and neurons were constructed. Three algorithms were used and tested: LM (Levenberg-Marquardt), SCG (Scaled Conjugate Gradient) and BayR (Bayesian Regularization). Artificial neural network have implemented using MATLAB 2015a version. As preliminary results, the spatial distribution of the fuel assemblies in the core using a neural network was slightly better than the standard core. (author)

  4. Fuel Assembly Damping Summary

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-10-15

    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

  5. Determination of power density distribution of fuel assemblies for research reactor by directly measuring the strontium-91 activities

    International Nuclear Information System (INIS)

    Yuan, Liq-Ji

    1987-01-01

    This work described the investigations of reactor core power peaking and three dimensional power density distribution of present core configuration of Tsing Hua Open-pool reactor (THOR). An experimental program, based on non-destructive fuel gamma scanning of 91 Sr activities, provides the data of fission density distribution for individual fuel pin of four-rod TRIGA-LEU cluster or for MTR-type fuel assembly. The informations are essentially important for the safety of reactor operation and for fuel management especially for the mixed loading with three different types of fuel at present. The relative power peaking values and the power density distribution for present core are discussed. (author)

  6. Human factors and safety issues associated with actinide retrieval from spent light water reactor fuel assemblies

    International Nuclear Information System (INIS)

    Spelt, P.F.

    1992-01-01

    A major problem in environmental restoration and waste management is the disposition of used fuel assemblies from the many light water reactors in the United States, which present a radiation hazard to those whose job is to dispose of them, with a similar threat to the general environment associated with long-term storage in fuel repositories around the country. Actinides resident in the fuel pins as a result of their use in reactor cores constitute a significant component of this hazard. Recently, the Department of Energy has initiated an Actinide Recycle Program to study the feasibility of using pyrochemical (molten salt) processes to recover actinides from the spent fuel assemblies of commercial reactors. This project concerns the application of robotics technology to the operation and maintenance functions of a plant whose objective is to recover actinides from spent fuel assemblies, and to dispose of the resulting hardware and chemical components from this process. Such a procedure involves a number of safety and human factors issues. The purpose of the project is to explore the use of robotics and artificial intelligence to facilitate accomplishment of the program goals while maintaining the safety of the humans doing the work and the integrity of the environment. This project will result in a graphic simulation on a Silicon Graphics workstation as a proof of principle demonstration of the feasibility of using robotics along with an intelligent operator interface. A major component of the operator-system interface is a hybrid artificial intelligence system developed at Oak Ridge National Laboratory, which combines artificial neural networks and an expert system into a hybrid, self-improving computer-based system interface. 10 refs

  7. Reactor fuel charging equipment

    International Nuclear Information System (INIS)

    Wade, Elman.

    1977-01-01

    In many types of reactor fuel charging equipment, tongs or a grab, attached to a trolley, housed in a guide duct, can be used for withdrawing from the core a selected spent fuel assembly or to place a new fuel assembly in the core. In these facilities, the trolley may have wheels that roll on rails in the guide duct. This ensures the correct alignment of the grab, the trolley and fuel assembly when this fuel assembly is being moved. By raising or lowering such a fuel assembly, the trolley can be immerged in the coolant bath of the reactor, whereas at other times it can be at a certain level above the upper surface of the coolant bath. The main object of the invention is to create a fuel handling apparatus for a sodium cooled reactor with bearings lubricated by the sodium coolant and in which the contamination of these bearings is prevented [fr

  8. One approach to accepting and transporting spent fuel from early-generation reactors with short fuel assemblies

    International Nuclear Information System (INIS)

    Peterson, R.W.; Bentz, E.J. Jr.; Bentz, C.B.

    1993-01-01

    In the early days of development of commercial nuclear power reactors in the U.S., the overall length and uranium loading of the fuel assemblies were considerably less than those of later generation facilities. In turn, some of these early facilities were designed for handling shorter casks than currently-certified casks. The spent fuel assemblies from these facilities are nearly all standard fuel within the definition in the Standard Contract (10 CFR 961) between the utilities and the U.S. Department of Energy (DOE) (the Big Rock Point fuel cross-section is outside the standard fuel dimension), and the utilities involved hold early delivery rights under DOE's oldest-fuel-first (OFF) allocation scenario. However, development of casks suitable for satisfying the acceptance and transportation requirements of some of these facilities is not currently underway in the DOE Cask System Development Program (CSDP). While the total MTU of these fuels is relatively small compared to the total program, the number of assemblies to be transported is significant, especially in the early years of operation according to the OFF allocation scenario. We therefore perceive a need for DOE to develop an approach and to implement plans to satisfy the unique acceptance and transportation requirements of these facilities. One such approach is outlined below. (author)

  9. Fuel assembly reconstitution

    International Nuclear Information System (INIS)

    Morgado, Mario M.; Oliveira, Monica G.N.; Ferreira Junior, Decio B.M.; Santos, Barbara O. dos; Santos, Jorge E. dos

    2009-01-01

    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)

  10. Optimization of the fuel assembly for the Canadian SuperCritical Water-cooled Reactor (SCWR)

    Energy Technology Data Exchange (ETDEWEB)

    French, C., E-mail: Corey.French@cnsc-ccsn.gc.ca [Canadian Nuclear Safety Commission, Ottawa, Ontario (Canada); Bonin, H.; Chan, P.K. [Royal Military College of Ontario, Kingston, Ontario (Canada)

    2013-07-01

    An approach to develop a parametric optimization tool to support the Canadian Supercritical Water-cooled Reactor (SCWR) fuel design is presented in this work. The 2D benchmark lattices for 78-pin and 64-pin fuel assemblies are used as the initial models from which fuel performance and subsequent optimization stem from. A tandem optimization procedure is integrated which employs the steepest descent method. The physics codes WIMS-AECL, MCNP6 and SERPENT are used to calculate and verify select performance factors. The results are used as inputs to an optimization algorithm that yield optimal fresh fuel isotopic composition and lattice geometry. Preliminary results on verifications of infinite lattice reactivity are demonstrated in this paper. (author)

  11. Fuel sub-assembly

    International Nuclear Information System (INIS)

    Jolly, R.

    1982-01-01

    A fuel sub-assembly for a liquid metal cooled nuclear reactor is described in which the bundle of fuel pins are braced apart by a series of spaced grids. The grids at the lower end are capable of yielding, thus allowing pins swollen by irradiation to be withdrawn with a reduced risk of damage. (U.K.)

  12. Experimental investigation of critical velocity in a parallel plate research reactor fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Castro, Alfredo J.A.; Scuro, Nikolas L.; Andrade, Delvonei A., E-mail: ajcastro@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNE-SP), Sao Paulo, SP (Brazil)

    2017-07-01

    The fuel elements of a MTR (Material Testing Reactor) type nuclear reactor are mostly composed of aluminum coated fuel plates containing the core of uranium silica (U{sub 3}Si{sub 2}) dispersed in an aluminum matrix. These plates have a thickness of the order of millimeters and are much longer in relation to their thickness. They are arranged in parallel in the assembly of the fuel element to form channels between them a few millimeters in thickness, through which there is a flow of the coolant. This configuration, combined with the need for a flow at high flow rates to ensure the cooling of the fuel element in operation, may create problems of mechanical failure of fuel plate due to the vibration induced by the flow in the channels. In the case of critical velocity excessive permanent deflections of the plates can cause blockage of the flow channel in the reactor core and lead to overheating in the plates. For this study an experimental bench capable of high volume flows and a test section that simulates a plate-like fuel element with three cooling channels were developed. The dimensions of the test section were based on the dimensions of the Fuel Element of the Brazilian Multipurpose Reactor (RMB), whose project is being coordinated by the National Commission of Nuclear Energy (CNEN). The experiments performed attained the objective of reaching Miller's critical velocity condition. The critical velocity was reached with 14.5 m/s leading to the consequent plastic deformation of the flow channel plates. (author)

  13. Nuclear fuel assembly

    International Nuclear Information System (INIS)

    Betten, P.R.

    1976-01-01

    Under the invention the fuel assembly is particularly suitable for liquid metal cooled fast neutron breeder reactors. Hence, according to the invention a fuel assembly cladding includes inward corrugations with respect to the remainder of the cladding according to a recurring pattern determined by the pitch of the metal wire helically wound round the fuel rods of the assembly. The parts of the cladding pressed inwards correspond to the areas in which the wire encircling the peripheral fuel rods is generally located apart from the cladding, thereby reducing the play between the cladding and the peripheral fuel rods situated in these areas. The reduction in the play in turn improves the coolant flow in the internal secondary channels of the fuel assembly to the detriment of the flow in the peripheral secondary channels and thereby establishes a better coolant fluid temperature profile [fr

  14. Nuclear reactor spacer assembly

    International Nuclear Information System (INIS)

    Anthony, A.J.; Groves, M.D.

    1979-01-01

    A fuel assembly for a nuclear reactor is disclosed wherein the fuel element receiving and supporting grid is comprised of a first metal, the guide tubes which pass through the grid assembly are comprised of a second metal and the grid is supported on the guide tubes by means of expanded sleeves located intermediate the grid and guide tubes. The fuel assembly is fabricated by inserting the sleeves, of initial outer diameter commensurate with the guide tube outer diameters, through the holes in the grid assembly provided for the guide tubes and thereafter expanding the sleeves radially outwardly along their entire length such that the guide tubes can subsequently be passed through the sleeves. The step of radial expansion, as a result of windows provided in the sleeves having dimensions commensurate with the geometry of the grid, mechanically captures the grid and simultaneously preloads the sleeve against the grid whereby relative motion between the grid and guide tube will be precluded

  15. Fuel assembly

    International Nuclear Information System (INIS)

    Ishibashi, Yoko; Aoyama, Motoo; Oyama, Jun-ichi.

    1995-01-01

    Burnable poison-incorporating fuel rods of a first group are disposed in a region in adjacent with a water rod having a large diameter (neutron moderator rod) disposed to the central portion of a fuel assembly. Burnable poison-incorporating fuel rods of a second group are disposed to a region other than peripheral zone in adjacent with a channel box and corners positioned at an inner zone, in adjacent with the channel box. The average concentration of burnable poisons of the burnable poison-incorporating fuel rods of the first group is made greater than that of the second group. With such a constitution, when the burnable poisons of the first group are burnt out, the burnable poisons of the second group are also burnt out at the same time. Accordingly, an amount of burnable poisons left unburnt at the final stage of the operation cycle is reduced, to improve the reactivity. This can improve the economical property. (I.N.)

  16. Apparatus for adjusting the elevation of fuel rods in a nuclear reactor fuel assembly

    International Nuclear Information System (INIS)

    Hale, D.L.; Culbreth, T.F.

    1988-01-01

    A tool adapted for adjusting the level of a nuclear fuel rod in a fuel assembly is described comprising: an expander comprising two elongate generally parallel and laterally spaced apart arms extending in a longitudinal direction, an actuator operatively mounted to the expander so as to be disposed between the two arms and being movable with respect thereto, and cooperating surface means mounted to the arms and the actuator for laterally separating the free ends of the arms to a predetermined maximum distance upon movement of the actuator with respect to the arms

  17. Nuclear fuel assemblies

    International Nuclear Information System (INIS)

    Butterfield, R.S.; Garner, D.L.M.

    1977-01-01

    Reference is made to nuclear fuel assemblies designed for cooling on the 'tube-in-shell' principle in which the fuel is contained by a shell and is cooled by coolant passed through tubes extending through the shell. It has been proposed to employ coated particle fuel as a porous bed on the tube side and the bleed coolant from the tubes into direct contact with the fuel particles. In this way heat is extracted both by direct contact with the fuel and by heat transfer through the coolant tube walls. The system described aims to provide an improved structure of tube and shell for a fuel assembly of this kind and is particularly suitable for use in a gas cooled fast reactor, being able to withstand the neutron flux and high temperature conditions in these reactors. Constructional details are given. (U.K.)

  18. Nuclear fuel assemblies

    International Nuclear Information System (INIS)

    Natori, Hisahide; Kurihara, Kunitoshi.

    1982-01-01

    Purpose: To increase the fuel safety by decreasing the gap conductance between fuels and cladding tubes, as well as improve the reactor core controllability by rendering the void coefficient negative. Constitution: Fuel assemblies in a pressure tube comprise a tie-rod, fuel rods in a central region, and fuel rods with burnable poison in the outer circumference region. Here, B 4 C is used as the burnable poison by 1.17 % by weight ratio. The degrees of enrichment for the fissile plutonium as PuO 2 -UO 2 fuel used in the assemblies are 2.7 %, 2.7 % and 1.5 % respectively in the innermost layer, the intermediate layer and the outermost layer. This increases the burn-up degree to improve the plant utilizability, whereby the void coefficient is rendered negative to improve the reactor core controllability. (Horiuchi, T.)

  19. Thermal-hydraulic calculations for a fuel assembly in a European Pressurized Reactor using the RELAP5 code

    Directory of Open Access Journals (Sweden)

    Skrzypek Maciej

    2015-09-01

    Full Text Available The main object of interest was a typical fuel assembly, which constitutes a core of the nuclear reactor. The aim of the paper is to describe the phenomena and calculate thermal-hydraulic characteristic parameters in the fuel assembly for a European Pressurized Reactor (EPR. To perform thermal-hydraulic calculations, the RELAP5 code was used. This code allows to simulate steady and transient states for reactor applications. It is also an appropriate calculation tool in the event of a loss-of-coolant accident in light water reactors. The fuel assembly model with nodalization in the RELAP5 (Reactor Excursion and Leak Analysis Program code was presented. The calculations of two steady states for the fuel assembly were performed: the nominal steady-state conditions and the coolant flow rate decreased to 60% of the nominal EPR flow rate. The calculation for one transient state for a linearly decreasing flow rate of coolant was simulated until a new level was stabilized and SCRAM occurred. To check the correctness of the obtained results, the authors compared them against the reactor technical documentation available in the bibliography. The obtained results concerning steady states nearly match the design data. The hypothetical transient showed the importance of the need for correct cooling in the reactor during occurrences exceeding normal operation. The performed analysis indicated consequences of the coolant flow rate limitations during the reactor operation.

  20. Structural behaviour of fuel assemblies for water cooled reactors. Proceedings of a technical meeting

    International Nuclear Information System (INIS)

    2005-07-01

    At the invitation of the Government of France and in response to a proposal of the IAEA Technical Working Group on Water Reactor Fuel Performance and Technology (TWGFPT), the IAEA convened a Technical Meeting on Fuel Assembly Structural Behaviour in Cadarache, France, from 22 to 26 November 2004. The meeting was hosted by the CEA Cadarache Centre, AREVA Framatome-ANP and Electricite de France. The meeting aimed to provide in depth technical exchanges on PWR and WWER operational experience in the field of fuel assembly mechanical behaviour and the potential impact of future high burnup fuel management on fuel reliability. It addressed in-service experience and remedial solutions, loop testing experience, qualification and damage assessment methods (analytic or experimental ones), mechanical behaviour of the fuel assembly including dynamic and fluid structure interaction aspects, modelling and numerical analysis methods, and impact of the in-service evolution of the structural materials. Sixty-seven participants from 17 countries presented 30 papers in the course of four sessions. The topics covered included the impact of hydraulic loadings on fuel assembly (FA)performance, FA bow and control rod (CR) drop kinetics, vibrations and rod-to-grid wear and fretting, and, finally, evaluation and modelling of accident conditions, mainly from seismic causes. FA bow, CR drop kinetics and hydraulics are of great importance under conditions of higher fuel duties including burnup increase, thermal uprates and longer fuel cycles. Vibrations and rod-to-grid wear and fretting have been identified as a key cause of fuel failure at PWRs during the past several years. The meeting demonstrated that full-scale hydraulic tests and modelling provide sufficient information to develop remedies to increase FA skeleton resistance to hydraulic loads, including seismic ones, vibrations and wear. These proceedings are presented as a book with an attached CD-ROM. The first part of the CD

  1. Nuclear fuel assembly

    International Nuclear Information System (INIS)

    Hayashi, Hiroshi; Watari, Yoshio; Hizahara, Hiroshi; Masuoka, Ryuzo.

    1970-01-01

    When exchanging nuclear fuel assemblies during the operation of a nuclear reactor, melting of fuel bodies, and severence of tubular claddings is halted at the time of insertion by furnishing a neutron absorbing material such as B 10 , Cd, Gd or the like at the forward end of the fuel assembly to thereby lower the power peak at the forward ends of the fuel elements to within tolerable levels and thus prevent both fuel liquification and excessive expansion. The neutron absorbing material may be attached in the form of a plate to the fuel assembly forward tie plate, or may be inserted as a pellet into the front end of the tubular cladding. (Owens, K.J.)

  2. A porous medium model for predicting the duct wall temperature of sodium fast reactor fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Yu, Yiqi, E-mail: yyu@anl.gov [Nuclear Engineering Division, Argonne National Laboratory, Lemont, IL 60439 (United States); Merzari, Elia; Obabko, Aleksandr [Mathematics and Computer Science Division, Argonne National Laboratory, Lemont, IL 60439 (United States); Thomas, Justin [Nuclear Engineering Division, Argonne National Laboratory, Lemont, IL 60439 (United States)

    2015-12-15

    Highlights: • The proposed models are 400 times less computationally expensive than CFD simulations. • The proposed models show good duct wall temperature agreement with CFD simulations. • The paper provides an efficient tool for coupled radial core expansion calculation. - Abstract: Porous medium models have been established for predicting duct wall temperature of sodium fast reactor rod bundle assembly, which is much less computationally expensive than conventional CFD simulations that explicitly represent the wire-wrap and fuel pin geometry. Three porous medium models are proposed in this paper. Porous medium model 1 takes the whole assembly as one porous medium of uniform characteristics in the conventional approach. Porous medium model 2 distinguishes the pins along the assembly's edge from those in the interior with two distinct regions, each with a distinct porosity, resistance, and volumetric heat source. This accounts for the different fuel-to-coolant volume ratio in the two regions, which is important for predicting the temperature of the assembly's exterior duct wall. In Porous medium model 3, a precise resistance distribution was employed to define the characteristic of the porous medium. The results show that both porous medium model 2 and 3 can capture the average duct wall temperature well. Furthermore, the local duct wall variations due to different sub-channel patterns in bare rod bundles are well captured by porous medium model 3, although the wire effect on the duct wall temperature in wire wrap rod bundle has not been fully reproduced yet.

  3. Fuel assembly

    International Nuclear Information System (INIS)

    Sano, Hiroki; Fushimi, Atsushi; Tominaga, Kenji; Aoyama, Motoo; Ishii, Kazuya.

    1997-01-01

    In burnable poison-incorporated uranium fuels of a BWR type reactor, the compositional ratio of isotopes of the burnable poisons is changed so as to increase the amount of those having a large neutron absorbing cross sectional area. For example, if the ratio of Gd-157 at the same burnable poison enrichment degree is made greater than the natural ratio, this gives the same effect as the increase of the enrichment degree per one fuel rod, thereby providing an effect of reducing a surplus reactivity. Gadolinium, hafnium and europium as burnable poisons have an absorbing cross sectional area being greater in odd numbered nuclei than in even numbered nuclei, on the contrary, boron has a cross section being greater in even numbered nucleus than odd numbered nuclei. Accordingly, if the ratio of isotopes having greater cross section at the same burnable poison enrichment degree is made greater than the natural ratio, surplus reactivity at the initial stage of the burning can be reduced without greatly increasing the amount of burnable poison-incorporated uranium fuels, fuel loading amount is not reduced and the fuel economy is not worsened. (N.H.)

  4. Benchmark exercise for fluid flow simulations in a liquid metal fast reactor fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Merzari, E., E-mail: emerzari@anl.gov [Mathematics and Computer Science Division, Argonne National Laboratory, 9700 S. Cass Avenue, Lemont, IL 60439 (United States); Fischer, P. [Mathematics and Computer Science Division, Argonne National Laboratory, 9700 S. Cass Avenue, Lemont, IL 60439 (United States); Yuan, H. [Nuclear Engineering Division, Argonne National Laboratory, Lemont, IL (United States); Van Tichelen, K.; Keijers, S. [SCK-CEN, Boeretang 200, Mol (Belgium); De Ridder, J.; Degroote, J.; Vierendeels, J. [Ghent University, Ghent (Belgium); Doolaard, H.; Gopala, V.R.; Roelofs, F. [NRG, Petten (Netherlands)

    2016-03-15

    Highlights: • A EUROTAM-US INERI consortium has performed a benchmark exercise related to fast reactor assembly simulations. • LES calculations for a wire-wrapped rod bundle are compared with RANS calculations. • Results show good agreement for velocity and cross flows. - Abstract: As part of a U.S. Department of Energy International Nuclear Energy Research Initiative (I-NERI), Argonne National Laboratory (Argonne) is collaborating with the Dutch Nuclear Research and consultancy Group (NRG), the Belgian Nuclear Research Centre (SCK·CEN), and Ghent University (UGent) in Belgium to perform and compare a series of fuel-pin-bundle calculations representative of a fast reactor core. A wire-wrapped fuel bundle is a complex configuration for which little data is available for verification and validation of new simulation tools. UGent and NRG performed their simulations with commercially available computational fluid dynamics (CFD) codes. The high-fidelity Argonne large-eddy simulations were performed with Nek5000, used for CFD in the Simulation-based High-efficiency Advanced Reactor Prototyping (SHARP) suite. SHARP is a versatile tool that is being developed to model the core of a wide variety of reactor types under various scenarios. It is intended both to serve as a surrogate for physical experiments and to provide insight into experimental results. Comparison of the results obtained by the different participants with the reference Nek5000 results shows good agreement, especially for the cross-flow data. The comparison also helps highlight issues with current modeling approaches. The results of the study will be valuable in the design and licensing process of MYRRHA, a flexible fast research reactor under design at SCK·CEN that features wire-wrapped fuel bundles cooled by lead-bismuth eutectic.

  5. Development of numerical simulation system for thermal-hydraulic analysis in fuel assembly of sodium-cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ohshima, Hiroyuki; Uwaba, Tomoyuki [Japan Atomic Energy Agency (4002 Narita, O-arai, Ibaraki 311-1393, Japan) (Japan); Hashimoto, Akihiko; Imai, Yasutomo [NDD Corporation (1-1-6 Jounan, Mito, Ibaraki 310-0803, Japan) (Japan); Ito, Masahiro [NESI Inc. (4002 Narita, O-arai, Ibaraki 311-1393, Japan) (Japan)

    2015-12-31

    A numerical simulation system, which consists of a deformation analysis program and three kinds of thermal-hydraulics analysis programs, is being developed in Japan Atomic Energy Agency in order to offer methodologies to clarify thermal-hydraulic phenomena in fuel assemblies of sodium-cooled fast reactors under various operating conditions. This paper gives the outline of the system and its applications to fuel assembly analyses as a validation study.

  6. Fuel assembly

    International Nuclear Information System (INIS)

    Ueda, Makoto.

    1991-01-01

    In a fuel assembly in which spectral shift type moderator guide members are arranged, the moderator guide member has a flow channel resistance member, that provides flow resistance against the moderators, in the upstream of a moderator flowing channel, by which the ratio of removing coolants is set greater at the upstream than downstream. With such a constitution, the void distribution increasing upward in the channel box except for the portion of the moderator guide member is moderated by the increase of the area of the void region that expands downward in the guide member. Accordingly, the axial power distribution is flattened throughout the operation cycle and excess distortion is eliminated to improve the fuel integrity. (T.M.)

  7. Innovative technologies on fuel assemblies cleaning for sodium fast reactors: First considerations on cleaning process

    International Nuclear Information System (INIS)

    Simon, N.; Lorcet, H.; Beauchamp, F.; Guigues, E.; Lovera, P.; Fleche, J. L.; Lacroix, M.; Carra, O.; Dechelette, F.; Prele, G.; Rodriguez, G.

    2012-01-01

    Within the framework of Sodium Fast Reactor development, innovative fuel assembly cleaning operations are investigated to meet the GEN IV goals of safety and of process development. One of the challenges is to mitigate the Sodium Water Reaction currently used in these processes. The potential applications of aqueous solutions of mineral salts (including the possibility of using redox chemical reactions) to mitigate the Sodium Water Reaction are considered in a first part and a new experimental bench, dedicated to this study, is described. Anhydrous alternative options based on Na/CO 2 interaction are also presented. Then, in a second part, a functional study conducted on the cleaning pit is proposed. Based on experimental feedback, some calculations are carried out to estimate the sodium inventory on the fuel elements, and physical methods like hot inert gas sweeping to reduce this inventory are also presented. Finally, the implementation of these innovative solutions in cleaning pits is studied in regard to the expected performances. (authors)

  8. Evaluation of neutron streaming in fast breeder reactor fuel assembly by double heterogeneous modelling

    International Nuclear Information System (INIS)

    Unesaki, Hironobu; Takeda, Toshikazu

    1988-01-01

    Neutron streaming in a fast breeder reactor fuel assembly caused by the double heterogeneity structure is estimated by double heterogeneous modelling. The conventional pin cell model, a two-region subassembly model and the exact pin cluster model are used to take into account the streaming effect caused by the pin cell structure and the surrounding wrapper tube structure. The heterogeneity of wrapper tube and its surrounding sodium is explicitly considered. The streaming effect is evaluated based on Benoist's diffusion coefficient. The total streaming effect caused by the double heterogeneity structure of a fuel subassembly is found to be -0.2 % dk/kk' for k eff , which is almost twice that obtained from the conventional pin cell model of -0.1 % dk/kk'. (author)

  9. Numerical investigation of a heat transfer within the prismatic fuel assembly of a very high temperature reactor

    International Nuclear Information System (INIS)

    Tak, Nam-il; Kim, Min-Hwan; Lee, Won Jae

    2008-01-01

    The complex geometry of the hexagonal fuel blocks of the prismatic fuel assembly in a very high temperature reactor (VHTR) hinders accurate evaluations of the temperature profile within the fuel assembly without elaborate numerical calculations. Therefore, simplified models such as a unit cell model have been widely applied for the analyses and designs of prismatic VHTRs since they have been considered as effective approaches reducing the computational efforts. In a prismatic VHTR, however, the simplified models cannot consider a heat transfer within a fuel assembly as well as a coolant flow through a bypass gap between the fuel assemblies, which may significantly affect the maximum fuel temperature. In this paper, a three-dimensional computational fluid dynamics (CFD) analysis has been carried out on a typical fuel assembly of a prismatic VHTR. Thermal behaviours and heat transfer within the fuel assembly are intensively investigated using the CFD solutions. In addition, the accuracy of the unit cell approach is assessed against the CFD solutions. Two example situations are illustrated to demonstrate the deficiency of the unit cell model caused by neglecting the effects of the bypass gap flow and the radial power distribution within the fuel assembly

  10. Method of transporting fuel assemblies

    International Nuclear Information System (INIS)

    Okada, Katsutoshi.

    1979-01-01

    Purpose: To enable safety transportation of fuel assemblies for FBR type reactors by surrounding each of fuel elements in a wrapper tube by a rubbery, hollow cylindrical container and by sealing medium such as air to the inside of the container. Method: A fuel element is contained in a hollow cylindrical rubber-like tube. The fuel element has an upper end plug, a lower end plug and a wire spirally wound around the outer periphery. Upon transportation of the fuel assemblies, each of the fuel elements is covered with the container and arranged in the wrapper tube and then the fuel assemblies are assembled. Then, medium such as air is sealed for each of the fuel elements by way of an opening and then the opening is tightly closed. Before loading the transported fuel assemblies in the reactor, the medium is discharged through the opening and the container is completely extracted and removed from the inside of the wrapper tube. (Seki, T.)

  11. Artificial neural networks for spatial distribution of fuel assemblies in reload of PWR reactors

    Energy Technology Data Exchange (ETDEWEB)

    Oliveira, Edyene; Castro, Victor F.; Velásquez, Carlos E.; Pereira, Claubia, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Programa de Pós-Graduação em Ciências e Técnicas Nucleares

    2017-07-01

    An artificial neural network methodology is being developed in order to find an optimum spatial distribution of the fuel assemblies in a nuclear reactor core during reload. The main bounding parameter of the modelling was the neutron multiplication factor, k{sub ef{sub f}}. The characteristics of the network are defined by the nuclear parameters: cycle, burnup, enrichment, fuel type, and average power peak of each element. These parameters were obtained by the ORNL nuclear code package SCALE6.0. As for the artificial neural network, the ANN Feedforward Multi{sub L}ayer{sub P}erceptron with various layers and neurons were constructed. Three algorithms were used and tested: LM (Levenberg-Marquardt), SCG (Scaled Conjugate Gradient) and BayR (Bayesian Regularization). Artificial neural network have implemented using MATLAB 2015a version. As preliminary results, the spatial distribution of the fuel assemblies in the core using a neural network was slightly better than the standard core. (author)

  12. Some Windscale experience of the underwater examination of water reactor fuel assemblies

    International Nuclear Information System (INIS)

    Banks, D.A.; Prestwood, J.; Stuttard, A.

    1981-01-01

    Windscale Nuclear Laboratories have been involved in the underwater post irradiation examination of irradiated water reactor fuel since the early 1970's. Since the work of the laboratories covers a wide range of fuel types, the equipment has had to be capable of handling any design, long or short, circular or square. There has so far been no element of routine work in the tasks performed at Windscale, for in this period fuel assemblies from 9 LWR's and WSGHWR have been examined successfully. Individual jobs have ranged from visual examination which may be carried out at several magnifications, to the complete breakdown of a PWR assembly to its separate rods and grids. Between these limits rod bow and rod diameter have been measured, rod withdrawal forces determined, and eddy current test methods devised. Cutting equipment has been used for a variety of dismantling tasks, and underwater cameras have been employed for monochrome and colour photography, using standard and macro lenses. The equipment is described. (author)

  13. Nuclear reactor core assembly

    International Nuclear Information System (INIS)

    Baxi, C.B.

    1978-01-01

    The object of the present invention is to provide a fast reactor core assembly design for use with a fluid coolant such as liquid sodium or carbon monoxide incorporating a method of increasing the percentage of coolant flow though the blanket elements relative to the total coolant flow through the blanket and fuel elements during shutdown conditions without using moving parts. It is claimed that deterioration due to reactor radiation or temperature conditions is avoided and ready modification or replacement is possible. (U.K.)

  14. Verification of spectral burn-up codes on 2D fuel assemblies of the GFR demonstrator ALLEGRO reactor

    International Nuclear Information System (INIS)

    Čerba, Štefan; Vrban, Branislav; Lüley, Jakub; Dařílek, Petr; Zajac, Radoslav; Nečas, Vladimír; Haščik, Ján

    2014-01-01

    Highlights: • Verification of the MCNPX, HELIOS and SCALE codes. • MOX and ceramic fuel assembly. • Gas-cooled fast reactor. • Burnup calculation. - Abstract: The gas-cooled fast reactor, which is one of the six GEN IV reactor concepts, is characterized by high operational temperatures and a hard neutron spectrum. The utilization of commonly used spectral codes, developed mainly for LWR reactors operated in the thermal/epithermal neutron spectrum, may be connected with systematic deviations since the main development effort of these codes has been focused on the thermal part of the neutron spectrum. To be able to carry out proper calculations for fast systems the used codes have to account for neutron resonances including the self-shielding effect. The presented study aims at verifying the spectral HELIOS, MCNPX and SCALE codes on the basis of depletion calculations of 2D MOX and ceramic fuel assemblies of the ALLEGRO gas-cooled fast reactor demonstrator in infinite lattice

  15. Method of reactor fueling

    International Nuclear Information System (INIS)

    Saito, Toshiro.

    1983-01-01

    Purpose: To decrease the cost and shorten the working time by saving fueling neutron detectors and their components. Method: Incore drive tubes for the neutron source range monitor (SRM) and intermediate range monitor (IRM) are disposed respectively within in a reactor core and a SRM detector assembly is inserted to the IRM incore drive tube which is most nearest to the neutron source upon reactor fueling. The reactor core reactivity is monitored by the SRM detector assembly. The SRM detector asesembly inserted into the IRM drive tube is extracted at the time of charging fuels up to the frame connecting the SRM and, thereafter, IRM detection assembly is inserted into the IRM drive tube and the SRM detector assembly is inserted into the SRM drive tube respectively for monitoring the reactor core. (Sekiya, K.)

  16. Mechanical fragmentation of nuclear reactor fuel assemblies by the double cutting method

    International Nuclear Information System (INIS)

    Voitsekhovskii, B.V.; Istomin, V.L.; Mitrofanov, V.V.

    1995-01-01

    A method is described for cutting a spent fuel assembly with straight shears into pieces of a prescribed size. The method does not require separation of the casing and the lattices. The double cutting method is briefly described, and experiments designed for cutting BN-350 and VVER-440 fuel assemblies are outlined. The testing showed that the cutting method was suitable for mechanical polarization of fuel assemblies. The investigations led to the development of turnkey industrial equipment for cutting spent fuel assemblies of different geometries with a maximum size up to 170 mm. 6 refs., 8 figs., 1 tab

  17. Reactor fueling system

    International Nuclear Information System (INIS)

    Hattori, Noriaki; Hirano, Haruyoshi.

    1983-01-01

    Purpose: To optimally position a fuel catcher by mounting a television camera to a fuel catching portion and judging video images by the use of a computer or the like. Constitution: A television camera is mounted to the lower end of a fuel catching mechanism for handling nuclear fuels and a fuel assembly disposed within a reactor core or a fuel storage pool is observed directly from above to judge the position for the fuel assembly by means of video signals. Then, the relative deviation between the actual position of the fuel catcher and that set in a memory device is determined and the positional correction is carried out automatically so as to reduce the determined deviation to zero. This enables to catch the fuel assembly without failure and improves the efficiency for the fuel exchange operation. (Moriyama, K.)

  18. Experimental investigations of heat transfer during sodium boiling in fuel assembly model in justification of advanced fast reactor safety

    International Nuclear Information System (INIS)

    Khafizov, R.R.; Poplavskij, V.M.; Rachkov, V.I.; Sorokin, A.P.; Ashurko, Yu.M.; Volkov, A.V.; Ivanov, E.F.; Privezentsev, V.V.

    2015-01-01

    The experimental facility is built up and investigation of heat exchange during sodium boiling in simulated fast reactor core assembly in conditions of natural and forced circulation with sodium plenum and upper end shield model are conducted. It is shown that in the presence of sodium plenum there is possibility to provide long-term cooling of fuel assembly when heat flux density on the surface of fuel element simulator up to 140 and 170 kW/m 2 in conditions of natural and forced circulation, respectively. The obtained data is used for improving calculational model of sodium boiling process in fuel assembly and calculational code COREMELT verification. It is pointed out that heat transfer coefficients in the case of liquid metal boiling in fuel assemblies are slightly over the ones in the case of liquid metals boiling in pipes and pool boiling [ru

  19. Nuclear fuel assembly

    International Nuclear Information System (INIS)

    1975-01-01

    The nuclear fuel assembly described includes a cluster of fuel elements supported at a distance from each other so that their axes are parallel in order to establish secondary channels between them reserved for the coolant. Several ducts for an auxiliary cooling fluid are arranged in the cluster. The wall of each duct is pierced with coolant ejection holes which are placed circumferentially to a pre-determined pattern established according to the position of the duct in the cluster and by the axial distance of the ejection hole along the duct. This assembly is intended for reactors cooled by light or heavy water [fr

  20. Comparison of hydrogen generation for TVSM and TVSA fuel assemblies for water water energy reactor (VVER)-1000

    International Nuclear Information System (INIS)

    Stefanova, A.E.; Groudev, P.P.; Atanasova, B.P.

    2009-01-01

    This paper presents the results received during investigation of hydrogen generation for both types fuel assemblies-the old modernistic type of fuel assemblies (TVSM) and recently installed new one alternative type of fuel assemblies (TVSA) in case of severe accident. There are some differences between both types FAs. They have different geometry as well as different burnable poisons. To investigate behavior of new fuel assemblies during the severe conditions it have been performed comparison of fuel behavior of old type TVSM fuel assembly to new one TVSA. To perform this investigation it has been used MELCOR 'input model' for Kozloduy Nuclear Power Plant (KNPP) VVER 1000. The model was developed by Institute for Nuclear Research and Nuclear Energy-Bulgarian Academy of Sciences (INRNE-BAS) for investigation of severe accident scenarios and Probabilistic Safety Analyses (PSA) level 2. The model provides a significant analytical capability for the Bulgarian technical specialists, working in the field of the NPP safety, for analysis of core and containment damaged states and the estimation of radionuclides release outside fuel cladding. It was accepted criteria for vessel integrity about hydrogen concentration to be 8%. This criterion was based on the decision of RSK (Germany commission for reactor safety). Generally based on the received results it was made conclusion that using both types of fuel assemblies it was not disturbance safety conditions of NPP

  1. Fuel assemblies

    International Nuclear Information System (INIS)

    Mukai, Hideyuki

    1987-01-01

    Purpose: To prevent bending of fuel rods caused by the difference of irradiation growth between coupling fuel rods and standards fuel rods thereby maintain the fuel rod integrity. Constitution: The f value for a fuel can (the ratio of pole of zirconium crystals in the entire crystals along the axial direction of the fuel can) of a coupling fuel rod secured by upper and lower tie plates is made smaller than the f value for the fuel can of a standard fuel rod not secured by the upper and the lower tie plates. This can make the irradiation growth of the fuel can of the coupling fuel rod greater than the irradiation growth of the fuel can of the standard fuel rod and, accordingly, since the elongation of the standard fuel rod can always by made greater, bending of the standard fuel rod can be prevented. (Yoshihara, M.)

  2. COUPLED SIMULATION OF GAS COOLED FAST REACTOR FUEL ASSEMBLY WITH NESTLE CODE SYSTEM

    Directory of Open Access Journals (Sweden)

    Filip Osusky

    2018-05-01

    Full Text Available The paper is focused on coupled calculation of the Gas Cooled Fast Reactor. The proper modelling of coupled neutronics and thermal-hydraulics is the corner stone for future safety assessment of the control and emergency systems. Nowadays, the system and channel thermal-hydraulic codes are accepted by the national regulatory authorities in European Union for license purposes, therefore the code NESTLE was used for the simulation. The NESTLE code is a coupled multigroup neutron diffusion code with thermal-hydraulic sub-channel code. In the paper, the validation of NESTLE code 5.2.1 installation is presented. The processing of fuel assembly homogeneous parametric cross-section library for NESTLE code simulation is made by the sequence TRITON of SCALE code package system. The simulated case in the NESTLE code is one fuel assembly of GFR2400 concept with reflective boundary condition in radial direction and zero flux boundary condition in axial direction. The results of coupled calculation are presented and are consistent with the GFR2400 study of the GoFastR project.

  3. Removable top nozzle and tool for a nuclear reactor fuel assembly

    International Nuclear Information System (INIS)

    Wilson, J.F.; Cerni, S.; Gjertsen, R.K.

    1986-01-01

    A fuel assembly is described for a nuclear reactor including a bottom nozzle, at least one longitudinally extending control rod guide thimble having an upper end and a lower end being attached to the bottom nozzle and projecting upwardly therefrom, transverse grids axially spaced along-the thimble for supporting an array of upstanding fuel rods, and a top nozzle subassembly removable mounted on the upper end of the guide thimble for obtaining top access to the fuel rods upon removal thereof. The top nozzle subassembly consists of: (a) a section integrally formed on the upper end of the guide thimble and having external threads thereon; (b) a lower adapter plate having a guide thimble hole for receiving the guide thimble so as to mount the adapter plate on the guide thimble for slidable movement therealong; (c) a retainer mounted on the guide thimble for restably supporting and limiting the downward movement of the adapter plate along the guide thimble; (d) an upper hold-down plate having a guide thimble passageway with an internal ledge for receiving the thimble so as to mount the hold-down plate on the thimble for slidable movement therealong; (e) spring means interposed between the upper hold-down plate and the lower adapter plate for biasing the hold-down plate upwardly when a downward force is applied thereon whereby the downward force is yieldably transmitted to the fuel assembly; and (f) a collar disposed within the passageway and in abutment with the ledge, the collar having an internal threaded section engageable with the externally threaded section to move the hold-down plate down against the spring means and thereby mounting of the subassembly on the guide thimble

  4. Examination process of a nuclear reactor fuel assembly and examination machine to bring the process into operation

    International Nuclear Information System (INIS)

    Delaroche, P.; Leseur, A.; Saglio, R.; Vaubert, Y.

    1983-01-01

    The machine to examine a fuel assembly of a nuclear reactor includes a support on which the assembly to be examined is placed, a source emitting waves, directed to the assembly to be examined, devices to examine the assembly to be examined that receive the emitted wave by the said source and that have been reflected by the assembly. The examination devices have an axis, this axis being directed to a mirror, this mirror being inclined in such a way that it reflects the waves reflected by the assembly to the examination devices, a radiation protection, to avoid the radiation emitted by the assembly, being diposed between the assembly and the examination devices [fr

  5. Nuclear fuel assembly

    International Nuclear Information System (INIS)

    Takeda, Tadashi; Sato, Kenji; Goto, Masakazu.

    1984-01-01

    Purpose: To facilitate identification of a fuel assembly upon fuel exchange in BWR type reactors. Constitution: Fluorescent material is coated or metal plating is applied to the impressed portion of a upper tie plate handle of a fuel assembly, and the fluorescent material or the metal plating surface is covered with a protective membrane made of transparent material. This enables to distinguish the impressed surface from a distant place and chemical reaction between the impressed surface and the reactor water can be prevented. Furthermore, since the protective membrane is formed such that it protrudes toward the upper side relative to the impressed surface, there is no risk of depositions of claddings thereover. (Moriyama, K.)

  6. Flow field measurements using LDA and numerical computation for rod bundle of reactor fuel assembly

    International Nuclear Information System (INIS)

    Hu Jun; Zou Zunyu

    1995-02-01

    Local mean velocity and turbulence intensity measurements were obtained with DANTEC 55 X two-dimensional Laser Dopper Anemometry (LDA) for rod bundle of reactor fuel assembly test model which was a 4 x 4 rod bundle. The data were obtained from different experimental cross-sections both upstream and downstream of the model support plate. Measurements performed at test Reynolds numbers of 1.8 x 10 4 ∼3.6 x 10 4 . The results described the local and gross effects of the support plate on upstream and downstream flow distributions. A numerical computation was also given, the experimental results are in good agreement with the numerical one and the others in references. Finally, a few suggestions were proposed for how to use the LDA system well. (11 figs.)

  7. Refitting of the 'Celimene' hot cell for following up the fuel assembly of 900 MWe PWR power reactors

    International Nuclear Information System (INIS)

    Lhermenier, Andre; Van Craeynest, J.-C.

    1980-05-01

    The 'Celimene' cell adjoining the EL3 reactor provides for the acceptance, handling and the examination of irradiated fuel assemblies from power reactors (length approximately 4m, weight approximately 700 kg). Within the framework of the PWR fuel behavior follow-up or reprocessing, it is possible to extract an assembly representative of the normal fuel cycle, carry out non destructive tests on this assembly, extract pencils from it and re-insert this assembly, after refitting the head, into the normal fuel cycle for handling in a reprocessing plant or storage pond. Given suitable refitting techniques, the re-irradiation of the assembly can be considered after examination. Significant changes have been made to the buildings and the hoist facilities for handling very heavy flasks. It was necessary to rearrange the handling, machining and in-cell storage facilities. The development of an inspection rig will make it possible, some time in 1980, to carry out non destructive tests of assemblies, optical and metrological examination of assemblies prior to dismantling or of the structure after dismantling [fr

  8. EXPERIMENTAL STUDY OF LOCAL HYDRODYNAMICS AND MASS EXCHANGE PROCESSES OF COOLANT IN FUEL ASSEMBLIES OF PRESSURIZED WATER REACTORS

    Directory of Open Access Journals (Sweden)

    S. M. Dmitriev

    2016-01-01

    Full Text Available The results of experimental studies of local hydrodynamics and mass exchange of coolant flow behind spacer and mixing grids of different structural versions that were developed for fuel assemblies of domestic and foreign nuclear reactors are presented in the article. In order to carry out the study the models of the following fuel assemblies have been fabricated: FA for VVER and VBER, FA-KVADRAT for PWR-reactor and FA for KLT-40C reactor. All the models have been fabricated with a full geometrical similarity with full-scale fuel assemblies. The study was carried out by simulating the flow of coolant in a core by air on an aerodynamic test rig. In order to measure local hydrodynamic characteristics of coolant flow five-channel Pitot probes were used that enable to measure the velocity vector in a point by its three components. The tracerpropane method was used for studying mass transfer processes. Flow hydrodynamics was studied by measuring cross-section velocities of coolant flow and coolant rates according to the model cells. The investigation of mass exchange processes consisted of a study of concentration distribution for tracer in experimental model, in determination of attenuation lengths of mass transfer processes behind mixing grids, in calculating of inter-cellar mass exchange coefficient. The database on coolant flow in fuel assemblies for different types of reactors had been accumulated that formed the basis of the engineering substantiation of reactor cores designs. The recommendations on choice of optimal versions of mixing grids have been taken into consideration by implementers of the JSC “OKBM Afrikantov” when creating commissioned fuel assemblies. The results of the study are used for verification of CFD-codes and CFD programs of detailed cell-by-cell calculation of reactor cores in order to decrease conservatism for substantiation of thermal-mechanical reliability.

  9. Accelerating the design and testing of LEU fuel assemblies for conversion of Russian-designed research reactors outside Russia

    International Nuclear Information System (INIS)

    Matos, J.E

    2003-01-01

    This paper identifies proposed geometries and loading specifications of LEU tube-type and pin-type test assemblies that would be suitable for accelerating the conversion of Russian-designed research reactors outside of Russia if these fuels are manufactured, qualified by irradiation testing, and made commercially available in Russia. (author)

  10. Numerical simulation of fuel assembly thermohydraulics of fast reactors with the partial blockage of cross section under the coolant

    International Nuclear Information System (INIS)

    Zhukov, A.V.; Sorokin, A.P.

    2000-01-01

    The problems of numerical modeling of thermohydraulics in assembly of fuel elements of fast reactors with the partial blockage of cross-section under the coolant are considered. The information about existing codes constructed on use of subchannel technique and model of porous body are presented. The results of calculation obtained by these codes are presented. (author)

  11. Fuel assembly

    International Nuclear Information System (INIS)

    Sakuyama, Tadashi; Mukai, Hideyuki.

    1988-01-01

    Purpose: To prevent the bending of a fuel rod caused by the difference in the elongation between a joined fuel rod and a standard fuel rod thereby maintain the fuel rod integrity. Constitution: A joined fuel rod is in a thread engagement at its lower end plug thereof with a lower plate, while passed through at its upper end plug into an upper tie plate and secured with a nut. Further, a standard fuel rod is engaged at its upper end plug and lower end plug with the upper tie plate and the lower tie plate respectively. Expansion springs are mounted to the upper end plugs of these bonded fuel rods and the standard fuel rods for preventing this lifting. Each of the fuel rods comprises a plurality of sintered pellets of nuclear fuel materials laminated in a zircaloy fuel can. The content of the alloy ingredient in the fuel can of the bonded fuel rod is made greater than that of the alloy ingredient of the standard fuel rod. this can increase the elongation for the bonded fuel rod, and the spring of the standard fuel rod is tightly bonded to prevent the bending. (Yoshino, Y.)

  12. CEA fuel pencil qualification under irradiation: from component conception to fuel assembly irradiation in a power reactor

    International Nuclear Information System (INIS)

    Marin, J.-F.; Pillet, Claude; Francois, Bernard; Morize, Pierre; Petitgrand, Sylvie; Atabek, R.-M.; Houdaille, Brigitte.

    1981-06-01

    Fabrication of fuel pins made of uranium oxide pellets and of a zircaloy 4 cladding is described. Irradiation experiment results are given. Thermomechanical behavior of the fuel pin in a power reactor is examined [fr

  13. Fuel assembly supporting structure

    International Nuclear Information System (INIS)

    Aisch, F.W.; Fuchs, H.P.; Knoedler, D.; Steinke, A.; Steven, J.

    1976-01-01

    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. 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.

  15. Three dimensional conjugated heat transfer analysis in sodium fast reactor wire-wrapped fuel assembly

    International Nuclear Information System (INIS)

    Peniguel, C.; Rupp, I.; Juhel, JP.; Rolfo, S.; Guillaud, M.; Gervais, N.

    2009-01-01

    Fast reactors with liquid metal coolant have recently received a renewed interest owing to a more efficient usage of the primary uranium resources, and they are one of the proposal for the next Generation IV. In order to evaluate nuclear power plant design and safety, 3D analysis of the flow and heat transfer in a wire spacer fuel assembly are ongoing at EDF. The introduction of the wire wrapped spacers, helically wound along the pin axis, enhances the mixing of the coolant between sub-channels and prevents contact between the fuel pins. The mesh generation step constitutes a challenging task if a reasonable amount of cells in conjunction with a suitable spatial discretization is wanted. Several approaches have been investigated and will be presented. Quite complex global flow patterns are found using either k-ε or preferably Reynolds Stress turbulent models. Preliminary conjugated heat transfer calculations using a coupling between the finite element thermal code SYRTHES and the finite volume CFD code Code Saturne are also shown. (author)

  16. Innovative technologies on fuel assemblies cleaning for sodium fast reactors: First considerations on cleaning process

    Energy Technology Data Exchange (ETDEWEB)

    Simon, N.; Lorcet, H.; Beauchamp, F.; Guigues, E. [CEA, DEN, DTN Cadarache, F-13108 Saint-Paul-lez-Durance (France); Lovera, P.; Fleche, J. L. [CEA, DEN, DPC Saclay, F-91191 Gif-sur-Yvette (France); Lacroix, M. [CEA, DEN, DTN Cadarache, F-13108 Saint-Paul-lez-Durance (France); Carra, O. [AREVA / NP, 10 Rue Juliette Recamier, 69003 Lyon (France); Dechelette, F. [CEA, DEN, DTN Cadarache, F-13108 Saint-Paul-lez-Durance (France); Prele, G. [EDF/SEPTEN, 12-14 avenue Dutrievoz, 69628 Villeurbane Cedex (France); Rodriguez, G. [CEA, DEN, DTN Cadarache, F-13108 Saint-Paul-lez-Durance (France)

    2012-07-01

    Within the framework of Sodium Fast Reactor development, innovative fuel assembly cleaning operations are investigated to meet the GEN IV goals of safety and of process development. One of the challenges is to mitigate the Sodium Water Reaction currently used in these processes. The potential applications of aqueous solutions of mineral salts (including the possibility of using redox chemical reactions) to mitigate the Sodium Water Reaction are considered in a first part and a new experimental bench, dedicated to this study, is described. Anhydrous alternative options based on Na/CO{sub 2} interaction are also presented. Then, in a second part, a functional study conducted on the cleaning pit is proposed. Based on experimental feedback, some calculations are carried out to estimate the sodium inventory on the fuel elements, and physical methods like hot inert gas sweeping to reduce this inventory are also presented. Finally, the implementation of these innovative solutions in cleaning pits is studied in regard to the expected performances. (authors)

  17. Study of short-time mechanical properties changes for BN-350 reactor spent fuel assemblies jacket material from vacancy swelling

    International Nuclear Information System (INIS)

    Karaulov, V.N.; Blynskij, A.P.; Yakovlev, I.L.; Golovin, S.V.; Lambert, D.

    1999-01-01

    Variations of mechanical properties (ultimate strength and limit of plasticity) for irradiated stainless steels, materials of BN-350 reactor cased fuel assemblies tubes, namely: 12X18H10T MTO, 08X16H11M3 MTO, 10X17H13M2T, 12X13M2BRF from vacancy swelling and neutron damaging doze have been studied. Flat samples cut out from hexagonal fuel assemblies casing were tested. The data on casing profilometry, and also the results from hydrostatic weighing of steel samples, were used to evaluate swelling. All measurements and testing were made at temperature 25 degrees C

  18. Thermal-hydraulic numerical simulation of fuel sub-assembly for Sodium-cooled Fast Reactor

    International Nuclear Information System (INIS)

    Saxena, Aakanksha

    2014-01-01

    The thesis focuses on the numerical simulation of sodium flow in wire wrapped sub-assembly of Sodium-cooled Fast Reactor (SFR). First calculations were carried out by a time averaging approach called RANS (Reynolds- Averaged Navier-Stokes equations) using industrial code STAR-CCM+. This study gives a clear understanding of heat transfer between the fuel pin and sodium. The main variables of the macroscopic flow are in agreement with correlations used hitherto. However, to obtain a detailed description of temperature fluctuations around the spacer wire, more accurate approaches like LES (Large Eddy Simulation) and DNS (Direct Numerical Simulation) are clearly needed. For LES approach, the code TRIO U was used and for the DNS approach, a research code was used. These approaches require a considerable long calculation time which leads to the need of representative but simplified geometry. The DNS approach enables us to study the thermal hydraulics of sodium that has very low Prandtl number inducing a very different behavior of thermal field in comparison to the hydraulic field. The LES approach is used to study the local region of sub-assembly. This study shows that spacer wire generates the local hot spots (∼20 C) on the wake side of spacer wire with respect to the sodium flow at the region of contact with the fuel pin. Temperature fluctuations around the spacer wire are low (∼1 C-2 C). Under nominal operation, the spectral analysis shows the absence of any dominant peak for temperature oscillations at low frequency (2-10 Hz). The obtained spectra of temperature oscillations can be used as an input for further mechanical studies to determine its impact on the solid structures. (author) [fr

  19. Nuclear fuel assembly

    International Nuclear Information System (INIS)

    Domoto, Noboru; Masuda, Hiroyuki

    1989-01-01

    In a nuclear fuel assembly loaded with a plurality of fuel rods, the inside of a fuel rod disposed at a high neutron flux region is divided into an inner region and an outer region, and more burnable poisons are mixed in the inner region than in the outer region. Alternatively, the central portion of a pellet disposed in a high neutron flux region is made hollow, in which burnable poisons are charged. This can prevent neutron infinite multiplication factor from decreasing extremely at the initial burning stage. Further, the burnable poisons are not rapidly burnt completely and local peaking coefficient can be controlled. Accordingly, in a case of suppressing a predetermined excess reactivity by using a fuel rod incorporated with the burnable poison, the fuel economy can be improved more and the reactor core controllability can also be improved as compared with the usual case. (T.M.)

  20. Nuclear fuel assembly

    International Nuclear Information System (INIS)

    Ito, Arata; Wakamatsu, Mitsuo.

    1976-01-01

    Object: To permit the coolant in an FBR type reactor to enter from the entrance nozzle into a nuclear fuel assembly without causing cavitation. Structure: In a nuclear fuel assembly, which comprises a number of thin fuel pines bundled together at a uniform spacing and enclosed within an outer cylinder, with a handling head connected to an upper portion of the outer cylinder and an entrance nozzle connected to a lower portion of the cylinder, the inner surface of the entrance nozzle is provided with a buffer member and an orifice successively in the direction of flow of the coolant. The coolant entering from a low pressure coolant chamber into the entrance nozzle strikes the buffer member and is attenuated, and thereafter flows through an orifice into the outer cylinder. (Horiuchi, T.)

  1. Transfer of fuel assemblies

    International Nuclear Information System (INIS)

    Vuckovich, M.; Burkett, J. P.; Sallustio, J.

    1984-01-01

    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

  2. Mechanisms of the initial stage of fuel elements degradation of BN reactor fuel assemblies

    International Nuclear Information System (INIS)

    Zagorul'ko, Yu.I.; Kashcheev, M.V.; Ganichev, N.S.

    2015-01-01

    On the base of developed calculational technique numerical evaluation is carried out to the time of fuel element fracture in conditions of loss of sodium flow through fuel element jacket. Data on mechanical properties of steel EhK-164 is used in calculations. Calculations are carried out for different conditions of jacket outer surface cooling: by sodium of 1073 K temperature, by boiling sodium and by sodium in condition of film boiling. It is shown that time to jacket fracture under considered rupture mechanisms essentially depends on fuel element cooling conditions [ru

  3. Establishing the fuel burn-up measuring system for 106 irradiated assemblies of Dalat reactor by using gamma spectrometer method

    International Nuclear Information System (INIS)

    Nguyen Minh Tuan; Pham Quang Huy; Tran Tri Vien; Trang Cao Su; Tran Quoc Duong; Dang Tran Thai Nguyen

    2013-01-01

    The fuel burn-up is an important parameter needed to be monitored and determined during a reactor operation and fuel management. The fuel burn-up can be calculated using computer codes and experimentally measured. This work presents the theory and experimental method applied to determine the burn-up of the irradiated and 36% enriched VVR-M2 fuel type assemblies of Dalat reactor. The method is based on measurement of Cs-137 absolute specific activity using gamma spectrometer. Designed measuring system consists of a collimator tube, high purity Germanium detector (HPGe) and associated electronics modules and online computer data acquisition system. The obtained results of measurement are comparable with theoretically calculated results. (author)

  4. Measuring device for the distribution of burn-up degree in fuel assembly irradiated in nuclear reactor

    International Nuclear Information System (INIS)

    Kumanomido, Hironori

    1989-01-01

    The object of the invention is to measure the distribution of burn-up degree, of fuel assemblies irradiated in a nuclear reactor in a short time and exactly. That is, the device comprises a device main body having substantially the same length as that for the axial length of a fuel assembly and a detector container disposed axially slidably to the main body. A plurality of radiation detectors are arranged at an equi-axial pitch and contained in the container. The container is caused to slide at a pitch equal to the equi-axial distance of the detectors. In the device having thus been constituted, measurement is conducted at least for twice at an axial position on the side of a fuel assembly irradiated in the nuclear reactor and a position caused to slide therefrom by one pitch. Based on the result, the sensitivities between each of the detectors are compared and the relative sensitivity of the radiation detectors is calibrated. Accordingly, the sensitivity between each of the detectors can be calibrated rapidly and easily. As a result, the distribution of the burn-up degree, etc of irradiated fuel assembly can be measured exactly. (K.M.)

  5. Removable top nozzle and tool for a nuclear reactor fuel assembly

    International Nuclear Information System (INIS)

    Wilson, J.F.; Cerni, S.; Gjertsen, R.K.

    1987-01-01

    A tool is described used in a nuclear reactor fuel assembly for connecting and disconnecting an internally threaded collar threadably engagable with an externally threaded upper end of a control rod guide thimble. The tool consists of: (a) rotation means engagable with the collar and operable to apply a torque in one direction to threadably connect the internally threaded collar on the thimble's externally threaded upper end and in an opposite direction to disconnect the collar from its threaded connection with the thimble; and (b) gripper means adapted to be inserted into the upper end of the guide thimble and operable to prevent the thimble from rotating about its longitudinal axis as the rotation means applies torque to the collar in connecting and disconnecting the collar on and off the guide thimble; (c) the gripper means being disposed within the rotation means and having a portion thereof projecting outwardly beyond the rotation means for insertion into the thimble when the rotation means is engaged with the collar

  6. Performance Specification Shippinpark Pressurized Water Reactor Fuel Drying and Canister Inerting System for PWR Core 2 Blanket Fuel Assemblies Stored within Shippingport Spent Fuel Canisters

    International Nuclear Information System (INIS)

    JOHNSON, D.M.

    2000-01-01

    This specification establishes the performance requirements and basic design requirements imposed on the fuel drying and canister inerting system for Shippingport Pressurized Water Reactor (PWR) Core 2 blanket fuel assemblies (BFAs) stored within Shippingport spent fuel (SSFCs) canisters (fuel drying and canister inerting system). This fuel drying and canister inerting system is a component of the U.S. Department of Energy, Richland Operations Office (RL) Spent Nuclear Fuels Project at the Hanford Site. The fuel drying and canister inerting system provides for removing water and establishing an inert environment for Shippingport PWR Core 2 BFAs stored within SSFCs. A policy established by the U.S. Department of Energy (DOE) states that new SNF facilities (this is interpreted to include structures, systems and components) shall achieve nuclear safety equivalence to comparable U.S. Nuclear Regulatory Commission (NRC)-licensed facilities. This will be accomplished in part by applying appropriate NRC requirements for comparable NRC-licensed facilities to the fuel drying and canister inerting system, in addition to applicable DOE regulations and orders

  7. Stepwise evolution of fuel assembly design toward a sustainable fuel cycle with hard neutron spectrum light water reactors

    International Nuclear Information System (INIS)

    Uchikawa, Sadao; Okubo, Tsutomu; Nakano, Yoshihiro

    2011-01-01

    An advanced LWR with hard neutron spectrum, FLWR, aims at efficient and flexible utilization of nuclear resources by evolving its fuel assembly design keeping the same core configuration. A proposed evolution process of the design toward a sustainable fuel cycle is composed of three stages, the first one based on the LWR fuel cycle infrastructures, the second one for transitioning from the LWR fuel cycle to the FR fuel cycle, and the third one based on the FR fuel cycle infrastructures. For the first stage, a fuel assembly design concept named FLWR/MIX has been developed in which enriched UO 2 fuel rods are arranged in the peripheral region of the assembly, surrounding the MOX fuel rods in the central region. The FLWR/MIX design realizes a breeder type operation under the framework of the LWR-MOX technologies and there experience. A modified FLWR/MIX design with low Pu inventory for the second stage has a potential of high Puf conversion ratio of 1.1 and can contribute to smooth and speedy transition from the LWR fuel cycle to the FR fuel cycle. For the third stage, the FLWR/MIX design is extended into a design with natural UO 2 fuel rods to realize multiple Pu recycling keeping a Puf conversion ratio of around 1.0. (author)

  8. The passive nondestructive assay of the plutonium content of spent-fuel assemblies from the BN-350 fast-breeder reactor in the city of Aqtau, Kazakhstan

    CERN Document Server

    Lestone, J P; Rennie, J A; Sprinkle, J K; Staples, P; Grimm, K N; Hill, R N; Cherradi, I; Islam, N; Koulikov, J; Starovich, Z

    2002-01-01

    The International Atomic Energy Agency is presently interested in developing equipment and techniques to measure the plutonium content of breeder reactor spent-fuel assemblies located in storage ponds before they are relocated to more secure facilities. We present the first quantitative nondestructive assay of the plutonium content of fast-breeder reactor spent-fuel assemblies while still underwater in their facility storage pond. We have calibrated and installed an underwater neutron coincidence counter (Spent Fuel Coincidence Counter (SFCC)) in the BN-350 reactor spent-fuel pond in Aqtau, Kazakhstan. A procedure has been developed to convert singles and doubles (coincidence) neutron rates observed by the SFCC into the total plutonium content of a given BN-350 spent-fuel assembly. The plutonium content has been successfully determined for spent-fuel assemblies with a contact radiation level as high as approx 10 sup 5 Rads/h. Using limited facility information and multiple measurements along the length of spe...

  9. Fuel assembly outlet temperature profile influence on core by-pass flow and power distribution determination in WWER -440 reactors

    International Nuclear Information System (INIS)

    Petenyi, V.; Klucarova, K.; Remis, J.

    2003-01-01

    The in core instrumentation of the WWER-440 reactors consists of the thermocouple system and the system of self powered detectors (SPD). The thermocouple systems are positioned about 50 cm above the fuel bundle upper flow-mixing grid. The usual assumption is that, the coolant is well mixed in the Tc location, i.e. the temperature is constant through the flow cross-section area. The present evaluations by using the FLUENT 5.5.14 code reveal that, this assumption is not fulfilled. There exists a temperature profile that depends on fuel assembly geometry and on inner power profile of the fuel assembly. The paper presents the estimation of this effect and its influence on the core power distribution and the core by-pass flow determination. Comparison with measurements in Mochovce NPP will also be a part of this presentation (Authors)

  10. Nuclear reactor fuel element assembly spacer grid and method of making

    International Nuclear Information System (INIS)

    Chetter, J.

    1975-01-01

    A cellular fuel element assembly spacer grid is described which provides for resilient bracing of fuel pins in the cells of the grid by bow spring locating members projecting inside the cells of the grid to hold the fuel pins against opposed rigid stops also projecting inside the cells of the grid. The grid comprises two tiers each formed from intersecting strip members defining cells which are penetrated by the fuel pins and arranged parallel to one another but spaced apart. The bow spring locating members extend longitudinally between the two tiers and have end ferrules which are a sliding fit on locating members which extend longitudinally from the facing inner edges of the strip members forming the two tiers. The grid tiers are fabricated individually by heat bonding the intersecting strip members prior to assembling the tiers into the spacer grid. (U.S.)

  11. Fuel assembly guide tube

    International Nuclear Information System (INIS)

    Jabsen, F.S.

    1979-01-01

    This invention is directed toward a nuclear fuel assembly guide tube arrangement which restrains spacer grid movement due to coolant flow and which offers secondary means for supporting a fuel assembly during handling and transfer operations

  12. CFD analysis of flow distribution of reactor core and temperature rise of coolant in fuel assembly for VVER reactor

    International Nuclear Information System (INIS)

    Du Daiquan; Zeng Xiaokang; Xiong Wanyu; Yang Xiaoqiang

    2015-01-01

    Flow field of VVER-1000 reactor core was investigated by using computational fluid dynamics code CFX, and the temperature rise of coolant in hot assembly was calculated. The results show that the maximum value of flow distribution factor is 1.12 and the minimum value is 0.92. The average value of flow distribution factor in hot assembly is 0.97. The temperature rise in hot assembly is higher than current warning limit value ΔT t under the deviated operation condition. The results can provide reference for setting ΔT t during the operation of nuclear power plant. (authors)

  13. A report on the transport of MTR-type spent fuel assemblies of the Philippine Research Reactor (PRR-1)

    International Nuclear Information System (INIS)

    Yoshisaki, Magno B.; Leopando, Leonardo S.

    1999-03-01

    Fifty one (51) fuel assemblies of mixed enrichment from the Philippine Research Reactor (PRR-1), consisting of 50 spent and 1 fresh, were shipped to the United States last 14 March 1999 under the U.S. Return of Foreign Research Reactor (FRR) fuel policy. The shipment was in line with the U.S. initiative to implement its Record of Decision (ROD) which took effect on 13 May 1996 to accept and manage all FRR uranium fuel of U.S. origin and enriched in the United States. The shipment program would last10 years, ending midnight of 13 May 2006. The ROD provided a 3 year extension period within which to accept FRR spent nuclear fuel (SNF) withdrawn from reactors after 2006. The U.S. policy gave priority to the NPT significance of high enriched U, as the prime target of the return of FRR policy. Classified as a developing country, the Philippines, through the PNRI, signed a contract with the U.S. Department of Energy for the cost-free shipment of PRR-1 spent fuel to the United States. Spent fuel loading and transport operations to the port area lasted seven (7) days, from 8 to 14 March 1999. (Author)

  14. Optimization of the fuel assembly for the Canadian Supercritical Water-cooled Reactor (SCWR)

    Energy Technology Data Exchange (ETDEWEB)

    French, C.; Bonin, H.; Chan, P., E-mail: Corey.French@rmc.ca [Royal Military College of Canada, Dept. of Chemistry and Chemical Engineering, Kingston, Ontario (Canada)

    2013-07-01

    A parametric optimization of the Canadian Supercritical Water-cooled Reactor (SCWR) lattice geometry and fresh fuel content is performed in this work. With the potential to improve core physics and performance, significant gains to operating and safety margins could be achieved through slight progressions. The fuel performance codes WIMS-AECL and SERPENT are used to calculate performance factors, and use them as inputs to an optimization algorithm. (author)

  15. Thermal assessment of Shippingport pressurized water reactor blanket fuel assemblies within a multi-canister overpack within the canister storage building

    International Nuclear Information System (INIS)

    HEARD, F.J.

    1999-01-01

    A series of analyses were performed to assess the thermal performance characteristics of the Shippingport Pressurized Water Reactor Core 2 Blanket Fuel Assemblies as loaded within a Multi-Canister Overpack within the Canister Storage Building. A two-dimensional finite element was developed, with enough detail to model the individual fuel plates: including the fuel wafers, cladding, and flow channels

  16. Thermal assessment of Shippingport pressurized water reactor blanket fuel assemblies within a multi-canister overpack within the canister storage building

    Energy Technology Data Exchange (ETDEWEB)

    HEARD, F.J.

    1999-04-09

    A series of analyses were performed to assess the thermal performance characteristics of the Shippingport Pressurized Water Reactor Core 2 Blanket Fuel Assemblies as loaded within a Multi-Canister Overpack within the Canister Storage Building. A two-dimensional finite element was developed, with enough detail to model the individual fuel plates: including the fuel wafers, cladding, and flow channels.

  17. Method for using a channel box for fuel assemblies in a reactor

    International Nuclear Information System (INIS)

    Ito, Ken-ichi; Shinpo, Katsutoshi; Watahiki, Minoru.

    1975-01-01

    Object: To extend a service life of a channel box used in a light water nuclear reactor. Structure: A channel box mounted in the outer periphery of a fuel bundle is removed from the fuel bundle after use for a predetermined period of time, and the removed channel box is re-mounted on the fuel bundle with opposite ends of the box reversed in position for further use. By this arrangement, structural deformation of the channel box may be minimized to extend the service life of the box. (Kamimura, M.)

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

    International Nuclear Information System (INIS)

    Hernandez, H.; Alonso, G.

    2001-01-01

    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)

  19. Method and device for storing irradiated respectively spent fuel assemblies from pressurized and boiling water reactors

    International Nuclear Information System (INIS)

    Pirk, H.; Klein, D.

    1978-01-01

    The storage compartment for the fuel elements from the PWR and the BWR consists of a concrete chamber containing at least one grating serving for vertical holding of supporting boxes e.g. of boron steel for the fuel assemblies. Further, the non-used openings of the supporting grid may be closed by means of lids so that to the space above it an underpressure may be applied for safety reasons. In the lower part of the concrete chamber there open out supply and exhaust air shafts guided in the same duct but separately from one another towards respectively away from the concrete chamber. The supply air shafts open out below or sideways from the boxes while the exhaust air shafts discharge below the upper most grid. The critical distance between the boxes or the fuel assemblies retained. (DG) [de

  20. Effect of lattice deformation on temperature fields and heat transfer in the fuel elements of characteristic zones for a model of fast reactor fuel assembly

    International Nuclear Information System (INIS)

    Zhukov, A.V.; Matyukhin, N.M.; Sviridenko, E.Ya.

    1980-01-01

    Given are the experimental results for temperature fields in the model assembly in nonribbed simulators of the BN-600-type reactor fuel elements in the course of deformation of the lattice caused by shifting of the central and peripheral (lateral, angular) fuel elements by the value of the gap between the fuel elements (the limiting case when the fuel elements touch each other along the whole length). An assembly consisting of 37 electroheated pipes arranged in a triangular lattice with a relative step of S/d=1.185 is used as a model. The experiments were carried out on the sodium stand at constant energy release along the length of the fuel element simulators and at the Pe number changing in the 14-700 range. The data obtained show considerable increase of nonuniformities of the fuel element temperatures for characteristic zones of the fuel cassette assembly models of the fast reactor at deviations of the lattice geometric sizes from the nominal ones. For the central nonribbed element the temperature nonuniformity increases approximately 7.5 times and for the lateral element approximately 6 times when the elements touch each other along the whole length. The shift the central nonribbed element by the value of the gap between the fu.el elements leads to the decrease of heat transfer in comparison with heat transfer at the nominal geometry approximately 3-7 times in the 10-450 range for the Pe numbers. It is shown that the coolant temperature distribution along the assembly radius has a complex character (with a peak between the centre and the perifery) caused by redistribution of coolant consumptions due to fuel element lattice deformation

  1. APPLICATION OF MULTIHOLE PRESSURE PROBE FOR RESEARCH OF COOLANT VELOCITY PROFILE IN NUCLEAR REACTOR FUEL ASSEMBLIES

    Directory of Open Access Journals (Sweden)

    S. M. Dmitriev

    2015-01-01

    Full Text Available Development of heat and mass transfer intensifiers is a major engineering task in the design of new and modernization of existing fuel assemblies. These devices create lateral mass flow of coolant. Design of intensifiers affects both the coolant mixing and the hydraulic resistance. The aim of this work is to develop a methodology of measuring coolant local velocity in the fuel assembly models with different mixing grids. To solve the problems was manufactured and calibrated multihole pressure probe. The air flow velocity measuring method with multihole pressure probe was used in the experimental studies on the coolant local hydrodynamics in fuel assemblies with mixing grids. Analysis of the coolant lateral velocity vector fields allowed to study the formation of the secondary vortex flows behind the mixing grids, and to determine the basic laws of coolant flow in experimental models. Quantitative data on the coolant flow velocity distribution obtained with a multihole pressure probe make possible to determine the magnitude of the flow lateral velocities in fuel rod gaps, as well as to determine the distance at which damping occurs during mixing. 

  2. Feasibility study for LEU conversion of the WWR-K reactor at the Institute of Nuclear Physics in Kazakhstan using a 5-tube fuel assembly

    International Nuclear Information System (INIS)

    Hanan, N.A.; Liaw, J.R.; Matos, J.E.

    2005-01-01

    A feasibility study by the RERTR program for possible LEU conversion of the 6 MW WWR-K reactor concludes that conversion is feasible using an LEU 5-tube Russian fuel assembly design. This 5-tube design is one of several LEU fuel assembly designs being studied (Ref. 1) for possible use in this reactor. The 5-tube assembly contains 200 g 235 U with an enrichment of 19.7% in four cylindrical inner tubes and an outer hexagonal tube with the same external dimensions as the current HEU (36%) 5-tube fuel assembly, which contains 112.5 g 235 U. The fuel meat material, LEU UO 2 -Al dispersion fuel with ∼ 2.5 g U/cm 3 , has been extensively irradiation tested in a number of reactors with uranium enrichments of 36% and 19.7%. Since the 235 U loading of the LEU assemblies is much larger than the HEU assemblies, a smaller LEU core with five rows of fuel assemblies is possible (instead of six rows of fuel assemblies in the HEU core). This smaller LEU core would consume about 60% as many fuel assemblies per year as the current HEU core and provide thermal neutron fluxes in the inner irradiation channels that are ∼ 17% larger than with the present HEU core. The current 21 day cycle length would be maintained and the average discharge burnup would be ∼ 42%. Neutron fluxes in the five outer irradiation channels would be smaller in the LEU core unless these channels can be moved closer to the LEU fuel assemblies. Results show that the smaller LEU core would meet the reactor's shutdown margin requirements and would have an adequate thermal-hydraulic safety margin to onset of nucleate boiling. (author)

  3. GAIA: AREVAs New PWR fuel assembly design

    Energy Technology Data Exchange (ETDEWEB)

    Vollmert, N.; Gentet, G.; Louf, P.H.; Mindt, M.; O' Brian, J.; Peucker, J.

    2015-07-01

    GAIA is the label of a new PWR Fuel Assembly design developed by AREVA with the objective to provide its customers an advanced fuel assembly design regarding both robustness and performance. Since 2012 GAIA lead fuel assemblies are under irradiation in a Swedish reactor and since 2015 in a U.S. reactor. Visual inspections and examinations carried out so far during the outages confirmed the intended reliability, robustness and the performance enhancement of the design. (Author)

  4. Device for measuring the local power in nuclear reactor fuel assemblies

    International Nuclear Information System (INIS)

    Thor, E.R.; Korpaas, Harald; Leyse, R.H.; Smith, R.D.

    1977-01-01

    The purpose of the invention is to make possible the fabrication of a gamma ray thermometer, capable of being accurately calibrated, before being mounted in the core of the reactor, by means of a system simulating the power released by the fuel. As already indicated, the principle used for performing the measurement with a gamma ray thermometer consists, schematically, in establishing a constant and given conductance path in the mass of an absorbent body and, after calibration, deducting from the temperature difference recorded at the ends of this path the value of the calorific power absorbed by the body and therefore generated by the surrounding nuclear fuel [fr

  5. Device for facilitating the insertion and withdrawal of fuel assemblies from a nuclear reactor

    International Nuclear Information System (INIS)

    Andrea, C.; Siegel, E.A.

    1976-01-01

    A device is provided which is installed in a reactor prior to carrying out refueling operations and which accurately locates and isolates a selected core location to permit rapid withdrawal and insertion of fuel subassemblies at that location. A shielded plug designed to cooperate with the refueling apparatus is inserted into an access port in the reactor head. A structural shroud extends down from the plug and carries at its lower end a radially floating, hexagonal spreader tube with mechanisms to rotate it for angular alignment purposes and a linear drive for inserting it into the core. The upper end of the spreader tube serves as a guide for leading the fuel handling apparatus into alignment with the chosen subassembly

  6. Prototype vibration measurement program for reactor internals (177-fuel assembly plant). Supplement 1

    International Nuclear Information System (INIS)

    Simonis, J.C.; Post, R.C.; Thoren, D.E.

    1976-08-01

    The surveillance specimen holder tubes installed in the Babcock and Wilcox 177-fuel assembly plants have been redesigned. The structural adequacy of this design has been verified through extensive analysis. The design adequacy will be further confirmed by measuring the vibrational response of the surveillance specimen holder tube during normal and transient flow operation. This report describes the vibration measurement program that will be conducted at Toledo Edison's Davis Besse 1 site

  7. Coupled neutronics/thermal-hydraulics analysis of a high-performance light-water reactor fuel assembly

    International Nuclear Information System (INIS)

    Waata, C.L.

    2006-07-01

    The use of water at supercritical pressure as coolant and moderator introduces a challenge in the design of a High-Performance Light-Water Reactor (HPLWR) fuel assembly. At supercritical pressure condition (P=25 MPa), the thermal-hydraulics behaviour of water differs strongly from that at sub-critical pressure due to a rapid variation of the thermal-physical properties across the pseudo-critical line. Due of the strong link between the water (moderation) and the neutron spectrum and subsequently the power distribution, a coupling of neutronics and thermal-hydraulics has become a necessity for reactor concepts operating at supercritical pressure condition. The effect of neutron moderation on the local parameters of thermal-hydraulics and vice-verse in a fuel assembly has to be considered for an accurate design analysis. In this study, the Monte Carlo N-Particle code (MCNP) and the sub-channel code STAFAS (Sub-channel Thermal-hydraulics Analysis of a Fuel Assembly under Supercritical conditions) have been coupled for the design analysis of a fuel assembly with supercritical water as coolant and moderator. Both codes are well known for complex geometry modelling. The MCNP code is used for neutronics analyses and for the prediction of power profiles of individual fuel rods. The sub-channel code STAFAS for the thermal-hydraulics analyses takes into account the coolant properties beyond the critical point as well as separate moderator channels. The coupling procedure is realized automatically. MCNP calculates the power distribution in each fuel rod, which is then transferred into STAFAS to obtain the corresponding thermal-hydraulic conditions in each sub-channel. The new thermal-hydraulic conditions are used to generate a new input deck for the next MCNP calculation. This procedure is repeated until a converged state is achieved. The coupled code system was tested on a proposed fuel assembly design of a HPLWR. An under-relaxation was introduced to achieve convergence

  8. Model for determining stresses in the structure of a fast reactor fuel assembly

    International Nuclear Information System (INIS)

    Kervevan, J.-J.

    1974-01-01

    Deformations in a reactor core are due to two metallurgical phenomena, swelling of the steel under irradiation and irradiation creep when the structure is under stress. The first step is to determine the deformation of each sub-assembly supposed free, subjected to a neutron flux or temperature gradient, and the second is to study the interactions amongst most of the sub-assemblies. Under the influence of the deformations the interaction value will change with time, and this development must be determined. Calculation methods were developed for the purpose. A number of computing codes already exist and it is necessary to complete them, modify them if necessary, create new ones as the case arises and form a coherent whole [fr

  9. Fuel assembly storage pool

    International Nuclear Information System (INIS)

    Hiranuma, Hiroshi.

    1976-01-01

    Object: To remove limitation of the number of storage of fuel assemblies to increase the number of storage thereof so as to relatively reduce the water depth required for shielding radioactive rays. Structure: Fuel assembly storage rack containers for receiving a plurality of spent fuel assembly racks are stacked in multi-layer fashion within a storage pool filled with water for shielding radioactive rays and removing heat. (Furukawa, Y.)

  10. Computer program for modelling the history of the in-service bending of fast power reactor fuel assemblies

    International Nuclear Information System (INIS)

    Dienstbier, J.

    1979-04-01

    The studies into stresses and deformations in the core are mainly focused on the fuel rod and the fuel assembly can. In high neutron doses austenitic steel swells and this is associated with a considerable increase in the volume of material. The SANDRA computer program is used for solving the problems of can deformations and stress during long-term reactor operation. The block for the mechanical interaction of cans is the key part of the program. The program input data include temperature distribution, fast neutron flux distribution and coolant overpressure inside the cans. Reactor operation is modelled using operating modes A, B, C which may arbitrarily be combined. Mode A computes bending deformations and the deformations of the can cross-section due to temperature dilatation in the change in temperature fields in the reactor; mode B computes deformations due to swelling and creep in long-term operation; mode C computes thermal deformations in reactor shut-down. A flowsheet is shown of program SANDRA as are examples of computed deformations. (M.S.)

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

    International Nuclear Information System (INIS)

    Matejka, Karel; Sklenka, Lubomir

    1997-01-01

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

  12. Statistical analysis of the vibration loading of the reactor internals and fuel assemblies of reactor units type WWER-440 from deferent projects

    International Nuclear Information System (INIS)

    Ovcharov, O.; Pavelko, V.; Usanov, A.; Arkadov, G.; Dolgov, A.; Molchanov, V.; Anikeev, J.; Pljush, A.

    2006-01-01

    In this paper the following items have been presented: 1) Vibration noise instrument channels; 2) Vibration loading characteristics of control assemblies, internals and design peculiarities of internals of WWER-440 deferent projects; 3) Coolant flow rate through the reactor, reactor core, fuel assemblies and control assemblies for different projects WWER-440 and 4) Noise measurements of coolant speed per channel. The change of auto power spectrum density of absolute displacement detector signal for the last 12 years of SUS monitoring of the Kola NPP unit 2; the coherence functions groups between two SPND of the same level for the Kola NPP unit 1; the measured coolant flow rate at Paks NPP and the auto power spectrum density group of SPND signals from 11 neutron measuring channels of the Kola NPP unit 1 are given. The main factors of vibration loading of internals and fuel assemblies for Kola NPP units 1-4, Bohunice NPP units 1 and 2 and Novovoronezh NPP units 3 and 4 are also discussed

  13. Detection and replacement of crud accumulating fuel assemblies at the Loviisa-2 reactor

    International Nuclear Information System (INIS)

    Antila, M.

    1995-01-01

    At unit 2 of Loviisa NPP fuel bundle outlet temperatures had been increasing slowly from the beginning of November 1994 mainly at the six lead fuel bundles with new zirconium spacer grids (Z). The unit was taken to cold shut-down and started up again in January 1995. After a few days the increase of the outlet temperatures was confirmed to continue. It was decided to shut down the reactor again to examine and remove the new Z-bundles. One bundle was taken to closer investigations in the poolside inspection stand. The hexagonal shroud tube was removed and quite a lot of crud was detected on the lower surfaces of the spacer grids. All the six Z-bundles were removed and replaced by older bundles with steel spacers. A small reshuffling was performed to control the radial power distribution. The cause for the partial flow blockage causing the increasing temperatures was accumulation of crud particularly on the Zr-spacers. The root cause of the accumulation of crud has not yet been definitely cleared. It seems evident that it is related to the decontamination of the primary circuit performed last August. The extensive incore temperature measurement instrumentation and reactor core monitoring system had a central role in the early detection of the anomaly. Thus core thermal margins were not violated and possible fuel failures could be avoided. (orig.) (3 refs., 5 figs.)

  14. Peripheral pin alignment system for fuel assemblies

    International Nuclear Information System (INIS)

    Anthony, A.J.

    1981-01-01

    An alignment system is provided for nuclear fuel assemblies in a nuclear core. The core support structure of the nuclear reactor includes upwardly pointing alignment pins arranged in a square grid and engage peripheral depressions formed in the lateral periphery of the lower ends of each of the fuel assemblies of the core. In a preferred embodiment, the depressions are located at the corners of the fuel assemblies so that each depression includes one-quarter of a cylindrical void. Accordingly, each fuel assembly is positioned and aligned by one-quarter of four separate alignment pins which engage the fuel assemblies at their lower exterior corners. (author)

  15. Study of the mixture in an assembly of clustered fuel elements of a nuclear reactor

    International Nuclear Information System (INIS)

    Tofani, Paulo de Carvalho

    1970-01-01

    An improvement of thermal performance of fuel clusters in a nuclear reactor is closely related to the knowledge of heat transmission in the solid part and of heat exchanges in the fluid. This research thesis thus aimed at studying the mixture effects in simple phase between sub-canals in order to adjust laws which govern these effects in analytical codes. After a review of published works on flows and heat exchanges in clusters, the author presents an experimental device, reports and analyses the obtained results [fr

  16. EXPEL - a computing module for molten fuel/coolant interactions in fast reactor sub-assemblies

    International Nuclear Information System (INIS)

    Fishlock, T.P.

    1975-10-01

    This report describes a module for computing the effects of a molten fuel/coolant interaction in a fast reactor subassembly. The module is to be incorporated into the FRAX code which calculates the consequences of hypothetical whole core accidents. Details of the interaction are unknown and in consequence the model contains a large number of parameters which must be set by assumption. By variation of these parameters the interaction may be made mild or explosive. Results of a parametric survey are included. (author)

  17. Nuclear fuel sub-assemblies

    International Nuclear Information System (INIS)

    Dodd, J.A.; Butterfield, C.E.; Waite, E.

    1979-01-01

    A fast reactor fuel sub-assembly has honeycomb grids for laterally supporting the fuel pins. The grids are of two series and are arranged alternately along the bundle. The grids of a first series provide a discrete cell for each pin but the grids of the second series have a peripheral group of cells only. The grids of the second series provide intermediate support of the edge pins to restrain bow. (author)

  18. A setup for active neutron analysis of the fissile material content in fuel assemblies of nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bushuev, A. V.; Kozhin, A. F., E-mail: alexfkozhin@yandex.ru; Aleeva, T. B.; Zubarev, V. N.; Petrova, E. V.; Smirnov, V. E. [National Research Nuclear University MEPhI (Russian Federation)

    2016-12-15

    An active neutron method for measuring the residual mass of {sup 235}U in spent fuel assemblies (FAs) of the IRT MEPhI research reactor is presented. The special measuring stand design and uniform irradiation of the fuel with neutrons along the entire length of the active part of the FA provide high accuracy of determination of the residual {sup 235}U content. AmLi neutron sources yield a higher effect/background ratio than other types of sources and do not induce the fission of {sup 238}U. The proposed method of transfer of the isotope source in accordance with a given algorithm may be used in experiments where the studied object needs to be irradiated with a uniform fluence.

  19. Nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    Hindle, E. D.

    1984-01-01

    An array of rods is assembled to form a fuel element for a pressurized water reactor, the rods comprising zirconium alloy sheathed nuclear fuel pellets and containing helium. The helium gas pressure is selected for each rod so that it differs substantially from the helium gas pressure in its closest neighbors. In a preferred arrangement the rods are arranged in a square lattice and the helium gas pressure alternates between a relatively high value and a relatively low value so that each rod has as its closest neighbors up to four rods containing helium gas at the other pressure value

  20. Nuclear reactor fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Hindle, E. D.

    1984-10-16

    An array of rods is assembled to form a fuel element for a pressurized water reactor, the rods comprising zirconium alloy sheathed nuclear fuel pellets and containing helium. The helium gas pressure is selected for each rod so that it differs substantially from the helium gas pressure in its closest neighbors. In a preferred arrangement the rods are arranged in a square lattice and the helium gas pressure alternates between a relatively high value and a relatively low value so that each rod has as its closest neighbors up to four rods containing helium gas at the other pressure value.

  1. NUPEC proves reliability of LWR fuel assemblies

    International Nuclear Information System (INIS)

    Anon.

    1987-01-01

    It is very important in assuring the safety of nuclear reactors to confirm the reliability of fuel assemblies. The test program of the Nuclear Power Engineering Center on the reliability of fuel assemblies has verified the high performance and reliability of Japanese LWR fuels, and confirmed the propriety of their design and fabrication. This claim is based on the data obtained from the fuel assemblies irradiated in commercial reactors. The NUPEC program includes irradiation test which has been conducted for 11 years since fiscal 1976, and the maximum thermal loading test using the out of pile test facilities simulating a real reactor which has been continued since fiscal 1978. The irradiation test on BWR fuel assemblies in No.3 reactor in Fukushima No.1 Nuclear Power Station, Tokyo Electric Power Co., Inc., and on PWR fuel assemblies in No.3 reactor in Mihama Power Station, Kansai Electric Power Co., Inc., and the maximum thermal loading test on BWR and PWR fuel assemblies are reported. The series of postirradiation examination of the fuel assemblies used for commercial reactors was conducted for the first time in Japan, and the highly systematic data on 27 items were obtained. (Kako, I.)

  2. Experience with the transport and storage casks CASTOR (registered) MTR 2 for spent nuclear fuel assemblies from research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Jack, Allen; Rettenbacher, Katharina; Skrzyppek, Juergen [GNS Gesellschaft fuer Nuklear-Service mbH, Essen (Germany)

    2011-07-01

    The CASTOR (registered) MTR 2 cask was designed and manufactured by the company GNS during the 1990's for the transport and interim storage of spent nuclear fuel assemblies from various types of research reactors. Casks of this type have been used at the VKTA Research Centre in Rossendorf near Dresden, Germany as well as at the European Commission's Joint Research Centre at Petten and at the HOR reactor at Delft in the Netherlands. A total of 24 units have been used for the functions of transport and storage with various spent fuel types (VVER, HFR-HEU, and HOR-HEU) for more than ten years now. This type of packaging for radioactive material is a member of the CASTOR (registered) family of spent nuclear fuel casks used worldwide. Over 1000 units are loaded and in storage in Europe, Asia, Africa and North America. This paper presents the experience from the use of the casks for transport and storage in the past, as well as the prospects for the future. (author)

  3. Nuclear reactor control assembly

    International Nuclear Information System (INIS)

    Negron, S.B.

    1991-01-01

    This patent describes an assembly for providing global power control in a nuclear reactor having the core split into two halves. It comprises a disk assembly formed from at least two disks each machined with an identical surface hole pattern such that rotation of one disk relative to the other causes the hole pattern to open or close, the disk assembly being positioned substantially at the longitudinal center of and coaxial with the core halves; and means for rotating at least one of the disks relative to the other

  4. Fuel assembly spacer

    International Nuclear Information System (INIS)

    Shirakawa, Ken-etsu.

    1988-01-01

    Purpose: To reduce the pressure loss of coolants by fuel assembly spacers. Constitution: Spacers for supporting a fuel assembly are attached by means of a plurality of wires to an outer frame. The outer frame is made of shape memory alloy such that the wires are caused to slacken at normal temperature and the slacking of the wires is eliminated in excess of the transition temperature. Since the wires slacken at the normal temperature, fuel rods can be inserted easily. After the insertion of the fuel rods, when the entire portion or the outer frame is heated by water or gas at a predetermined temperature, the outer frame resumes its previously memorized shape to tighten the wires and, accordingly, the fuel rods can be supported firmly. In this way, since the fuel rods are inserted in the slacken state of the wires and, after the assembling, the outer frame resumes its memorized shape, the assembling work can be conducted efficiently. (Kamimura, M.)

  5. The fuel of nuclear reactors

    International Nuclear Information System (INIS)

    1995-03-01

    This booklet is a presentation of the different steps of the preparation of nuclear fuels performed by Cogema. The documents starts with a presentation of the different French reactor types: graphite moderated reactors, PWRs using MOX fuel, fast breeder reactors and research reactors. The second part describes the fuel manufacturing process: conditioning of nuclear materials and fabrication of fuel assemblies. The third part lists the different companies involved in the French nuclear fuel industry while part 4 gives a short presentation of the two Cogema's fuel fabrication plants at Cadarache and Marcoule. Part 5 and 6 concern the quality assurance, the safety and reliability aspects of fuel elements and the R and D programs. The last part presents some aspects of the environmental and personnel protection performed by Cogema. (J.S.)

  6. Method and apparatus for removably mounting a top nozzle on a nuclear reactor fuel assembly

    International Nuclear Information System (INIS)

    Wilson, J.F.; Gjertsen, R.K.; Schallenberger, J.M.

    1986-01-01

    In a fuel assembly having a top nozzle and control rod guide thimbles, a method is described of removably mounting the top nozzle on the ends of the guide thimbles, comprising the steps of: (a) releasably mating hollow outer sockets defined in the top nozzle with hollow inner sockets defined on the ends of the guide thimbles. The inner sockets are movable between compressed conditions for removing and inserting the inner sockets from and into the outer sockets in mounting and removing the top nozzle on and from the guide thimbles and expanded conditions for mating the inner and outer sockets together and the top nozzle on the guide thimbles; (b) supporting elongated locking tubes such that end portions thereof extend into the outer sockets defined in the top nozzle; and (c) moving all of the locking tubes at the same time between unlocking and locking positions to displace their end portions axially within the outer sockets between first and second locations

  7. Experience in arranging shipments of spent fuel assemblies of commercial and research reactors

    International Nuclear Information System (INIS)

    Komarov, S.; Barinkov, O.; Eshcherkin, A.; Lozhnikov, V.; Smirnov, A.

    2008-01-01

    At present the key activities of Sosny Company are to inspect physical conditions, handle and arrange shipment of SFA including failed SFA. In 2003 after obtaining the license of Gosatomnadzor (Rostechnadzor now) entitled to handle nuclear materials in the process of their shipment, Sosny Company started preparing certification and arranging SFA shipment on its own. About 40 shipments of SFA were performed with participation of Sosny Company. Experience in handling failed SFA - an example of development of a new technology could be the transport and technological scheme of RBMK-1000 SFA shipment from Leningradskaya NPP that was designed by Sosny Company. TUK-11 cask was selected for this shipment. The example of change of transport and technological scheme is modification of the technology for handling and shipment of WWER-440 SFA from Kola NPP. Experience in arranging transportation - based on the results of development of logistics schemes for shipping SFA of reactor facilities Sosny Company justified and implemented composition of mixed trains containing rail cars of many types that enabled to perform shipment more efficiently in time and cost. Experience in arranging handling and shipment of research reactor SFA - over the past years the activity of Sosny Company was aimed at implementing international Russian Research Reactor Fuel Return (RRRFR) program. Since equipment of the majority of research centers doesn't allow for the large casks to be accepted and loaded, special casks of less mass and dimensions are used to ship SFA from research reactors. In RRRFR program it is assumed to use different casks for RR SFA such as Russian TUK- 19, TUK-128 and foreign SKODA VPVR/M and NAC-LWT. At present Sosny Company is involved in coordination of the efforts of the affected organizations in creating the type 'C' package for RR SFA in the RF. Conclusion: Under conditions of constant increase of the requirements to shipment safety and complication of regulations of all

  8. Nuclear reactor assembly

    International Nuclear Information System (INIS)

    Dorner, H.; Scholz, M.; Jungmann, A.

    1975-01-01

    A nuclear reactor assembly includes a reactor pressure tank having a substantially cylindrical side wall surrounded by the wall of a cylindrical cavity formed by a biological shield. A rotative cylindrical wall is interposed between the walls and has means for rotating it from outside of the shield, and a probe is carried by the rotative wall for monitoring the pressure tank's wall. The probe is vertically movable relative to the rotative cylindrical wall, so that by the probe's vertical movement and rotation of the rotative cylinder, the reactor's wall can be very extensively monitored. If the reactor pressure tank's wall fails, it is contained by the rotative wall which is backed-up by the shield cavity wall. (Official Gazette)

  9. Nuclear fuel assembly

    International Nuclear Information System (INIS)

    Ueda, Tomihiro.

    1970-01-01

    The present invention relates to fuel assemblies employing wire wrap spacers for retaining uniform spatial distribution between fuel elements. Clad fuel elements are helically wound in the oxial direction with a wave-formed wire strand. The strand is therefore provided with spring action which permits the fuel elements to expand freely in the axial and radial directions so as to retain proper spacing and reduce stresses due to thermal deformation. (Ownes, K.J.)

  10. Thermal-hydraulic study of the LBE-cooled fuel assembly in the MYRRHA reactor: Experiments and simulations

    Energy Technology Data Exchange (ETDEWEB)

    Pacio, J., E-mail: Julio.pacio@kit.edu [Karlsruhe Institute of Technology (KIT), Institute for Nuclear and Energy Technologies (IKET), Hermann-von-Helmholtz Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Wetzel, T. [Karlsruhe Institute of Technology (KIT), Institute for Nuclear and Energy Technologies (IKET), Hermann-von-Helmholtz Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Doolaard, H.; Roelofs, F. [Nuclear Research and Consultancy Group (NRG), Westerduinweg 3, 1755 ZG Petten (Netherlands); Van Tichelen, K. [Belgian Nuclear Reseach Center (SCK-CEN), Boeretang 200, Mol (Belgium)

    2017-02-15

    Heavy liquid metals (HLMs), such as lead-bismuth eutectic (LBE) and pure lead are prominent candidate coolants for many advanced systems based on fast neutrons. In particular, LBE is used in the first-of-its-kind MYRRHA fast reactor, to be built in Mol (Belgium), which can be operated either in critical mode or as a sub-critical accelerator-driven system. With a strong focus on safety, key thermal-hydraulic aspects of these systems, such as the proper cooling of fuel assemblies, must be assessed. Considering the complex geometry and low Prandtl number of LBE (Pr ∼ 0.025), this flow scenario is challenging for the models used in Computational Fluid Dynamics (CFD), e.g. for relating the turbulent transport of momentum and heat. Thus, reliable experimental data for the relevant scenario are needed for validation. In this general context, this topic is studied both experimentally and numerically in the framework of the European FP7 project SEARCH (2011–2015). An experimental campaign, including a 19-rod bundle with wire spacers, cooled by LBE is undertaken at KIT. With prototypical geometry and operating conditions, it is intended to evaluate the validity of current empirical correlations for the MYRRHA conditions and, at the same time, to provide validation data for the CFD simulations performed at NRG. The results of one benchmarking case are presented in this work. Moreover, this validated approach is then used for simulating a complete MYRRHA fuel assembly (127 rods).

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

    International Nuclear Information System (INIS)

    Oh, Jinho

    2013-01-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Jinho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-07-01

    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.

  13. Safety for fuel assembly handling in the nuclear ship Mutsu

    International Nuclear Information System (INIS)

    Ando, Yoshio

    1978-01-01

    The safety for fuel assembly handling in the nuclear ship Mutsu is deliberated by the committee of general inspection and repair technique examination for Mutsu. The result of deliberation for both cases of removing fuel assemblies and keeping them in the reactor is outlined. The specification of fuel assemblies, and the nuclides and designed radioactivity of fission products of fuel are described. The possibility of shielding repair work and general safety inspection keeping the fuel assemblies in the reactor, the safety consideration when the fuel assemblies are removed at a quay, in a dry dock and on the ocean, the safety of fuel transport in special casks and fuel storage are explained. It is concluded finally that the safety of shielding repair work and general inspection work is secured when the fuel assemblies are kept in the reactor and also when the fuel assemblies are removed from the reactor by cautious working. (Nakai, Y.)

  14. Fuel Fabrication and Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Karpius, Peter Joseph [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-02-02

    The uranium from the enrichment plant is still in the form of UF6. UF6 is not suitable for use in a reactor due to its highly corrosive chemistry as well as its phase diagram. UF6 is converted into UO2 fuel pellets, which are in turn placed in fuel rods and assemblies. Reactor designs are variable in moderators, coolants, fuel, performance etc.The dream of energy ‘too-cheap to meter’ is no more, and now the nuclear power industry is pushing ahead with advanced reactor designs.

  15. Facility for the storage of spent, heat-emitting and container-enclosed nuclear reactor fuel assemblies

    International Nuclear Information System (INIS)

    Hennings, U.

    1987-01-01

    Patent for facility for the storage of spent, heat-emitting and container-enclosed nuclear reactor fuel assemblies, which are arranged within a building in a horizontal position and are cooled by a gas stream, whereby the building has a storage and a loading zone, characterized by the fact that pallet trucks arranged one above the other in a row and such that an interspace is left for the receiving positions for the containers, the the pallet trucks can be moved along rails that extend between two side walls arranged opposite to one another in the storage zone, that the storage zone can be loaded and unloaded by opening located in these two side walls, and that the gas stream only circulates within the building

  16. Fuel nozzle assembly

    Science.gov (United States)

    Johnson, Thomas Edward [Greer, SC; Ziminsky, Willy Steve [Simpsonville, SC; Lacey, Benjamin Paul [Greer, SC; York, William David [Greer, SC; Stevenson, Christian Xavier [Inman, SC

    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.

  17. A porous medium approach for the fluid structure interaction modelling of a water pressurized nuclear reactor core fuel assemblies: simulation and experimentation

    International Nuclear Information System (INIS)

    Ricciardi, G.

    2008-10-01

    The designing of a pressurized water reactor core subjected to seismic loading, is a major concern of the nuclear industry. We propose, in this PhD report, to establish the global behaviour equations of the core, in term of a porous medium. Local equations of fluid and structure are space averaged on a control volume, thus we define an equivalent fluid and an equivalent structure, of which unknowns are defined on the whole space. The non-linear fuel assemblies behaviour is modelled by a visco-elastic constitutive law. The fluid-structure coupling is accounted for by a body force, the expression of that force is based on empirical formula of fluid forces acting on a tube subject to an axial flow. The resulting equations are solved using a finite element method. A validation of the model, on three experimental device, is proposed. The first one presents two fuel assemblies subjected to axial flow. One of the two fuel assemblies is deviated from its position of equilibrium and released, while the other is at rest. The second one presents a six assemblies row, immersed in water, placed on a shaking table that can simulate seismic loading. Finally, the last one presents nine fuel assemblies network, arranged in a three by three, subject to an axial flow. The displacement of the central fuel assembly is imposed. The simulations are in agreement with the experiments, the model reproduces the influence of the flow of fluid on the dynamics and coupling of the fuel assemblies. (author)

  18. Nuclear fuel: modelling the advanced plutonium assembly

    International Nuclear Information System (INIS)

    Kaoua, Th.; Lenain, R.

    2004-01-01

    The benefits of modeling in the nuclear sector are illustrated by the example of the design study for a new plutonium fuel assembly, APA, capable of ensuring maximum consumption of this fuel in pressurized-water reactors. Beyond the physical design of the assembly and its integration into the reactor, this serves for the working out of a complete materials flow and assists in modeling production from the entire inventory of nuclear power stations. (authors)

  19. Nuclear fuel: modelling the advanced plutonium assembly

    International Nuclear Information System (INIS)

    N'kaoua, Th.; Lenain, R.

    2002-01-01

    The benefits of modeling in the nuclear sector are illustrated by the example of the design study for a new plutonium fuel assembly, APA, capable of ensuring maximum consumption of this fuel in pressurized-water reactors. Beyond the physical design of the assembly and its integration into the reactor, this serves for the working out of a complete materials flow and assists in modeling production from the entire inventory of nuclear power stations. (authors)

  20. Reconstitutable nuclear reactor fuel assembly with unitary removable top nozzle subassembly

    International Nuclear Information System (INIS)

    Shallenberger, J.M.

    1987-01-01

    A reconstitutable fuel assembly is described having at least one control rod guide thimble and a top nozzle, the guide thimble including an upper extension, the top nozzle including at least one hold-down spring, an upper hold-down plate and a lower adapter plate, an improved attaching structure removably mounting the top nozzle as a unitary subassembly on the guide thimble. The attaching structure comprises: (a) a coupling member interfitting the lower adapter plate, the upper hold-down plate and the hold-down spring disposed between the plates so as to capture and retain the plates and spring together as a unitary subassembly in which the upper plate is slidably moveable along the coupling member relative to the lower plate with the spring biasing the upper plate away from the lower plate. The coupling member has spaced apart upper and lower portions with a central passageway extending for slidably receiving the upper extension of the guide thimble in a nonattached relationship in which the coupling member is slidably movable relative to the guide thimble extension for respectively inserting and removing the coupling member on and from the guide thimble extension

  1. Study of the neutronic behavior of a fuel assembly with gadolinium of a reactor HPLWR; Estudio del comportamiento neutronico de un ensamble combustible con gadolinia de un reactor HPLWR

    Energy Technology Data Exchange (ETDEWEB)

    Barragan M, A.; Martin del Campo M, C.; Francois L, J. L. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Mexico D. F. (Mexico); Espinosa P, G., E-mail: albrm29@yahoo.com [Universidad Autonoma Metropolitana, Unidad Iztapalapa, Area de Ingenieria en Recursos Energeticos, Av. San Rafael Atlixco 186, Col. Vicentina, 09340 Mexico D. F. (Mexico)

    2012-10-15

    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{infinity}) and the maximum power peaking factor (PPF{sub max}), as well as the reactivity coefficients by the fuel temperature. The k{infinity} and PPF{sub max} 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 - Gd{sub 2O3}) in some rods of the same assembly. The obtained results show values k{infinity} and PPF{sub max} 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)

  2. Parameters calculation of fuel assembly with complex geometry

    International Nuclear Information System (INIS)

    Wu Hongchun; Ju Haitao; Yao Dong

    2006-01-01

    The code DRAGON was developed for CANDU reactor by Ecole Polytechnique de Montreal of Canada. In order to validate the DRAGON code's applicability for complex geometry fuel assembly calculation, the rod shape fuel assembly of PWR benchmark problem and the plate shape fuel assembly of MTR benchmark problem were analyzed by DRAGON code. Some other shape fuel assemblies were also discussed simply. Calculation results show that the DRAGON code can be used to calculate variform fuel assembly and the precision is high. (authors)

  3. Experience in WWER fuel assemblies vibration analysis

    International Nuclear Information System (INIS)

    Ovtcharov, O.; Pavelko, V.; Usanov, A.; Arkadov, G.; Dolgov, A.; Molchanov, V.

    2003-01-01

    It is stated that the vibration studies of internals and the fuel assemblies should be conducted during the reactor designing, commissioning and commercial operation stages and the analysis methods being used should complement each other. The present paper describes the methods and main results of the vibration noise studies of internals and the fuel assemblies of the operating NPPs with WWER reactors, as an example of the implementation of the comprehensive approach to the analysis on equipment flow-induced vibration. At that, the characteristics of internals and fuel assemblies vibration loading were dealt jointly as they are elements of the same compound oscillating system and their vibrations have the interrelated nature

  4. Fuel assembly inspection device

    International Nuclear Information System (INIS)

    Yaginuma, Yoshitaka

    1998-01-01

    The present invention provides a device suitable to inspect appearance of fuel assemblies by photographing the appearance of fuel assemblies. Namely, the inspection device of the present invention measures bowing of fuel assembly or each of fuel rods or both of them based on the partially photographed images of fuel assembly. In this case, there is disposed a means which flashily projects images in the form of horizontal line from a direction intersecting obliquely relative to a horizontal cross section of the fuel assembly. A first image processing means separates the projected image pictures including projected images and calculates bowing. A second image processing means replaces the projected image pictures of the projected images based on projected images just before and after the photographing. Then, images for the measurement of bowing and images for inspection can be obtained simultaneously. As a result, the time required for the photographing can be shortened, the time for inspection can be shortened and an effect of preventing deterioration of photographing means by radiation rays can be provided. (I.S.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-05-15

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

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

    International Nuclear Information System (INIS)

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

    2012-01-01

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

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

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  8. Investigation of the burn-up behavior of boron poison rods, placed in a fuel assembly of a pressurized water reactor

    International Nuclear Information System (INIS)

    Arnold, C.; Lutz, D.C.

    1979-09-01

    The excess reactivity of a pressurized water reactor is compensated by boron, disolved in the moderator. In addition during the first cycle boron poison rods are placed in fuel assemblies without control rods. The burn-up behavior of a poison rod in a Biblis B fuel assembly is analysed in the present paper. Multigroup spectrum calculations were performed. The influence of critical boron concentration depending from burn-up, the changes of fuel concentration and the concentration of burnable poison were taken into consideration. Furthermore the built-up of rapidly saturating fisson products 135 Xe and 149 Sm was considered. The interaction of these effects are discussed. Spatial influences are emphasized most. Finally two group cross sections were calculated. The results are compared with calculations for a fuel assembly of the same type without burnable poison rods. (orig.) [de

  9. Integral nuclear fuel element assembly

    International Nuclear Information System (INIS)

    Schluderberg, D. C.

    1985-01-01

    An integral nuclear fuel element assembly utilizes longitudinally finned fuel pins. The continuous or interrupted fins of the fuel pins are brazed to fins of juxtaposed fuel pins or directly to the juxtaposed fuel pins or both. The integrally brazed fuel assembly is designed to satisfy the thermal and hydraulic requirements of a fuel assembly lattice having moderator to fuel atom ratios required to achieve high conversion and breeding ratios

  10. Fuel assembly, channel box of fuel assembly, fuel spacer of fuel assembly and method of manufacturing channel box

    International Nuclear Information System (INIS)

    Chaki, Masao; Kanazawa, Toru; Orii, Akihito; Nagayoshi, Takuji; Nishida, Koji; Kawasaki, Terufumi.

    1997-01-01

    In a fuel assembly of a BWR type reactor, fuel rods disposed at corners of side walls of a channel box or in the periphery of the side walls are partially removed, and recessed portions are formed on the side walls of the channel box from which the fuel rods are removed. Spaces closed at the sides are formed in the inner side of the corner portions. Openings are formed for communicating the closed space with the outside of the channel box. Then, the channel area of the outer side of the channel box is increased, through which much water flows to increase the amount of water in the reactor core thereby promoting the moderation of neutrons and providing thermal neutrons suitable to nuclear fission. The degree of freedom for distribution of the spaces in the reactor core is increased to improve neutron economy thereby enabling to utilize reactor fuels effectively. (N.H.)

  11. Rod drop in the LR-0 reactor core comprising 55 fuel assemblies

    International Nuclear Information System (INIS)

    Hadek, J.; Grundmann, U.

    1989-09-01

    Data from the third stage of kinetic measurements on the LR-0 reactor, performed in 1988, were employed for additional calculations using the 3-dimensional neutron kinetics code HEXDYN3D. The reactor consists of subassemblies similar to those in the WWER-1000 (PWR) reactor. The theoretical and experimental results are compared for the time behavior of the neutron flux caused by drop of the control rod cluster in various subassemblies of the reactor. The results demonstrate that the HEXDYN3D code is well suited to the treatment of the space-time behavior of the neutron flux. (author). 21 figs., 2 tabs., 16 refs

  12. Fueling method in LMFBR type reactors

    International Nuclear Information System (INIS)

    Kawashima, Katsuyuki; Inoue, Kotaro.

    1985-01-01

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

  13. Fuel bundle for nuclear reactor

    International Nuclear Information System (INIS)

    Long, J.W.; Flora, B.S.; Ford, K.L.

    1977-01-01

    The invention concerns a new, simple and inexpensive system for assembling and dismantling a nuclear reactor fuel bundle. Several fuel rods are fitted in parallel rows between two retaining plates which secure the fuel rods in position and which are maintained in an assembled position by means of several stays fixed to the two end plates. The invention particularly refers to an improved apparatus for fixing the stays to the upper plate by using locking fittings secured to rotating sleeves which are applied against this plate [fr

  14. Method of fueling for a nuclear reactor

    International Nuclear Information System (INIS)

    Igarashi, Takao.

    1983-01-01

    Purpose: To enable the monitoring of reactor power with sufficient accuracy, upon starting even without existence of neutron source in case of a low average burnup degree in the reactor core. Constitution: Each of fuel assemblies is charged such that neutron source region monitors for the start-up system in a reactor core neutron instrumentation system having nuclear fuel assemblies and a neutron instrumentation system are surrounded with 4 or 16 fuel assemblies of a low burnup degree. Then, the average burnup degree of the fuel assemblies surrounding the neutron source region monitors are increased than the reactor core burnup degree, whereby neutrons released from the peripheral fuels are increased, sufficient number of neutron counts can be obtained even with no neutron sources upon start-up and the reactor power can be monitored at a sufficient accuracy. (Sekiya, K.)

  15. Installation and method for handling fuel assemblies of fast nuclear reactors

    International Nuclear Information System (INIS)

    Aubert, Michel; Renaux, Charley.

    1982-01-01

    This invention concerns an installation and a method for handling the assemblies which makes it possible to have a large revolving plug smaller in diameter than that of the presently known solutions. This large, coaxial to the core, revolving plug has a handling arm enabling a fraction of the assemblies to be reached and deposited in a handling well. Through a small offset revolving plug the remainder of the assemblies can be reached and deposited in a pick-up well accessible to the arm of the large revolving plug [fr

  16. Gross gamma-ray measurements of light water reactor spent-fuel assemblies in underwater storage arrays

    International Nuclear Information System (INIS)

    Moss, C.E.; Lee, D.M.

    1980-12-01

    Two gross gamma-ray detection systems have been developed for rapid measurement of spent-fuel assemblies in underwater storage racks. One system uses a scintillator as the detector and has a 2% crosstalk between a fuel assembly and an adjacent void. The other system uses an ion chamber as the detector. The measurements with both detectors correlate well with operator-declared burnup and cooling-time values

  17. Thermal hydraulics of accelerator breeder systems for regeneration of reactor fuel assemblies

    International Nuclear Information System (INIS)

    Yu, W.S.; Powell, J.R.

    1979-01-01

    The following conclusions are obtained with regard to the thermal-hydraulic behavior of the Linear Accelerator Fuel Regenerator for PWR and CANDU fuel: (1) two-phase flow is a feasible coolant option for fuel element heat fluxes up to 1 x PWR (or CANDU) average value, which is the maximum design value for a LAFR; (2) two-phase flow pressure drops are low (typically 10 to 30 psi) and film temperature drops very low (typically approx. 10 0 F) for PWR fuel, with inlet velocity range (50 to 75 ft/sec). A somewhat higher inlet velocity range (75 to 100 ft/sec) and pressure drop (50 to 100 psi) is necessary for CANDU fuel, however, to prevent dry out

  18. Fuel assembly manufacturing device

    International Nuclear Information System (INIS)

    Picard, P.; Villaeys, R.

    1995-01-01

    The device comprises a central support on which the frame is mounted, a magazine which supports the fuel rods in passages aligned with those in the frame and a traction assembly on the opposite side of the magazine and including an array of pull rods designed to be advanced through the passages in the frame, to grip respective fuel rods in magazine and to pull those rods into the passages on the return stroke. 13 figs

  19. The problems of calculation of heat transfer crisis in fuel assemblies of PW reactors based on modern versions of thermohydraulic codes

    International Nuclear Information System (INIS)

    Fialko, N.M.; Sharaevskij, G.I.; Sharaevskaya, E.I.; Babak, E.I.

    2014-01-01

    The article gives an analysis of the adequacy of computer software systems FASCICLE BM-DF and COBRA, which are designed to calculate the main parameters of the safety of water-cooled nuclear reactors. This calculation is based on determining the local thermal-hydraulic parameters of the flow of coolant in the fuel rod assembled elements. In this article introduced the results of the comparison of experiments performed to determine the distribution of the main thermal-hydraulic flow parameters characteristic of subchannels of fuel rod assembled elements with the data for calculating these parameters on the basis of declared computer codes. Particular attention is paid to the analysis of experimental and calculation data, by definition, burnout in rod fuel assembled elements

  20. Facilities of fuel transfer for nuclear reactors

    International Nuclear Information System (INIS)

    Wade, E.E.

    1977-01-01

    This invention relates to sodium cooled fast breeder reactors. It particularly concerns facilities for the transfer of fuel assemblies between the reactor core and a fuel transfer area. The installation is simple in construction and enables a relatively small vessel to be used. In greater detail, the invention includes a vessel with a head, fuel assemblies housed in this vessel, and an inlet and outlet for the coolant covering these fuel assemblies. The reactor has a fuel transfer area in communication with this vessel and gear inside the vessel for the transfer of these fuel assemblies. These facilities are borne by the vessel head and serve to transfer the fuel assemblies from the vessel to the transfer area; whilst leaving the fuel assemblies completely immersed in a continuous mass of coolant. A passageway is provided between the vessel and this transfer area for the fuel assemblies. Facilities are provided for closing off this passageway so that the inside of the reactor vessel may be isolated as desired from this fuel transfer area whilst the reactor is operating [fr

  1. Fuel and fuel cycles with high burnup for WWER reactors

    International Nuclear Information System (INIS)

    Chernushev, V.; Sokolov, F.

    2002-01-01

    The paper discusses the status and trends in development of nuclear fuel and fuel cycles for WWER reactors. Parameters and main stages of implementation of new fuel cycles will be presented. At present, these new fuel cycles are offered to NPPs. Development of new fuel and fuel cycles based on the following principles: profiling fuel enrichment in a cross section of fuel assemblies; increase of average fuel enrichment in fuel assemblies; use of refuelling schemes with lower neutron leakage ('in-in-out'); use of integrated fuel gadolinium-based burnable absorber (for a five-year fuel cycle); increase of fuel burnup in fuel assemblies; improving the neutron balance by using structural materials with low neutron absorption; use of zirconium alloy claddings which are highly resistant to irradiation and corrosion. The paper also presents the results of fuel operation. (author)

  2. Nuclear fuel assembly seismic amplitude limiter

    International Nuclear Information System (INIS)

    Anthony, A.J.

    1977-01-01

    The ability of a nuclear reactor to withstand high seismic loading is enhanced by including, on each fuel assembly, at least one seismic grid which reduces the magnitude of the possible lateral deflection of the individual fuel elements and the entire fuel assembly. The reduction in possible deflection minimizes the possibility of impact of the spacer grids of one fuel assembly on those of an adjacent fuel assembly and reduces the magnitude of forces associated with any such impact thereby minimizing the possibility of fuel assembly damage as a result of high seismic loading. The seismic grid is mounted from the fuel assembly guide tubes, has greater external dimensions when compared to the fuel assembly spacer grids and normally does not support or otherwise contact the fuel elements. The reduction in possible deflection is achieved through reduction of the clearance between adjacent fuel assemblies made possible by the use in the seismic grid of a high strength material characterized by favorable thermal expansion characteristics and minimal irradiation induced expansion

  3. Reactor core and control rod assembly in FBR type reactor

    International Nuclear Information System (INIS)

    Fujimura, Koji; Kawashima, Katsuyuki; Itooka, Satoshi.

    1993-01-01

    Fuel assemblies and control rod assemblies are attached respectively to reactor core support plates each in a cantilever fashion. Intermediate spacer pads are disposed to the lateral side of a wrapper tube just above the fuel rod region. Intermediate space pads are disposed to the lateral side of a control rod guide tube just above a fuel rod region. The thickness of the intermediate spacer pad for the control rod assembly is made smaller than the thickness of the intermediate spacer pad for the fuel assembly. This can prevent contact between intermediate spacer pads of the control guide tube and the fuel assembly even if the temperature of coolants is elevated to thermally expand the intermediate spacer pad, by which the radial displacement amount of the reactor core region along the direction of the height of the control guide tube is reduced substantially to zero. Accordingly, contribution of the control rod assembly to the radial expansion reactivity can be reduced to zero or negative level, by which the effect of the negative radial expansion reactivity of the reactor is increased to improve the safety upon thermal transient stage, for example, loss of coolant flow rate accident. (I.N.)

  4. Nuclear fuel assembly

    International Nuclear Information System (INIS)

    Borrman, B.; Nylund, O.

    1984-01-01

    A fuel assembly with a fuel channel which surrounds a plurality of fuel rods and which is divided, by means of a stiffening device of cruciform cross-section and four wings, into four sub-channels each of which comprises a bundle of fuel rods. Each fuel channel side has a plurality of stamped, inwardly-directed projections, arranged vertically one after the other, aid projections being welded to one and the same stiffening wing. Each one of the wall portions located between the projections defines, together with two adjacently positioned projections and a portion of the stiffening wing, a communiation opening between two bundles located on on one side each of the stiffening wing. (Author)

  5. Device for absorbing the axial forces occurring on the fuel assemblies during operation of a nuclear reactor

    International Nuclear Information System (INIS)

    Sankovich, M.F.

    1978-01-01

    The fuel assemblies consisting of rod-shaped fuel rods stand on a grid plate. Opposite the projections of the upper grid plate mounted on a support barrel the fuel assemblies are elastically supported in order to compensate the mechanical vibrations and thermal expansions occurring during operation. This is achieved by combined bending and torsion springs bridging the distance between projections and fuel assembly end pieces. The bending and torsion springs consist of a bending arm, a torsion piece, and another bending arm being deflected by 90 0 and provided at the end with an upsetting. Each spring consists of round stock. In order to increase the flexibility one of the bending arms is designed conically or stepped. (DG) [de

  6. Impact analysis of spent fuel jacket assemblies

    International Nuclear Information System (INIS)

    Aramayo, G.A.

    1994-01-01

    As part of the analyses performed in support of the reracking of the High Flux Isotope Reactor pool, it became necessary to prove the structural integrity of the spent fuel jacket assemblies subjected to gravity drop that result from postulated accidents associated with the handling of these assemblies while submerged in the pool. The spent fuel jacket assemblies are an integral part of the reracking project, and serve to house fuel assemblies. The structure integrity of the jacket assemblies from loads that result from impact from a height of 10 feet onto specified targets has been performed analytically using the computer program LS-DYNA3D. Nine attitudes of the assembly at the time of impact have been considered. Results of the analyses show that there is no failure of the assemblies as a result of the impact scenarios considered

  7. Nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    Hindle, E.D.

    1984-01-01

    The fuel elements for a pressurised water reactor comprise arrays of rods of zirconium alloy sheathed nuclear fuel pellets. The helium gas pressure within each rod differs substantially from that of its closest neighbours

  8. Unification of fuel elements for research reactors

    International Nuclear Information System (INIS)

    Vatulyn, A.V.; Stetskyi, Y.A.; Dobrikova, I.V.

    1997-01-01

    To the purpose of fuel elements unification the possibility of rod fuel assembly (FA) using in the cores of research reactors have been considered in this paper. The calculation results of geometric, hydraulic and thermotechnical parameters of rod assembly are submitted. Several designs of finned square fuel element and fuel assembly are proposed on base of analysis of rod FA characteristics in compare of tube ones. The fuel elements specimens and the model assembly are manufactured. The developed designs are the basis for further optimization after neutron-physical calculations of cores. (author)

  9. Sodium-cooled fast reactor (SFR) fuel assembly design with graphite-moderating rods to reduce the sodium void reactivity coefficient

    Energy Technology Data Exchange (ETDEWEB)

    Won, Jong Hyuck; Cho, Nam Zin, E-mail: nzcho@kaist.ac.kr; Park, Hae Min; Jeong, Yong Hoon, E-mail: jeongyh@kaist.ac.kr

    2014-12-15

    Highlights: • The graphite rod-inserted SFR fuel assembly is proposed to achieve low sodium void reactivity. • The neutronics/thermal-hydraulics analyses are performed for the proposed SFR cores. • The sodium void reactivity is improved about 960–1030 pcm compared to reference design. - Abstract: The concept of a graphite-moderating rod-inserted sodium-cooled fast reactor (SFR) fuel assembly is proposed in this study to achieve a low sodium void reactivity coefficient. Using this concept, two types of SFR cores are analyzed; the proposed SFR type 1 core has new SFR fuel assemblies at the inner/mid core regions while the proposed SFR type 2 core has a B{sub 4}C absorber sandwich in the middle of the active core region as well as new SFR fuel assemblies at the inner/mid core regions. For the proposed SFR core designs, neutronics and thermal-hydraulic analyses are performed using the DIF3D, REBUS3, and the MATRA-LMR codes. In the neutronics analysis, the sodium void reactivity coefficient is obtained in various void situations. The two types of proposed core designs reduce the sodium void reactivity coefficient by about 960–1030 pcm compared to the reference design. However, the TRU enrichment for the proposed SFR core designs is increased. In the thermal hydraulic analysis, the temperature distributions are calculated for the two types of proposed core designs and the mass flow rate is optimized to satisfy the design constraints for the highest power generating assembly. The results of this study indicate that the proposed SFR assembly design concept, which adopts graphite-moderating rods which are inserted into the fuel assembly, can feasibly minimize the sodium void reactivity coefficient. Single TRU enrichment and an identical fuel slug diameter throughout the SFR core are also achieved because the radial power peak can be flattened by varying the number of moderating rods in each core region.

  10. Research reactor fuel - an update

    International Nuclear Information System (INIS)

    Finlay, M.R.; Ripley, M.I.

    2003-01-01

    In the two years since the last ANA conference there have been marked changes in the research reactor fuel scene. A new low-enriched uranium (LEU) fuel, 'monolithic' uranium molybdenum, has shown such promise in initial trials that it may be suitable to meet the objectives of the Joint Declaration signed by Presidents Bush and Putin to commit to converting all US and Russian research reactors to LEU by 2012. Development of more conventional aluminium dispersion UMo LEU fuel has continued in the meantime and is entering the final qualification stage of multiple full sized element irradiations. Despite this progress, the original 2005 timetable for UMo fuel qualification has slipped and research reactors, including the RRR, may not convert from silicide to UMo fuel before 2007. The operators of the Swedish R2 reactor have been forced to pursue the direct route of qualifying a UMo lead test assembly (LTA) in order to meet spent fuel disposal requirements of the Swedish law. The LTA has recently been fabricated and is expected to be loaded shortly into the R2 reactor. We present an update of our previous ANA paper and details of the qualification process for UMo fuel

  11. Experience with reactor assembly of FBTR

    International Nuclear Information System (INIS)

    Srinivasan, G.; Ravishankar, K.; Babu, A.; Varadarajan, S.; Arumugam, P.; Sekhar, P.

    2006-01-01

    Reactor Assembly, also called Block Pile, is the heart of FBTR and houses the core, top and lateral shields, control rod drive mechanisms (CRDM), sodium inlet pipe and outlet pipes etc. Two major problems which arose during commissioning were reactor vessel tilt due to convection in cover gas space and failure of inflatable seals. The reactor vessel tilt was solved by Helium injection. Reactor was operated without pressurising the inflatable seals till 2005, when the seals were replaced. Other major problems in the course of twenty years of reactor operation were failure of three CRDM lower parts, Core Cover plate which houses the core thermocouples getting stuck in the fuel handling position, water leaks from the Biological Shield Cooling (BSC) coils around the reactor, failure of core wires in the trailing cables during fuel handling etc. This paper addresses the major problems faced and modifications carried out. (author)

  12. UKAEA fast reactor project research and development programme on fuel element cladding and sub-assembly wrapper materials

    International Nuclear Information System (INIS)

    Harries, D.R.

    1977-01-01

    Research and development work on fuel element component (cladding, subassembly wrappers, etc.) materials for the U.K. sodium cooled fast reactor programme has been conducted at the United Kingdom Atomic Energy Authority (UKAEA) establishments at Dounreay, Harwell, Risley, and Springfields during the past fifteen years or so. This work has formed an integral part of, and has been co-ordinated by, the UKAEA Fast Reactor Project and has involved close liaison with the Nuclear Power Company (NPC) and the Central Electricity Generating Board (CEGB). The research and development were initially related to the Prototype Fast Reactor (PFR) but the scope has now been extended to cover the first Civil Fast Reactor (CFR1), which has recently been re-designated the Civil Demonstration Fast Reactor (CDFR). The paper outlines the present status of the development of sodium cooled fast reactors in the U.K. and proceeds to summarize the principal PFR and CDFR core and fuel element parameters which have determined the planning and direction of the fuel element materials programme. The current position on the fuel element cladding and wrapper research and development programme is reviewed, and the facilities and future irradiation programme to be carried out in PFR are described

  13. Nuclear fuel assembly

    International Nuclear Information System (INIS)

    Hirano, Yasushi; Hirukawa, Koji; Sakurada, Koichi.

    1994-01-01

    A bundle of fuel rods is divided into four fuel rod group regions of small fuel rod bundles by a cross-shaped partitioning structure consisting of paired plate-like structures which connect two opposing surfaces of a channel box. A water removing material with less neutron absorption (for example, Zr or a Zr alloy) or a solid moderator is inserted and secured to a portion of a non-boiling water region interposed between the paired plate-like structure. It has a structure that light water flows to the region in the plate-like structure. The volume, density or composition of the water removing material is controlled depending on the composition of the fuels, to change the moderating characteristics of neutrons in the non-boiling water region. This can easily moderate the difference of nuclear characteristics between each of fuel assemblies using fuel materials of different fuel compositions. Further, the reactivity control effect of the burnable poisons can be enhanced without worsening fuel economy or linear power density. (I.N.)

  14. Experience and results of material science research conducted on spent fuel assemblies from the BN-350 fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Maksimkin, O.; Gusev, M.; Turubarova, L.G.; Tsai, K.V.; Yarovchuk, A.V. [Institute of Nuclear Physics, Almaty (Kazakhstan)

    2007-07-01

    Full text of publication follows: The BN-350 fast reactor was commissioned in 1973, ran successfully for many years and is now in the decommission stage. Its unique operational parameters (low temperature of sodium at the input, wide range of damage rates, etc. ) allowed the investigation of a number of new radiation effects on both austenitic and ferritic-martensitic steels. The latter class of steel was extensively employed as wrappers for fuel assemblies. Much of the accumulated experience in BN-350 is relevant to development of fusion devices. Results are presented on post-operational research of steels 12Cr18Ni10Ti, 08Cr16Ni11Mo3, and 12Cr13Mo2BFR, all serving as hexagonal shrouds of fuel assemblies. Structural materials in the active core zone operated at temperatures of 280-430 deg. C, and were irradiated the range of 0.25-83 dpa with damage rates of 10{sup -9} - 10{sup -6} dpa/s). Investigations of irradiated hexagonal shroud materials were performed with using traditional techniques of transmission and scanning electron microscopy, metallography, mechanical tests, hydrostatic weighing, magnetometry, etc. Additionally, new techniques have been developed and employed with great success on these highly irradiated materials, such as optical computer extensometry, and magnetization cartography. Typical results to be covered in this presentation are: a) In 12Cr18Ni10Ti steel irradiated at a low dose rate of 0.12 x 10{sup -8} dpa/s voids were found at 281 deg. C after only 0.65 dpa, demonstrating once again the acceleration of swelling at low dpa rates observed in other steels. b) Data on helium release during annealing of highly irradiated sample are presented. c) Differences in deformation-induced hardening between the shroud's corners and faces leads to post-irradiation differences in swelling and mechanical properties. d) During room temperature mechanical tests of 12Cr18Ni10Ti steel at {approx}56 dpa at 350 deg. C it was found that ductility lost at

  15. Nuclear fuel assembly

    International Nuclear Information System (INIS)

    Wakamatsu, Mitsuo.

    1974-01-01

    Object: To improve a circulating flow passage of coolant so as to be able to accurately detect the temperature of coolant, rare gases contained, and the like. Structure: A fuel assembly comprising a flow regulating lattice provided with a plurality of communication holes in an axial direction, said lattice being positioned at the upper end of an outer tube in which nuclear fuel elements are received, and a neutron shielding body having a plurality of spiral coolant flow passages disposed between the lattice and the nuclear fuel elements, whereby a coolant comprised of liquid sodium or the like, which moves up passing through the coolant flow passages and the flow regulating passage, is regulated and passed through a detector mounted at the upper part of the flow regulating lattice to detect coolant temperature, flow rate, and rare gases or the like as the origin of nuclear fission contained in the coolant due to breakage of fuel elements. (Kamimura, M.)

  16. A classification scheme for LWR fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Moore, R.S.; Williamson, D.A.; Notz, K.J.

    1988-11-01

    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.

  17. A classification scheme for LWR fuel assemblies

    International Nuclear Information System (INIS)

    Moore, R.S.; Williamson, D.A.; Notz, K.J.

    1988-11-01

    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

  18. Computer simulation of variform fuel assemblies using Dragon code

    International Nuclear Information System (INIS)

    Ju Haitao; Wu Hongchun; Yao Dong

    2005-01-01

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

  19. On-line system for monitoring of boiling in nuclear reactor fuel assemblies

    International Nuclear Information System (INIS)

    Tuerkcan, E.; Kozma, R.; Nabeshima, K.; Verhoef, J.P.

    1993-01-01

    The performance of the boiling detection system has been tested on boiling signals coming from the research reactor HOR during experiments with the NIOBE boiling setup. Several detection methods utilizing frequency domain analysis have been tested both on- and off-line. Results of these methods indicate that boiling detection is possible in real-time even in the incipient stage of the boiling. Both DC and AC components of the in-core and ex-core neutron detector signals can be used for boiling detection; these two components provide complementary information. Advanced signal analysis application to the DC signals may give information about the dynamic changes of the reactor, provided that the changes of the signal exceed the inherent noise of the measured channel. At the same time, AC signal analysis will characterize the changes even in the inherent signal fluctuation level. Boiling experiments of HOR and the methods implemented for signal analysis validates the techniques used for these experiments. (orig./HP)

  20. Fuel assembly insertion system

    International Nuclear Information System (INIS)

    Barkhurst, D.J.

    1987-01-01

    This patent describes a nuclear reactor facility having fuel bundles: a system for the insertion of a fuel bundle into a position where vertically arranged fuel bundles surround and are adjacent the system comprising, in combination, separate and individual centering devices secured to and disposed on top of each fuel bundle adjacent the position. Each such centering device has a generally box-like cap configuration on the upper end of each fuel bundle and includes: a top wall; first and second side walls, each secured along and upper edge to the top wall; a rear plate attached along opposite vertical edges to the first and second side walls; a front inclined wall joined along an upper edge to the top to the wall and attached along opposite vertical edges first and second side walls; pad means secured to the lower edge of the first and second side walls, the front inclined wall and the rear plate for mounting each centering device on top of an associated fuel bundle; pin means carried by at least two of the pad means engageable with an associated aperature for locating and laterally fixing each centering device on top of its respective fuel bundle. Each front inclined wall of each of the centering devices is orientated on top of its respective fuel bundle to slope upwardly and away from the position where upon downward insertion of a fuel bundle any contact between the lower end of the fuel bundle inserted with a front inclined wall of a centering device will laterally deflect the fuel bundle. Each centering device further includes a central socket means secured to the top wall, and an elongated handling pole pivotally attached to the socket

  1. Fuel element shipping shim for nuclear reactor

    International Nuclear Information System (INIS)

    Gehri, A.

    1975-01-01

    A shim is described for use in the transportation of nuclear reactor fuel assemblies. It comprises a member preferably made of low density polyethylene designed to have three-point contact with the fuel rods of a fuel assembly and being of sufficient flexibility to effectively function as a shock absorber. The shim is designed to self-lock in place when associated with the fuel rods. (Official Gazette)

  2. Fuel transfer system for a nuclear reactor

    International Nuclear Information System (INIS)

    Katz, L.R.; Marshall, J.R.; Desmarchais, W.E.

    1977-01-01

    Disclosed is a fuel transfer system for moving nuclear reactor fuel assemblies from a new fuel storage pit to a containment area containing the nuclear reactor, and for transferring spent fuel assemblies under water from the reactor to a spent fuel storage area. The system includes an underwater track which extends through a wall dividing the fuel building from the reactor containment and a car on the track serves as the vehicle for moving fuel assemblies between these two areas. The car is driven by a motor and linkage extending from an operating deck to a chain belt drive on the car. A housing pivotally mounted at its center on the car is hydraulically actuated to vertically receive a fuel assembly which then is rotated to a horizontal position to permit movement through the wall between the containment and fuel building areas. Return to the vertical position provides for fuel assembly removal and the reverse process is repeated when transferring an assembly in the opposite direction. Limit switches used in controlling operation of the system are designed to be replaced from the operating deck when necessary by tools designed for this purpose. 5 claims, 8 figures

  3. Conceptual design of reactor assembly of prototype fast breeder reactor

    International Nuclear Information System (INIS)

    Selvaraj, A.; Balasubramaniyan, V.; Raghupathy, S.; Elango, D.; Sodhi, B.S.; Chetal, S.C.; Bhoje, S.B.

    1996-01-01

    The conceptual design of Reactor Assembly of 500 MWe Prototype Fast Breeder Reactor (as selected in 1985) was reviewed with the aim of 'simplification of design', 'Compactness of the reactor assembly' and 'ease in construction'. The reduction in size has been possible by incorporating concentric core arrangement, adoption of elastomer seals for Rotatable plugs, fuel handling with one transfer arm type mechanism, incorporation of mechanical sealing arrangement for IHX at the penetration in Inner vessel redan and reduction in number of components. The erection of the components has been made easier by adopting 'hanging' support for roof slab with associated changes in the safety vessel design. This paper presents the conceptual design of the reactor assembly components. (author). 8 figs, 2 tabs

  4. Fuel sub-assembly

    International Nuclear Information System (INIS)

    Jolly, R.

    1988-01-01

    A nuclear fuel sub-assembly includes a hexagonal bundle of parallel, spaced apart fuel pins coupled at one end to an end-holding grid comprising a number of transverse spaced apart rails to each of which is connected a series of pin-receiving cells which render the pins axially captive with the rails. The series of cells are defined by a pair of metal strips each of which has a series of pocket formations such that when the pocket formations are in registry they define cylindrical shaped cells provided with internal projections which engage annular recesses in the end caps of the fuel pins to effect axial constraint of the pins. (author)

  5. Nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    Butterfield, C.E.; Waite, E.

    1982-01-01

    A nuclear reactor fuel element comprising a column of vibration compacted fuel which is retained in consolidated condition by a thimble shaped plug. The plug is wedged into gripping engagement with the wall of the sheath by a wedge. The wedge material has a lower coefficient of expansion than the sheath material so that at reactor operating temperature the retainer can relax sufficient to accommodate thermal expansion of the column of fuel. (author)

  6. Unified fuel elements development for research reactors

    International Nuclear Information System (INIS)

    Vatulin, A.; Stetsky, Y.; Dobrikova, I.

    1998-01-01

    Square cross-section rod type fuel elements have been developed for russian pool-type research reactors. new fuel elements can replace the large nomenclature of tubular fuel elements with around, square and hexahedral cross-sections and to solve a problem of enrichment reduction. the fuel assembly designs with rod type fuel elements have been developed. The overall dimensions of existing the assemblies are preserved in this one. the experimental-industrial fabricating process of fuel elements, based on a joint extrusion method has been developed. The fabricating process has been tested in laboratory conditions, 150 experimental fuel element samples of the various sizes were produced. (author)

  7. Siemens advance PWR fuel assemblies (HTP) and cladding

    International Nuclear Information System (INIS)

    Stout, R. B.; Woods, K. N.

    1997-01-01

    This paper describes the key features of the Siemens HTP (High Thermal Performance) fuel design, the current in-reactor performance of this advanced fuel assembly design, and the advanced cladding types available

  8. Commissioning of the STAR test section for experimental simulation of loss of coolant accident using the EC-208 instrumented fuel assembly of the IEA-R1 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Maprelian, Eduardo; Torres, Walmir M.; Prado, Adelk C.; Umbehaun, Pedro E.; Franca, Renato L.; Santos, Samuel C.; Macedo, Luiz A.; Sabundjian, Gaiane, E-mail: emaprel@ipen.br, E-mail: wmtorres@ipen.br, E-mail: acprado@ipen.br, E-mail: umbehaun@ipen.br, E-mail: rlfranca@ipen.br, E-mail: samuelcs@ipen.br, E-mail: lamacedo@ipen.br, E-mail: gdjian@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SO (Brazil)

    2015-07-01

    The three basic safety functions of Research Reactors (RR) are the safe shutdown of the reactor, the proper cooling of the decay heat of the fuel elements and the confinement of radioactive materials. Compared to Nuclear Power Reactors, RR power release is small, yet its three safety functions must be met to ensure the integrity of the reactor. During a loss of coolant accident (LOCA) in pool type RR, partial or complete loss of pool water may occur, with consequent partial or complete uncovering of the fuel assemblies. In such an accident, the decay heat removal safety function must not be compromised. The Test Section for Experimental Simulation of Loss of Coolant Accident (STAR) is in commissioning phase. This test section will provide experimental data on partial and total uncovering of the EC-208 instrumented fuel assembly (IFA) irradiated in the IEA-R1. Experimental results will be useful in validation of computer codes for RR safety analysis, particularly on heat removal efficiency aspects (safety function) in accident conditions. STAR comprises a base on which is installed the IFA, the cylindrical stainless steel hull, the compressed air system for the test section emptying and refilling, and the instrumentation for temperature and level measurements. The commissioning tests or pre-operational check, consist of several preliminary tests to verify experimental procedures, the difficulties during assembling of STAR in the pool, the difficulties in control the emptying and refilling velocities, as well as, the repeatability capacity, tests of equipment, valves and systems and tests of instrumentation and data acquisition system. Safety, accuracy and easiness of operation will be checked. (author)

  9. Experiments on simulation of coolant mixing in fuel assembly head and core exit channel of WWER-440 reactor

    International Nuclear Information System (INIS)

    Kobzar, L.L; Oleksyuk, D.A.

    2006-01-01

    RRC 'Kurchatov Institute' has performed coolant mixing investigation in a head of a full-size simulator of WWER-440 fuel assembly. The experiments were focused on obtaining the data important for investigating the trends in temperature difference between the value registered by a ICIS thermocouple and the value of average temperature. The completed experiments ensure representative of configuration simulation by reproducing every construction peculiar feature of flow part of fuel assembly in the domain between the lower spacing grid and thermocouple location, and also by slightly modified fuel assembly regular elements (or analogues thereof). For the purpose of effectiveness of coolant mixing assessment within the head cross section of FA simulator, we measured coolant temperature distribution both in the place where coolant flow leaves the rod bundle simulator (in 39 data points along the cross section) and in the cross section location of regular ICIS thermocouple simulator (30 data points). The testing was conducted with pressure of (90 - 95) bar, mass coolant flow rates up to 2000 kg/(m 2 .s), temperature of coolant heating in 'hot' parts of the bundle up to 35.. and differences between coolant temperature extremes measured in rod bundle simulator outlet up to 20... Temperature fields were registered in 63 conditions that differ in coolant flow and inlet coolant temperature, electrical heating rate of FA simulator, and radial coolant distribution. In certain registered conditions we simulated coolant leakage to the space between the fuel assemblies. The received test data may be important both for investigation of dependencies between the coolant temperature in regular thermocouple location or average outlet temperature in assembly head, and for validation of CFD codes or subchannel codes (Authors)

  10. Fuel cell sub-assembly

    Science.gov (United States)

    Chi, Chang V.

    1983-01-01

    A fuel cell sub-assembly comprising a plurality of fuel cells, a first section of a cooling means disposed at an end of the assembly and means for connecting the fuel cells and first section together to form a unitary structure.

  11. Fast reactor fuel design and development

    International Nuclear Information System (INIS)

    Bishop, J.F.W.; Chamberlain, A.; Holmes, J.A.G.

    1977-01-01

    Fuel design parameters for oxide and carbide fast reactor fuels are reviewed in the context of minimising the total uranium demands for a combined thermal and fast reactor system. The major physical phenomena conditioning fast reactor fuel design, with a target of high burn-up, good breeding and reliable operation, are characterised. These include neutron induced void swelling, irradiation creep, pin failure modes, sub-assembly structural behaviour, behaviour of defect fuel, behaviour of alternative fuel forms. The salient considerations in the commercial scale fabrication and reprocessing of the fuels are reviewed, leading to the delineation of possible routes for the manufacture and reprocessing of Commercial Reactor fuel. From the desiderata and restraints arising from Surveys, Performance and Manufacture, the problems posed to the Designer are considered, and a narrow range of design alternatives is proposed. The paper concludes with a consideration of the development areas and the conceptual problems for fast reactors associated with those areas

  12. Fuel arrangement for high temperature gas cooled reactor

    International Nuclear Information System (INIS)

    Tobin, J.M.

    1978-01-01

    Disclosed is a fuel arrangement for a high temperature gas cooled reactor including fuel assemblies with separate directly cooled fissile and fertile fuel elements removably inserted in an elongated moderator block also having a passageway for control elements

  13. Blockages in LMFBR fuel assemblies: a review

    International Nuclear Information System (INIS)

    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. 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.

  15. Inter renewal travelling wave reactor with rotary fuel columns

    International Nuclear Information System (INIS)

    Terai, Yuzo

    2016-01-01

    To realize the COP21 decision, this paper proposes Inter Renewal Travelling Wave Reactor that bear high burn-up rate 50% and product TRU fuel efficiently. The reactor is based on 4S Fast Reactor and has Reactor Fuel Columns as fuel assemblies that equalize temperature in the fuel assembly so that fewer structure is need to restrain thermal transformation. To equalize burn-up rate of all fuel assemblies in the reactor, each rotary fuel column has each motor-lifter. The rotary fuel column has two types (Cylinder type and Heat Pipe type using natrium at 15 kPa which supply high temperature energy for Ultra Super Critical power plant). At 4 years cycle all rotary fuel columns of the reactor are renewed by the metallurgy method (vacuum re-smelting) and TRU fuel is gotten from the water fuel. (author)

  16. Study of a fuel assembly for the nuclear reactor of IV generation cooled with supercritical water; Estudio de un ensamble de combustible para el reactor nuclear de generacion IV enfriado con agua supercritica

    Energy Technology Data Exchange (ETDEWEB)

    Barragan M, A.; Martin del Campo M, C.; Francois L, J. L. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Mexico D. F. (Mexico); Espinosa P, G., E-mail: albrm29@yahoo.com [Universidad Autonoma Metropolitana, Unidad Iztapalapa, Area de Ingenieria en Recursos Energeticos, Av. San Rafael Atlixco 186, Col. Vicentina, 09340 Mexico D. F. (MX)

    2011-11-15

    In this work a neutron study is presented about a square assembly design of double line of fuel rods, with moderator box to the center of the arrangement, for a nuclear reactor cooled with supercritical water (SCWR). The SCWR reactor was chosen by the characteristics of its design, mainly because is based in light water reactors (PWR and BWR), and the operational experience that has of them allow to use models and similar programs to simulate the fuel and the nucleus of this type of reactors. To develop the necessary models and to carry out the design and analysis of the SCWR reactor, the neutron codes MCNPX and Helios were used. The reason of using both codes, is because the code MCNPX used thoroughly in the neutron simulation of these reactors, it has been our reference code to analyze the results obtained with the Helios code which results are more efficient because its calculation times are minors. In the nucleus design the same parameters for both codes were considered. The results show that the design with Helios is a viable option to simulate these reactors since their values of the neutrons multiplication factor are very similar to those obtained with MCNPX. On the other hand, it could be corroborated that the CASMO-4 code is inadequate to simulate the fuel to the temperature conditions and water pressure in the SCWR. (Author)

  17. Reconstitution of fuel assemblies and core components

    Energy Technology Data Exchange (ETDEWEB)

    Hummel, Wolfgang; Langenberger, Jan [AREVA NP GmbH (Germany)

    2012-11-01

    Due to AREVA's experience and big portfolio of techniques, reconstitution of fuel assemblies and core components at light water reactors is possible within a reasonable timeframe and with interesting cost benefit. Customer feedback indicates the sustainability of such reconstitutions. As a result, a long-term maintenance of value can be assured and early waste disposal can be avoided. (orig.)

  18. Conditioning of spent fuel assemblies from the Rossendorf RFR research reactor in transport and storage containers of the type CASTOR MTR 2

    International Nuclear Information System (INIS)

    Schneider, B.; Hofmann, G.

    1994-09-01

    Most of the spent fuel assemblies are temporarily stored in the flooded fuel ponds AB 1 and AB 2 of the RFR, and some are still in the reactor core. The conditioning task described here is part of the RFR spent fuel management concept and covers the safe emplacement of the spent fuel elements in the CASTOR MTR 2 shipping containers and the sealing of the containers in compliance with the nuclear licence issued for the conditioning task. The transfer of the spent fuel assemblies from the present wet storage conditions to the dry storage conditions in the CASTOR MTR 2 containers is done by a mobile manipulation equipment consisting essentially of the transfer sluice gate and a transfer container. Subsequent to conditioning, the shipping containers are to be transported to a licensed intermediate storage facility to await their transport to a national radwaste repository. The technical handling tools for the transfer and manipulation are briefly described, as well as the process steps involved, putting emphasis on the detailed description of processes and the accompanying time frame, so that the conditioning task can be incorporated into the work plan of the entire project. The report further presents the EDP concept established for the task, including the required data archivation and documentation. (orig.) [de

  19. BRET fuel assembly dismantling machine

    International Nuclear Information System (INIS)

    Titzler, P.A.; Bennett, K.L.; Kelley, R.S. Jr.; Stringer, J.L.

    1984-08-01

    An automated remote nuclear fuel assembly milling and dismantling machine has been designed, developed, and demonstrated at the Hanford Engineering Development Laboratory (HEDL) in Richland, Washington. The machine can be used to dismantle irradiated breeder fuel assemblies from the Fast Flux Test Facility prior to fuel reprocessing. It can be installed in an existing remotely operated shielded hot cell facility, the Fuels and Materials Examination Facility (FMEF), at the Hanford Site in Richland, Washington

  20. Water rod and fuel assembly

    International Nuclear Information System (INIS)

    Tsutsumi, Shinro; Tada, Nobuo; Nakajima, Junjiro; Aizawa, Yasuhiro.

    1995-01-01

    A water rod disposed in a fuel assembly comprises a larger diameter tube constituting an upwarding flow channel for coolants flown from the lower portion of a reactor core, and a smaller diameter tube connected fixedly to the larger diameter tube at the periphery of the upper end thereof and constituting a downwarding flow channel for coolants upwardly flown in the larger diameter tube. The larger diameter tube is formed by subjecting a base tube made of a zirconium alloy to PILGER mil fabrication and annealing in α region repeatingly for several times, then subjecting it to α + β treatment for once. The smaller diameter tube is formed by subjecting a base tube made of a zirconium alloy to PILGER mil fabrication and annealing in α region repeatingly for several times, then subjecting it to β treatment for once. With such procedures, the amount of irradiation growth of the tube in the axial direction is made greater in the larger diameter tube than that in the smaller diameter tube. Accordingly, since the smaller diameter tube is never bent by pressing, mechanical integrity of the fuel assembly is never lost. (I.N.)

  1. High burnup MOX fuel assembly

    International Nuclear Information System (INIS)

    Blanpain, P.; Brunel, L.

    1999-01-01

    From the outset, the MOX product was required to have the same performance as UO 2 in terms of burnup and operational flexibility. In fact during the first years the UO 2 managements could not be applied to MOX. The changeover to an AFA 2G type fuel allowed an improvement in NPP operational flexibility. The move to the AFA 3G design fuel will enable an increase in the burnup of the MOX assemblies to the level of the UO 2 ones ('MOX Parity' project). But the FRAMATOME fuel development objective does not stop at the obtaining of parity between the current MOX and UO 2 products: this parity must remain guaranteed and the MOX managements must evolve in the same way as the UO 2 managements. The goal of the MOX product development programmes underway with COGEMA and the CEA is the demonstration over the next 10 years of a fuel capable of reaching burnups of 70 GWD/T. The research programmes focus on the fission gas release aspect, with three issues explored: optimization of pellet microstructures and validation in experimental reactor ; build-up of experience feedback from fission gas release at elevated burnups in commercial reactors, both for current and experimental products; adaptation and qualification of the design models and tools, over the ranges and for the products concerned. The product arising from these development programmes should be offered on the market around 2010. While meeting safety requirements, it will cater for the needs of the utilities in terms of product reliability, personnel dosimetry and kWh output costs (increase in burnup, NPP maneuverability and availability, minimization of process waste). (authors)

  2. Nuclear fuel bundle disassembly and assembly tool

    International Nuclear Information System (INIS)

    Yates, J.; Long, J.W.

    1975-01-01

    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.)

  3. Establishment of China Nuclear Fuel Assembly Database

    International Nuclear Information System (INIS)

    Chen Peng; Jin Yongli; Zhang Yingchao; Lu Huaquan; Chen Jianxin

    2009-01-01

    China Nuclear Fuel Assembly Database (CNFAD) is developed based on Oracle system. It contains the information of fuel assemblies in the stages of its design, fabrication and post irradiation (PIE). The structure of Browser Sever is adopted in the development of the software, which supports the HTTP protocol. It uses Java interface to transfer the codes from server to clients and make the sources of server and clients be utilized reasonably and sufficiently, so it can perform complicated tasks. Data in various stages of the fuel assemblies in Pressure Water Reactor (PWR), such as the design,fabrication, operation, and post irradiation examination, can be stored in this database. Data can be shared by multi users and communicated within long distances. By using CNFAD, the problem of decentralization of fuel data in China nuclear power plants will be solved. (authors)

  4. Cermet fuel reactors

    International Nuclear Information System (INIS)

    Cowan, C.L.; Palmer, R.S.; Van Hoomissen, J.E.; Bhattacharyya, S.K.

    1987-01-01

    Cermet fueled nuclear reactors are attractive candidates for high performance space power systems. The cermet fuel consists of tungsten-urania hexagonal fuel blocks characterized by high strength at elevated temperatures, a high thermal conductivity and resultant high thermal shock resistance. The concept evolved in the 1960's with the objective of developing a reactor design which could be used for a wide range of mobile power generation systems including both Brayton and Rankine power conversion cycles. High temperature thermal cycling tests and in-reactor irradiation tests using cermet fuel were carried out by General Electric in the 1960's as part of the 710 Development Program and by Argonne National laboratory in a subsequent activity. Cermet fuel development programs are currently underway at Argonne National laboratory and Pacific Northwest Laboratory as part of the Multi-Megawatt Space Power Program. Key features of the cermet fueled reactor design are 1) the ability to achieve very high coolant exit temperatures, and 2) thermal shock resistance during rapid power changes, and 3) two barriers to fission product release - the cermet matrix and the fuel element cladding. Additionally, there is a potential for achieving a long operating life because of 1) the neutronic insensitivity of the fast-spectrum core to the buildup of fission products and 2) the utilization of a high strength refractory metal matrix and structural materials. These materials also provide resistance against compression forces that potentially might compact and/or reconfigure the core

  5. Fuel handling system of nuclear reactor plants

    International Nuclear Information System (INIS)

    Faulstich, D.L.

    1991-01-01

    This patent describes a fuel handing system for nuclear reactor plants comprising a reactor vessel having an openable top and removable cover for refueling and containing therein, submerged in coolant water substantially filling the reactor vessel, a fuel core including a multiplicity of fuel bundles formed of groups of sealed tube elements enclosing fissionable fuel assembled into units. It comprises a fuel bundle handing platform moveable over the open top of the reactor vessel; a fuel bundle handing mast extendable downward from the platform with a lower end projecting into the open top reactor vessel to the fuel core submerged in water; a grapple head mounted on the lower end of the mast provided with grappling hook means for attaching to and transporting fuel bundles into and out from the fuel core; and a camera with a prismatic viewing head surrounded by a radioactive resisting quartz cylinder and enclosed within the grapple head which is provided with at least three windows with at least two windows provided with an angled surface for aiming the camera prismatic viewing head in different directions and thereby viewing the fuel bundles of the fuel core from different perspectives, and having a cable connecting the camera with a viewing monitor located above the reactor vessel for observing the fuel bundles of the fuel core and for enabling aiming of the camera prismatic viewing head through the windows by an operator

  6. PWR fuel assembly

    International Nuclear Information System (INIS)

    Yamada, Yuji.

    1995-01-01

    A lower end plug is secured to a lower end of a thimble tube. A bolt-like thimble screw is screw-coupled and fastened to a female screw disposed to the end plug by way of a bushing screw-coupled to a lower nozzle. Then, the thimble screw and the lower nozzle are welded to secure the thimble tube and the lower nozzle. The lower portion of the bushing extends near the lower surface of the lower nozzle. The extended portion is provided with a recess to which a bolt head of the thimble screw is tightly inserted and a seating-face portion against which a seating-face of the bolt head abuts. Then, the extended portion of the bushing and the lower nozzle are spot-welded on the side of the lower surface of the nozzle, to prevent rotation of the bushing. This can easily prevent the rotation of the bushing after adjustment, to simplify the assembling of the fuel assembly. (I.N.)

  7. Calibration of spent fuel measurement assembly

    International Nuclear Information System (INIS)

    Koleska, Michal; Viererbl, Ladislav; Marek, Milan

    2014-01-01

    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 110m Ag isotope with known activity was performed. This isotope was chosen for having energies similar to isotopes measured in fuel assemblies. The 110m Ag 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 110m Ag isotope used for the measurement. • Calculated self-shielding factor for the IRT-2M fuel. • Applicable to other research reactor fuel geometries

  8. Assessment for the Applicability of Effective Thermal Conductivity Models on the Prismatic Fuel Assembly of Very High Temperature Reactor

    International Nuclear Information System (INIS)

    Shin, Dong-ho; Cho, Hyoung-kyu; Tak, Nam-il; Park, Goon-cherl

    2014-01-01

    A prismatic gas-cooled reactor is promising reactor type in the Nuclear Hydrogen Development and Demonstration (NHDD) project which was launched at KAERI (Korea Atomic Energy Research Institute). One of the most favorable characteristics of a prismatic gas-cooled reactor is its inherent and passive safety. As one of its inherent safety features, the heat flows through the prismatic core radially during the High Pressure Conduction Cooling (HPCC) or Low Pressure Conduction Cooling (LPCC) event and the radial heat transfer cools down the reactor core passively under such conditions. To verify the inherent safety of its design, the GAMMA+ code that is used to analyze VHTR thermo-fluid transients has been developed by KAERI. The code adopts effective thermal conductivity (ETC) model to analyze radial heat transfer in the core as a lumped parameter model. It is because the fuel block has complex geometry with large number of coolant holes and fuel compacts and the detail heat transfer calculations on that geometry needs excessive computation resources. GAMMA+ is adopting the Maxwell-based ETC model, however, there are several ETC models that could be applied to the GAMMA+ code. In this study, several ETC models will be introduced. They will be compared to CFD calculations which have similar condition with the fuel block. And then the most appropriate ETC model will be suggested for calculating the ETC of the fuel block. For the CFD calculation, unit cell tests with simple geometries were conducted. With unit cell test, the applicability of the ETC models were investigated. And proper ETC models were used to calculate the ETC of the fuel block and the results were compared to that of CFD calculation on the fuel block. In this study, the ETC models are introduced and the applicability of the ETC models to VHTR fuel block was investigated. The results of the ETC models were compared to those of CFD calculation. The CFD calculations were conducted for square graphite block

  9. LEU WWR-M2 fuel assemblies burnable test

    International Nuclear Information System (INIS)

    Kirsanov, G.A.; Konoplev, K.A.; Pikulik, R.G.; Sajkov, Yu. P.; Tchmshkyan, D.V.; Tedoradze, L.V.; Zakharov, A.S.

    2000-01-01

    The results of in-pile irradiation tests of LEU WWR-M2 fuel assemblies with reduced enrichment of fuel are submitted in the report. The tests are made according to the Russian Program on Reduced Enrichment for Research and Test Reactors (RERTR). United States Department of Energy and the Ministry of Atomic Energy of Russian Federation jointly fund this Program. The irradiation tests of 5 WWR-M2 experimental assemblies are carried out at WWR-M reactor of the Petersburg Nuclear Physics Institute (PNPI). The information on assembly design and technique of irradiation tests is presented. In the irradiation tests the integrity of fuel assemblies is periodically measured. The report presents the data for the integrity maintained during the burnup of 5 fuel assemblies up to 45%. These results demonstrate the high reliability of the experimental fuel assemblies within the guaranteed burnup limits specified by the manufacturer. The tests are still in progress; it is planned to test and analyze the change in integrity for burnup of up to 70% - 75% or more. LEU WWR-M2 fuel assemblies are to be offered for export by their Novosibirsk manufacturer. Currently, HEU WWR-M2 fuel assemblies are used in Hungary, Ukraine and Vietnam. LEU WWR-M2 fuel assemblies were designed as a possible replacement for the HEU WWR-M2 fuel assemblies in those countries, but their use can be extended to other research reactors. (author)

  10. Three-dimensional analysis of the coolant flow characteristics in the fuel assemblies of VVER-1000 reactors

    International Nuclear Information System (INIS)

    Dinh Van Thin; Tran Thi Nhan

    2015-01-01

    Computational Fluid Dynamics (CFD) is a widely used method around the world for complex flow and heat industrial problems. In this paper, the coolant flow parameters were investigated in subchannels of VVER-1000 reactor’s fuel assemblies by ANSYS V14.5 programme. The different mesh solutions and turbulence models were carried out to deal with the water flow problems such as velocity distribution, streamline, temperature and pressure change as well as the hydraulic resistances of the spacer grids. The obtained results are good agreement with the measured values and the published reports from other authors. (author)

  11. Determining initial enrichment, burnup, and cooling time of pressurized-water-reactor spent fuel assemblies by analyzing passive gamma spectra measured at the Clab interim-fuel storage facility in Sweden

    Energy Technology Data Exchange (ETDEWEB)

    Favalli, A., E-mail: afavalli@lanl.gov [Los Alamos National Laboratory, Los Alamos, NM (United States); Vo, D. [Los Alamos National Laboratory, Los Alamos, NM (United States); Grogan, B. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Jansson, P. [Uppsala University, Uppsala (Sweden); Liljenfeldt, H. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Mozin, V. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Schwalbach, P. [European Commission, DG Energy, Euratom Safeguards Luxemburg, Luxemburg (Luxembourg); Sjöland, A. [Swedish Nuclear Fuel and Waste Management Company, Stockholm (Sweden); Tobin, S.J.; Trellue, H. [Los Alamos National Laboratory, Los Alamos, NM (United States); Vaccaro, S. [European Commission, DG Energy, Euratom Safeguards Luxemburg, Luxemburg (Luxembourg)

    2016-06-01

    The purpose of the Next Generation Safeguards Initiative (NGSI)–Spent Fuel (SF) project is to strengthen the technical toolkit of safeguards inspectors and/or other interested parties. The NGSI–SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins; (3) estimate the plutonium mass [which is also a function of the variables in (1)]; (4) estimate the decay heat; and (5) determine the reactivity of spent fuel assemblies. Since August 2013, a set of measurement campaigns has been conducted at the Central Interim Storage Facility for Spent Nuclear Fuel (Clab), in collaboration with Swedish Nuclear Fuel and Waste Management Company (SKB). One purpose of the measurement campaigns was to acquire passive gamma spectra with high-purity germanium and lanthanum bromide scintillation detectors from Pressurized Water Reactor and Boiling Water Reactor spent fuel assemblies. The absolute {sup 137}Cs count rate and the {sup 154}Eu/{sup 137}Cs, {sup 134}Cs/{sup 137}Cs, {sup 106}Ru/{sup 137}Cs, and {sup 144}Ce/{sup 137}Cs isotopic ratios were extracted; these values were used to construct corresponding model functions (which describe each measured quantity’s behavior over various combinations of burnup, cooling time, and initial enrichment) and then were used to determine those same quantities in each measured spent fuel assembly. The results obtained in comparison with the operator declared values, as well as the methodology developed, are discussed in detail in the paper.

  12. Determining initial enrichment, burnup, and cooling time of pressurized-water-reactor spent fuel assemblies by analyzing passive gamma spectra measured at the Clab interim-fuel storage facility in Sweden

    Science.gov (United States)

    Favalli, A.; Vo, D.; Grogan, B.; Jansson, P.; Liljenfeldt, H.; Mozin, V.; Schwalbach, P.; Sjöland, A.; Tobin, S. J.; Trellue, H.; Vaccaro, S.

    2016-06-01

    The purpose of the Next Generation Safeguards Initiative (NGSI)-Spent Fuel (SF) project is to strengthen the technical toolkit of safeguards inspectors and/or other interested parties. The NGSI-SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins; (3) estimate the plutonium mass [which is also a function of the variables in (1)]; (4) estimate the decay heat; and (5) determine the reactivity of spent fuel assemblies. Since August 2013, a set of measurement campaigns has been conducted at the Central Interim Storage Facility for Spent Nuclear Fuel (Clab), in collaboration with Swedish Nuclear Fuel and Waste Management Company (SKB). One purpose of the measurement campaigns was to acquire passive gamma spectra with high-purity germanium and lanthanum bromide scintillation detectors from Pressurized Water Reactor and Boiling Water Reactor spent fuel assemblies. The absolute 137Cs count rate and the 154Eu/137Cs, 134Cs/137Cs, 106Ru/137Cs, and 144Ce/137Cs isotopic ratios were extracted; these values were used to construct corresponding model functions (which describe each measured quantity's behavior over various combinations of burnup, cooling time, and initial enrichment) and then were used to determine those same quantities in each measured spent fuel assembly. The results obtained in comparison with the operator declared values, as well as the methodology developed, are discussed in detail in the paper.

  13. Cermet fuel reactors

    International Nuclear Information System (INIS)

    Cowan, C.L.; Palmer, R.S.; Van Hoomissen, J.E.; Bhattacharyya, S.K.; Barner, J.O.

    1987-09-01

    Cermet fueled nuclear reactors are attractive candidates for high performance space power systems. The cermet fuel consists of tungsten-urania hexagonal fuel blocks characterized by high strength at elevated temperatures, a high thermal conductivity and resultant high thermal shock resistance. Key features of the cermet fueled reactor design are (1) the ability to achieve very high coolant exit temperatures, and (2) thermal shock resistance during rapid power changes, and (3) two barriers to fission product release - the cermet matrix and the fuel element cladding. Additionally, thre is a potential for achieving a long operating life because of (1) the neutronic insensitivity of the fast-spectrum core to the buildup of fission products and (2) the utilization of a high strength refractory metal matrix and structural materials. These materials also provide resistance against compression forces that potentially might compact and/or reconfigure the core. In addition, the neutronic properties of the refractory materials assure that the reactor remains substantially subcritical under conditions of water immersion. It is concluded that cermet fueled reactors can be utilized to meet the power requirements for a broad range of advanced space applications. 4 refs., 4 figs., 3 tabs

  14. Analysis of a total flow blockage of a Fuel Assembly in a typical MTR Research Reactor by RELAP5/MOD3.3

    International Nuclear Information System (INIS)

    Adorni, M.; Salah, A.B.; Di Maro, B.; Pierro, F.; D'Auria, F.; Hamidouche, T.

    2004-01-01

    The lack of full understanding of complex mechanisms connected with the interaction between thermal-hydraulics and neutronics still challenge the design and the operation of nuclear reactors by the adoption of conservative safety limits. The recent availability of powerful computer and computational techniques together with the continuing increase in operational experience imposes the revisiting of those areas and eventually the identification of design/safety requirements that can be relaxed [1]. Currently, the enlarged commercial exploitation of nuclear Research Reactors (RR) has increased the consideration to their corresponding safety issues. Almost all of the safety analyses have so far been performed using conservative computational tools [2]. Nowadays, the application of Best-Estimate (BE) methods constitutes a real necessity in order to increase their commercial productivity. In this framework, an attempt is made to apply the BE technique to perform a safety evaluation under research reactors operational conditions. In fact, this technique has been largely verified and validated for power reactors using coupled system thermal-hydraulic and three-dimensional neutron kinetics [1]. For this purpose, as typical representative of research reactors, the IAEA 10 MW MTR Research Reactors problem [3] is considered. The system thermal-hydraulic RELAP5 [4] code was developed to simulate transient scenarios in Power reactors such PWR, BWR, VVER, etc. However, only limited work was performed to access the applicability of the code to Research Reactors operating conditions (low pressure, mass flow rates, power, etc) [5]. Previous works performed in this field are reported in [5], [6] and [7]. In this framework, total and partial blockage of a single Fuel Assembly cooling channel are investigated. As a first attempt the calculations are performed by applying the BE thermal-hydraulic system code RELAP5 alone using its point kinetic model to derive the instantaneous core

  15. History of research reactor fuel fabrication at Babcock and Wilcox

    International Nuclear Information System (INIS)

    Freim, James B.

    1983-01-01

    B and W Research Reactor Fuel Element facility at Lynchburg, Virginia now produces national laboratory and university fuel assemblies. The Company's 201000 square foot facility is devoted entirely to supplying research fuel and related products. B and W re-entered the research reactor fuel market in 1981

  16. Nuclear fuel assemblies and fuel pins usable in such assemblies

    International Nuclear Information System (INIS)

    Jolly, R.

    1982-01-01

    A novel end cap for a nuclear fuel assembly is described in detail. It consists of a trisection arrangement which is received within a cell of a cellular grid. The cell contains abutment means with which the trisection comes into abutment. The grid also contains an abutment means for preventing the trisections from being inserted into the cell in an incorrect orientation. The present design allows fuel pins to be securely held in a hold-down grid of a sub-assembly. The design also allows easier dis-assembly of the swollen and embrittled fuel pins prior to reprocessing. (U.K.)

  17. Gaseous fuel reactors for power systems

    Science.gov (United States)

    Kendall, J. S.; Rodgers, R. J.

    1977-01-01

    Gaseous-fuel nuclear reactors have significant advantages as energy sources for closed-cycle power systems. The advantages arise from the removal of temperature limits associated with conventional reactor fuel elements, the wide variety of methods of extracting energy from fissioning gases, and inherent low fissile and fission product in-core inventory due to continuous fuel reprocessing. Example power cycles and their general performance characteristics are discussed. Efficiencies of gaseous fuel reactor systems are shown to be high with resulting minimal environmental effects. A technical overview of the NASA-funded research program in gaseous fuel reactors is described and results of recent tests of uranium hexafluoride (UF6)-fueled critical assemblies are presented.

  18. Fusion reactor fuel processing

    International Nuclear Information System (INIS)

    Johnson, E.F.

    1972-06-01

    For thermonuclear power reactors based on the continuous fusion of deuterium and tritium the principal fuel processing problems occur in maintaining desired compositions in the primary fuel cycled through the reactor, in the recovery of tritium bred in the blanket surrounding the reactor, and in the prevention of tritium loss to the environment. Since all fuel recycled through the reactor must be cooled to cryogenic conditions for reinjection into the reactor, cryogenic fractional distillation is a likely process for controlling the primary fuel stream composition. Another practical possibility is the permeation of the hydrogen isotopes through thin metal membranes. The removal of tritium from the ash discharged from the power system would be accomplished by chemical procedures to assure physiologically safe concentration levels. The recovery process for tritium from the breeder blanket depends on the nature of the blanket fluids. For molten lithium the only practicable possibility appears to be permeation from the liquid phase. For molten salts the process would involve stripping with inert gas followed by chemical recovery. In either case extremely low concentrations of tritium in the melts would be desirable to maintain low tritium inventories, and to minimize escape of tritium through unwanted permeation, and to avoid embrittlement of metal walls. 21 refs

  19. Modernization in-core monitoring system, of WWER-1000 reactors (W-320) by fuel assemblies with individual characteristics using

    International Nuclear Information System (INIS)

    Mitin, V. I.; Semchenkov, J. M.; Kalinushkin, A. E.

    2007-01-01

    The present report covers object, conception, engineering solution of construction of modern system of high-powered reactor in-core control, including WWER-1000 (V-320) reactors. It is known that ICMS main task is on-line monitoring distribution of power release field and its functioning independently of design programs to avoid common reason error. It is shown in what way field of power release recovery has been carrying on; rest on the signals of in-core neutron and temperature sensors. On the base of the obtained and refined information there have being automatically generated signals of preventive and emergency protection on local parameters (linear power to the maximum intensive fuel elements, reserve to heat exchange crisis, 'picking factor'). There have represented technology on sensors and processing methods of SPND and TC signals, ICIS composition and structure, program hard ware, system and applied software. Structure, composition and the taken decisions allow combining class 1E and class B and C tasks in accordance with international norms of separation and safety classes' realization. At present ICIS-M is a system, providing implementation of control, safety, information and diagnostic functions, which allow securing actual increase of quality, reliability and safety in operation of nuclear fuel and NPP units. And at the same time it reduces human factor negative influence to core work thermo technical eliability in the operational process (Authors)

  20. Method for calculating the forces and deformations in the fast reactor fuel assembly accounting for the effects of reactor control system elements and shutdown

    International Nuclear Information System (INIS)

    Likhachev, Yu.I.; Vashlyaev, Yu.N.; Kravchenko, I.N.

    1980-01-01

    Methods for calculating deformations and interaction forces of heat-generating assemblies (HGA) of fast reactor core with account for the effect of control and protection system (CPS) elements at the reactor operation and change of interaction efforts between HGA at the reactor shutdown, are described. The results of testing the suggested methods on example of estimate of HGA behaviour of the BN-350 reactor are presented. For estimating the effect of CPS elements on HGA bending the sector model has been used. It is assumed that HGA deformation inside each sector is independent of HGA deformation of other sectors. A higher calculation accuracy is attained by means of laying out of sectors into regions of preferable influence of emergency protection elements and compensating packets. When determining deformation and interaction efforts between HGA caused by temperature change in the course of shutdown it is supposed that the HGA deformation is purely elastic. The methods described are realized in the form of ABRI-CPS and ABRI-HOL programs written in FORTRAN for the BESM-6 computer. The results of HGA calculations of the BN-350 reactor core show that CPS elements decrease contact efforts in the middle of the central packet, increase contact efforts in the peak of the central packet, increase contact efforts in the peaks of packets from the eight row to the periphery and increase contact efforts in the middles of packets from the 5th to 9th row [ru

  1. Spent nuclear fuel discharges from U.S. reactors 1994

    International Nuclear Information System (INIS)

    1996-02-01

    Spent Nuclear Fuel Discharges from US Reactors 1994 provides current statistical data on fuel assemblies irradiated at commercial nuclear reactors operating in the US. This year's report provides data on the current inventories and storage capacities at these reactors. Detailed statistics on the data are presented in four chapters that highlight 1994 spent fuel discharges, storage capacities and inventories, canister and nonfuel component data, and assembly characteristics. Five appendices, a glossary, and bibliography are also included. 10 figs., 34 tabs

  2. Pattern fuel assembly loading system

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  3. Advanced fuel cycles of WWER-1000 reactors

    International Nuclear Information System (INIS)

    Lunin, G.; Novikov, A.; Pavlov, V.; Pavlovichev, A.

    2003-01-01

    The present paper considers characteristics of fuel cycles for the WWER-1000 reactor satisfying the following conditions: duration of the campaign at the nominal power is extended from 250 EFPD up to 470 and more ones; fuel enrichment does not exceed 5 wt.%; fuel assemblies maximum burnup does not exceed 55 MWd/kgHM. Along with uranium fuel, the use of mixed Uranium-Plutonium fuel is considered. Calculations were conducted by codes TVS-M, BIPR-7A and PERMAK-A developed in the RRC Kurchatov Institute, verified for the calculations of uranium fuel and certified by GAN RF

  4. A study on 80 fuel assemblies core for HFETR

    International Nuclear Information System (INIS)

    Sun Shouhua; Wu Yinghua; Bu Yongxi; Liu Shuiqing; Duan Tianyuan; Zhang Liangwan; Lin Jisen

    1996-12-01

    The performance of 80 and 60 fuel assemblies cores for High Flux Engineering Test Reactor (HFETR) has been compared with theoretical analysis and operating results. These results show that the core performance of 80 fuel assemblies is the same as that of 60 fuel assemblies in the following aspects: the permission power of core, the irradiation test of materials, the transmutation doping of single crystalline silicon, the production of Mo-Tc isotopes, etc. The core of 80 fuel assemblies is more convenient in operation after 500 kw test loop installed, and in greatly raising the production of 60 Co source with high specific radioactivity and the usage of fuel. As compared to the production of 60 Co source of 60 fuel assemblies core, the benefit of 80 fuel assemblies core can increase more than 3.8 millions RMB yuan per year. (2 refs., 2 tabs.)

  5. Appearance detection device for fuel assembly

    International Nuclear Information System (INIS)

    Matsuoka, Toshihiro

    1998-01-01

    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. Determining the axial power profile of partly flooded fuel in a compact core assembled in reactor LR-0

    International Nuclear Information System (INIS)

    Košťál, Michal; Švadlenková, Marie; Baroň, Petr; Rypar, Vojtěch; Milčák, Ján

    2016-01-01

    Highlights: • Fission density in partly flooded compact core. • Calculation of fission density axial profile. • Significant calculational under prediction of experimental axial profile. - Abstract: Measurement and calculation of the axial power profile near the boundary of a moderated and non-moderated core is used to analyze the suitability of the neutron-physical process description, mainly the angular cross-section of a water-moderated uranium system. This is also an important issue because it affects the radiation situation above the partly flooded core of a water-moderated reactor. Axial power profiles of various fuel pins irradiated on reactor LR-0 were measured and the results were compared with MCNP6 code calculations using the ENDF/B-VII.0 nuclear data library. The calculated power profile in positions above the moderator level significantly underestimates experimental results. This might be caused by an improper description of the angular distribution of scattered neutrons in a water-moderated uranium system.

  7. Advanced fuel in the Budapest research reactor

    International Nuclear Information System (INIS)

    Hargitai, T.; Vidovsky, I.

    1997-01-01

    The Budapest Research Reactor, the first nuclear facility of Hungary, started to operate in 1959. The main goal of the reactor is to serve neutron research, but applications as neutron radiography, radioisotope production, pressure vessel surveillance test, etc. are important as well. The Budapest Research Reactor is a tank type reactor, moderated and cooled by light water. After a reconstruction and upgrading in 1967 the VVR-SM type fuel elements were used in it. These fuel elements provided a thermal power of 5 MW in the period 1967-1986 and 10 MW after the reconstruction from 1992. In the late eighties the Russian vendor changed the fuel elements slightly, i.e. the main parameters of the fuel remained unchanged, however a higher uranium content was reached. This new fuel is called VVR-M2. The geometry of VVR-SM and VVR-M2 are identical, allowing the use to load old and new fuel assemblies together to the active core. The first new type fuel assemblies were loaded to the Budapest Research Reactor in 1996. The present paper describes the operational experience with the new type of fuel elements in Hungary. (author)

  8. Storage experience in Hungary with fuel from research reactors

    International Nuclear Information System (INIS)

    Gado, J.; Hargitai, T.

    1996-01-01

    In Hungary several critical assemblies, a training reactor and a research reactor have been in operation. The fuel used in the research and training reactors are of Soviet origin. Though spent fuel storage experience is fairly good, medium and long term storage solutions are needed. (author)

  9. Modular nuclear fuel assembly rack

    International Nuclear Information System (INIS)

    Davis, C.J.

    1982-01-01

    A modular nuclear fuel assembly rack constructed of an array of identical cells, each cell constructed of a plurality of identical flanged plates. The unique assembly of the plates into a rigid rack provides a cellular compartment for nuclear fuel assemblies and a cavity between the cells for accepting neutron absorbing materials thus allowing a closely spaced array. The modular rack size can be easily adapted to conform with available storage space. U-shaped flanges at the edges of the plates are nested together at the intersection of four cells in the array. A bar is placed at the intersection to lock the cells together

  10. Holddown device for nuclear fuel assembly

    International Nuclear Information System (INIS)

    Anthony, A.J.

    1978-01-01

    An apparatus for preventing ''floating'' of nuclear-reactor fuel assemblies due to hydraulic forces is disclosed. The apparatus uses a holddown column made of the same material as the core barrel. The column is positioned in a center guide-tube location in the fuel assembly in such a manner as to enable it either to slide within the center guide tube or, if the center guide tube is replaced by the column, to slide through openings in the spacer grids. The lower end of the holddown column engages the lower end fitting of the fuel assembly, and the upper end of the column engages a flow plate to which holddown force is applied. As a consequence of this arrangement, holddown force is transmitted from the flow plate through the holddown column to the lower end fitting. Movement of the fuel assembly is thereby prevented without a compression load being applied to the fuel-assemb1ly structure. In addition, variations due to thermal expansion in the distance between the lower core plate and the upper core plate are largely made up for by corresponding variations in the holddown column because the holddown column and the core barrel can be made of the same material

  11. Polarized advanced fuel reactors

    International Nuclear Information System (INIS)

    Kulsrud, R.M.

    1987-07-01

    The d- 3 He reaction has the same spin dependence as the d-t reaction. It produces no neutrons, so that if the d-d reactivity could be reduced, it would lead to a neutron-lean reactor. The current understanding of the possible suppression of the d-d reactivity by spin polarization is discussed. The question as to whether a suppression is possible is still unresolved. Other advanced fuel reactions are briefly discussed. 11 refs

  12. Assembly mechanism for nuclear fuel bundles

    International Nuclear Information System (INIS)

    Long, J.W.; Flora, B.S.; Ford, K.L.

    1980-01-01

    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)

  13. Natural circulation in simulated LMFBR fuel assemblies

    International Nuclear Information System (INIS)

    Levin, A.E.; Carbajo, J.J.; Lloyd, D.B.; Montgomery, B.H.; Rose, S.D.; Wantland, J.L.

    1985-01-01

    Natural circulation experiments have been performed using simulated liquid metal fast breeder reactor fuel assemblies in the Thermal-Hydraulic Out-of-Reactor Safety (THORS) facility, an engineering-scale sodium loop. Objective of these tests has been to provide experimental data under conditions that might be encountered during a partial or total loss of the shutdown heat removal system (SHRS) in a reactor. The experiments have included single- and two-phase tests under quasi-steady and transient conditions, at both nominal and non-nominal system conditions. Results from these test indicate that the potential for reactor damage during degraded SHRS operation is extremely slight, and that natural circulation can be a major contributor to safe operation of the system in both single- and two-phase flow during such operation

  14. Status of research reactor spent fuel world-wide

    International Nuclear Information System (INIS)

    Ritchie, I.G.

    2004-01-01

    Results compiled in the research reactor spent fuel database are used to assess the status of research reactor spent fuel world-wide. Fuel assemblies, their types, enrichment, origin of enrichment and geological distribution among the industrialised and developed countries of the world are discussed. Fuel management practices in wet and dry storage facilities and the concerns of reactor operators about long-term storage of their spent fuel are presented and some of the activities carried out by the International Atomic Energy Agency to address the issues associated with research reactor spent fuel are outlined. (author)

  15. A study of the friction and wear processes of the structural components of fuel assemblies for water-cooled and water moderated power reactors

    International Nuclear Information System (INIS)

    Makarov, V.; Afanasiev, A.; Matvienko, I.; Drozdov, Y.; Puchkov, V.

    2011-01-01

    The friction forces affect the fuel assembly (FA) strength at all the stages of its lifecycle. The paper covers the methods and the results of the pre-irradiation experimental studies of the static and dynamic processes the friction forces are involved in. These comprise the FA assembling at the manufacturer, fuel rod flow-induced vibration and fretting-wear in the fuel rod-to-cell friction pairs, rod cluster control assembly (RCCA) movement in the FA guide tubes, FA bowing, FA loading-unloading into the core, irradiation-induced growth and thermal-mechanical fuel rod-to-spacer grid interaction. (authors)

  16. Cleaning device for fuel assemblies

    International Nuclear Information System (INIS)

    Kita, Kaoru.

    1986-01-01

    Purpose: To completely remove obstacles deposited to the lower sides of supporting lattices for fuel assemblies by utilizing water within a pit before reloading of the fuel assemblies. Constitution: A cylindrical can, to which a fuel assembly is inserted through the upper end opening thereof, is vertically disposed within water of a pit and the bottom of the can is communicated with a pump by way of a suction pipe and a filter device disposed out of the pit. While on the other hand, a fuel assembly is suspended downwardly by a crane and inserted to the inside of the can through the upper end of the opening thereof and supported therein followed by starting the pump. As a result, water in the pit is circulated through the inside of the can, suction pipe, filtering device, pump, discharge pipe and to the inside of the pit thereby enabling to completely eliminate obstacles deposited to the lower surface, etc. of supporting lattices for the fuel assembly supported within the can. (Takahashi, M.)

  17. Container for spent fuel assembly

    International Nuclear Information System (INIS)

    Sawai, Takeshi.

    1996-01-01

    The container of the present invention comprises a container main body having a body portion which can contain spent fuel assemblies and a lid, and heat pipes having an evaporation portion disposed along the outer surface of the spent fuel assemblies to be contained and a condensation portion exposed to the outside of the container main body. Further, the heat pipe is formed spirally at the evaporation portions so as to surround the outer circumference of the spent fuel assemblies, branched into a plurality of portions at the condensation portion, each of the branched portion of the condensation portion being exposed to the outside of the container main body, and is tightly in contact with the periphery of the slit portions disposed to the container main body. Then, since released after heat is transferred to the outside of the container main body from the evaporation portion of the heat pipe along the outer surface of the spent fuel assemblies by way of the condensation portion of the heat pipes exposed to the outside of the container main body, the efficiency of the heat transfer is extremely improved to enhance the effect of removing heat of spent fuel assemblies. Further, cooling effect is enhanced by the spiral form of the evaporation portion and the branched condensation portion. (N.H.)

  18. Post-irradiation examination of the vipac fuel assemblies IFA-104 and IFA-203, irradiated in the Halden boiling water reactor

    International Nuclear Information System (INIS)

    Linde, A. van der; Lucas Luyckx, H.J.B.

    1976-11-01

    Two seven pin assemblies, IFA-104 and IFA-203, have been irradiated in the Halding Boiling Water Reactor. The Zircaloy clad pins contained vibrationally compacted, sharp edged UO 2 particles with smear densities of 84-88% theoretical density and a fuel stack length of about 1520 mm. The IFA-104 pins operated satisfactorily during approximately 240 effective full power days at an average linear heat rating of about 340 W/cm, and a peak rating of about 480 W/cm. Ingress of water via a pressure transducer into one of the IFA-203 pins necessitated the premature termination of the irradiation after approximately 30 effective full power days at an average rating of about 460 W/cm, and a peak rating of about 550 W/cm. One IFA-104 pin and two IFA-203 pins failed early. The primary cause of these failures has been sought in the oxidation reaction of the tantalum tube which protected the W/Re thermocouples with the surrounding UO 2 or with moisture in the fuel at temperatures in excess of approximately 1700degC. The fuel centre temperature and the internal gas pressure measurements during the beginning of lifetime and the results of the post-irradiation chemical and microscopic analysis on fuel pin cross-sections have been correlated with corresponding data calculated on a Gapcon-Thermal computer programme written for pellet fuel pins. Fairly good agreement could be achieved between the measured data on the vipac pins and the calculated data of a pellet pin. During the corrosion of the Zircaloy-2 cladding of the IFA-104 pins which were autoclaved in 400degC steam prior to the irradiation, relatively large amounts of H 2 entered the cladding from the D 2 O coolant side. With 0.66 mole% H 2 O in the D 2 O the calculated hydrogen uptake preference ratio was 33

  19. Advanced research reactor fuel development

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Chang Kyu; Pak, H. D.; Kim, K. H. [and others

    2000-05-01

    The fabrication technology of the U{sub 3}Si fuel dispersed in aluminum for the localization of HANARO driver fuel has been launches. The increase of production yield of LEU metal, the establishment of measurement method of homogeneity, and electron beam welding process were performed. Irradiation test under normal operation condition, had been carried out and any clues of the fuel assembly breakdown was not detected. The 2nd test fuel assembly has been irradiated at HANARO reactor since 17th June 1999. The quality assurance system has been re-established and the eddy current test technique has been developed. The irradiation test for U{sub 3}Si{sub 2} dispersed fuels at HANARO reactor has been carried out in order to compare the in-pile performance of between the two types of U{sub 3}Si{sub 2} fuels, prepared by both the atomization and comminution processes. KAERI has also conducted all safety-related works such as the design and the fabrication of irradiation rig, the analysis of irradiation behavior, thermal hydraulic characteristics, stress analysis for irradiation rig, and thermal analysis fuel plate, for the mini-plate prepared by international research cooperation being irradiated safely at HANARO. Pressure drop test, vibration test and endurance test were performed. The characterization on powders of U-(5.4 {approx} 10 wt%) Mo alloy depending on Mo content prepared by rotating disk centrifugal atomization process was carried out in order to investigate the phase stability of the atomized U-Mo alloy system. The {gamma}-U phase stability and the thermal compatibility of atomized U-16at.%Mo and U-14at.%Mo-2at.%X(: Ru, Os) dispersion fuel meats at an elevated temperature have been investigated. The volume increases of U-Mo compatibility specimens were almost the same as or smaller than those of U{sub 3}Si{sub 2}. However the atomized alloy fuel exhibited a better irradiation performance than the comminuted alloy. The RERTR-3 irradiation test of nano

  20. Disassembling and rebuilding 900 MW unit fuel assemblies in Celimene

    International Nuclear Information System (INIS)

    Giquel, G.; Leseur, A.; Pillet, C.; Van Craeynest, J.C.

    1987-01-01

    The Celimene high activity laboratory, in the Nuclear Research Centre of Saclay, has equipment for and experience of disassembling and rebuilding fuel assemblies from 900 MW light water reactors. These operations have been performed for R and D purposes; they allow removal for investigation of some of the fuel rods and examination of the skeleton. The rebuilt assemblies are sent to the fuel reprocessing plant. Reirradiation of these assemblies has not been considered so far and would require modifications of the procedure and of parts of the new skeleton. Disassembling and rebuilding have already been performed on three assemblies and a fourth one will be rebuilt in the coming months [fr

  1. Device for identifying fuel assembly

    International Nuclear Information System (INIS)

    Imai, Tetsuo; Miyazawa, Tatsuo.

    1982-01-01

    Purpose: To accurately identify a symbol printed on a hanging tool at the upper part of a fuel assembly. Constitution: Optical fibers are bundled to prepare a detector which is disposed at a predetermined position on a hanging tool. This position is set by a guide. Thus, the light emitted from an illumination lamp arrives at the bottom of a groove printed on the upper surface of the tool, and is divided into a weak light reflected upwardly and a strong light reflected on the surface lower than the groove. When these lights are received by the optical fibers, the fibers corresponding to the grooved position become dark, and the fibers corresponding to the ungrooved position become bright. Since the fuel assembly is identified by the dark and bright of the optical fibers as symbols, different machining can be performed every fuel assembly on the upper surface of the tool. (Yoshihara, H.)

  2. Reactor fuel exchanging facility

    International Nuclear Information System (INIS)

    Kubota, Shin-ichi.

    1981-01-01

    Purpose: To enable operation of an emergency manual operating mechanism for a fuel exchanger with all operatorless trucks and remote operation of a manipulator even if the exchanger fails during the fuel exchanging operation. Constitution: When a fuel exchanging system fails while connected to a pressure tube of a nuclear reactor during a fuel exchanging operation, a stand-by self-travelling truck automatically runs along a guide line to the position corresponding to the stopping position at that time of the fuel exchanger based on a command from a central control chamber. At this time the truck is switched to manual operation, and approaches the exchanger while being monitored through a television camera and then stops. Then, a manipurator is connected to the emergency manual operating mechanism of the exchanger, and is operated through necessary emergency steps by driving the snout, the magazine, the grab or the like in the exchanger in response to the problem, and necessary operations for the emergency treatment are thus performed. (Sekiya, K.)

  3. Weight simulation fuel assembly

    International Nuclear Information System (INIS)

    Sumikawa, Kiyokazu; Tokomatsu, Mutsuo.

    1993-01-01

    A tungsten pellet is not applied with hollow fabrication but a tungsten rod is inserted and filled into a zircaloy fuel cladding tube, as well as different kind of material having a density lower than that of tungsten, for example, stainless steel rods, are disposed successively intermittently and alternately for simulating the weight of one fuel rod. The filling method and the length of the individual pellets are optional depending on the method of usage, providing that the outer diameter of the simulation pellet is made identical with that of the actual fuel pellet. With such a constitution, there is no need to dispose a hollow portion as in the case of using only tungsten pellets, and the costs for both the materials and the fabrication can be saved. (T.M.)

  4. Irradiation experience of IPEN fuel at IEA-R1 research reactor

    International Nuclear Information System (INIS)

    Perrotta, Jose A.; Neto, Adolfo; Durazzo, Michelangelo; Souza, Jose A.B. de; Frajndlich, Roberto

    1998-01-01

    IPEN/CNEN-SP produces, for its IEA-R1 Research Reactor, MTR fuel assemblies based on U 3 O 8 -Al dispersion fuel type. Since 1985 a qualification program on these fuel assemblies has been performed. Average 235 U burnup of 30% and peak burnup of 50% was already achieved by these fuel assemblies. This paper presents some results acquire, by these fuel assemblies, under irradiation at IEA-R1 Research Reactor. (author)

  5. Fossil fuel furnace reactor

    Science.gov (United States)

    Parkinson, William J.

    1987-01-01

    A fossil fuel furnace reactor is provided for simulating a continuous processing plant with a batch reactor. An internal reaction vessel contains a batch of shale oil, with the vessel having a relatively thin wall thickness for a heat transfer rate effective to simulate a process temperature history in the selected continuous processing plant. A heater jacket is disposed about the reactor vessel and defines a number of independent controllable temperature zones axially spaced along the reaction vessel. Each temperature zone can be energized to simulate a time-temperature history of process material through the continuous plant. A pressure vessel contains both the heater jacket and the reaction vessel at an operating pressure functionally selected to simulate the continuous processing plant. The process yield from the oil shale may be used as feedback information to software simulating operation of the continuous plant to provide operating parameters, i.e., temperature profiles, ambient atmosphere, operating pressure, material feed rates, etc., for simulation in the batch reactor.

  6. Manufacturing method for fuel assembly

    International Nuclear Information System (INIS)

    Yamaguchi, Takashi.

    1997-01-01

    In an FBR type reactor, uranium/plutonium mixed oxide fuels (MOX fuels) are used. Nuclear fuel materials containing uranium and plutonium are filled to a portion or all of a plurality of fuel rods. In this case, an equivalent fissile coefficient (B) based on a combustion guarantee method defined by the formula: (B) = (M) · (F) is determined. (M) is a combustion matrix constituted based on the solution of equation of combustion which is a differential equation representing change with time of each of nuclear fuel materials during combustion. (F) is an equivalent fissile coefficient based on a reactivity keeping method which is a coefficient representing a reactivity worth equivalent with plutonium-239. The content of each of the nuclear fuel materials is determined so that the effective multiplication factor at the final stage of the operation cycle is substantially constant by using the equivalent fissile coefficient (B) based on the combustion guarantee method. (I.N.)

  7. Cooling nuclear reactor fuel

    International Nuclear Information System (INIS)

    Porter, W.H.L.

    1975-01-01

    Reference is made to water or water/steam cooled reactors of the fuel cluster type. In such reactors it is usual to mount the clusters in parallel spaced relationship so that coolant can pass freely between them, the coolant being passed axially from one end of the cluster in an upward direction through the cluster and being effective for cooling under normal circumstances. It has been suggested, however, that in addition to the main coolant flow an auxiliary coolant flow be provided so as to pass laterally into the cluster or be sprayed over the top of the cluster. This auxiliary supply may be continuously in use, or may be held in reserve for use in emergencies. Arrangements for providing this auxiliary cooling are described in detail. (U.K.)

  8. Recent BWR fuel management reactor physics advances

    International Nuclear Information System (INIS)

    Crowther, R.L.; Congdon, S.P.; Crawford, B.W.; Kang, C.M.; Martin, C.L.; Reese, A.P.; Savoia, P.J.; Specker, S.R.; Welchly, R.

    1982-01-01

    Improvements in BWR fuel management have been under development to reduce uranium and separative work (SWU) requirements and reduce fuel cycle costs, while also maintaining maximal capacity factors and high fuel reliability. Improved reactor physics methods are playing an increasingly important role in making such advances feasible. The improved design, process computer and analysis methods both increase knowledge of the thermal margins which are available to implement fuel management advance, and improve the capability to reliably and efficiently analyze and design for fuel management advances. Gamma scan measurements of the power distributions of advanced fuel assembly and advanced reactor core designs, and improved in-core instruments also are important contributors to improving 3-d predictive methods and to increasing thermal margins. This paper is an overview of the recent advances in BWR reactor physics fuel management methods, coupled with fuel management and core design advances. The reactor physics measurements which are required to confirm the predictions of performance fo fuel management advances also are summarized

  9. Nuclear reactor internals construction and failed fuel rod detection system

    International Nuclear Information System (INIS)

    Frisch, E.; Andrews, H.N.

    1976-01-01

    A system is provided for determining during operation of a nuclear reactor having fluid pressure operated control rod mechanisms the exact location of a fuel assembly with a defective fuel rod. The construction of the reactor internals is simplified in a manner to facilitate the testing for defective fuel rods and the reduce the cost of producing the upper internals of the reactor. 13 claims, 10 drawing figures

  10. Development of nuclear fuel for integrated reactor

    Energy Technology Data Exchange (ETDEWEB)

    Song, Kee Nam; Kim, H. K.; Kang, H. S.; Yoon, K. H.; Chun, T. H.; In, W. K.; Oh, D. S.; Kim, D. W.; Woo, Y. M

    1999-04-01

    The spacer grid assembly which provides both lateral and vertical support for the fuel rods and also provides a flow channel between the fuel rods to afford the heat transfer from the fuel pellet into the coolant in a reactor, is one of the major structural components of nuclear fuel for LWR. Therefore, the spacer grid assembly is a highly ranked component when the improvement of hardware is pursued for promoting fuel performance. Main objective of this project is to develop the inherent spacer grid assembly and to research relevant technologies on the spacer grid assembly. And, the UO{sub 2}-based SMART fuel is preliminarily designed for the 330MWt class SMART, which is planned to produce heat as well as electricity. Results from this project are listed as follows. 1. Three kinds of spacer grid candidates have been invented and applied for domestic and US patents. In addition, the demo SG(3x3 array) were fabricated, which the mechanical/structural test was carried out with. 2. The mechanical/structural technologies related to the spacer grid development are studied and relevant test requirements were established. 3. Preliminary design data of the UO{sub 2}-based SMART fuel have been produced. The structural characteristics of several components such as the top/bottom end piece and the holddown spring assembly were analysed by consulting the numerical method.

  11. Development of nuclear fuel for integrated reactor

    International Nuclear Information System (INIS)

    Song, Kee Nam; Kim, H. K.; Kang, H. S.; Yoon, K. H.; Chun, T. H.; In, W. K.; Oh, D. S.; Kim, D. W.; Woo, Y. M.

    1999-04-01

    The spacer grid assembly which provides both lateral and vertical support for the fuel rods and also provides a flow channel between the fuel rods to afford the heat transfer from the fuel pellet into the coolant in a reactor, is one of the major structural components of nuclear fuel for LWR. Therefore, the spacer grid assembly is a highly ranked component when the improvement of hardware is pursued for promoting fuel performance. Main objective of this project is to develop the inherent spacer grid assembly and to research relevant technologies on the spacer grid assembly. And, the UO 2 -based SMART fuel is preliminarily designed for the 330MWt class SMART, which is planned to produce heat as well as electricity. Results from this project are listed as follows. 1. Three kinds of spacer grid candidates have been invented and applied for domestic and US patents. In addition, the demo SG(3x3 array) were fabricated, which the mechanical/structural test was carried out with. 2. The mechanical/structural technologies related to the spacer grid development are studied and relevant test requirements were established. 3. Preliminary design data of the UO 2 -based SMART fuel have been produced. The structural characteristics of several components such as the top/bottom end piece and the holddown spring assembly were analysed by consulting the numerical method

  12. Fast breeder reactor fuel reprocessing in France

    International Nuclear Information System (INIS)

    Bourgeois, M.; Le Bouhellec, J.; Eymery, R.; Viala, M.

    1984-08-01

    Simultaneous with the effort on fast breeder reactors launched several years ago in France, equivalent investigations have been conducted on the fuel cycle, and in particular on reprocessing, which is an indispensable operation for this reactor. The Rapsodie experimental reactor was associated with the La Hague reprocessing plant AT1 (1 kg/day), which has reprocessed about one ton of fuel. The fuel from the Phenix demonstration reactor is reprocessed partly at the La Hague UP2 plant and partly at the Marcoule pilot facility, undergoing transformation to reprocess all the fuel (TOR project, 5 t/y). The fuel from the Creys Malville prototype power plant will be reprocessed in a specific plant, which is in the design stage. The preliminary project, named MAR 600 (50 t/y), will mobilize a growing share of the CEA's R and D resources, as the engineering needs of the UP3 ''light water'' plant begins to decline. Nearly 20 tonnes of heavy metals irradiated in fast breeder reactors have been processed in France, 17 of which came from Phenix. The plutonium recovered during this reprocessing allowed the power plant cycle to be closed. This power plant now contains approximately 140 fuel asemblies made up with recycled plutonium, that is, more than 75% of the fuel assemblies in the Phenix core

  13. Status of control assembly materials in Indian water reactors

    International Nuclear Information System (INIS)

    Date, V.G.; Kulkarni, P.G.

    2000-01-01

    India's present operating water cooled power reactors comprise boiling water reactors of Tarapur Atomic Power Station (TAPS) and pressurized heavy water reactors (PHWRs) at Kota (RAPS), Kalpakkam (MAPS), Narora (NAPS) and Kakrapara (KAPS). Boiling water reactors of TAPS use boron carbide control blades for control of power as well as for shut down (scram). PHWRs use boron steel and cobalt absorber rods for power control and Cd sandwiched shut off rods (primary shut down system) and liquid poison rods (secondary shut down system) for shut down. In TAPS, Gadolinium rods (burnable poison rods) are also incorporated in fuel assembly for flux flattening. Boron carbide control blades and Gadolinium rods for TAPS, cobalt absorber rods and shut down assemblies for PHWRs are fabricated indigenously. Considerable development work was carried out for evolving material specifications, component and assembly drawings, and fabrication processes. Details of various control and shut off assemblies being fabricated currently are highlighted in the paper. (author)

  14. Reactor fuel rod

    International Nuclear Information System (INIS)

    Inui, Mitsuhiro; Mori, Kazuma.

    1990-01-01

    In a high burnup degree reactor core, a problem of fuel can corrosion caused by coolants occurs due to long stay in a reactor. Then, the use of fuel cladding tubes with improved corrosion resistance is now undertaken and use of corrosion resistant alloys is attempted. However, since the conventional TIG welding melts the entire portion, the welded portion does not remain only in the corrosive resistant alloy but it forms new alloys of the corrosion resistant alloy and zircaloy as the matrix material or inter-metallic compounds, which degrades the corrosion resistance. In the present invention, a cladding tube comprising a dual layer structure using a corrosion resistant alloy only for a required thickness and an end plug made of the same material as the corrosion resistant alloy are welded at the junction portion by using resistance welding. Then, they are joined under welding by the heat generated to the junction surfaces between both of them, to provide corrosion resistant alloys substantially at the outside of the welded portion as well. Accordingly, the corrosion resistance is not degradated. (T.M.)

  15. Stationary Liquid Fuel Fast Reactor

    International Nuclear Information System (INIS)

    Yang, Won Sik; Grandy, Andrew; Boroski, Andrew; Krajtl, Lubomir; Johnson, Terry

    2015-01-01

    For effective burning of hazardous transuranic (TRU) elements of used nuclear fuel, a transformational advanced reactor concept named SLFFR (Stationary Liquid Fuel Fast Reactor) was proposed based on stationary molten metallic fuel. The fuel enters the reactor vessel in a solid form, and then it is heated to molten temperature in a small melting heater. The fuel is contained within a closed, thick container with penetrating coolant channels, and thus it is not mixed with coolant nor flow through the primary heat transfer circuit. The makeup fuel is semi- continuously added to the system, and thus a very small excess reactivity is required. Gaseous fission products are also removed continuously, and a fraction of the fuel is periodically drawn off from the fuel container to a processing facility where non-gaseous mixed fission products and other impurities are removed and then the cleaned fuel is recycled into the fuel container. A reference core design and a preliminary plant system design of a 1000 MWt TRU- burning SLFFR concept were developed using TRU-Ce-Co fuel, Ta-10W fuel container, and sodium coolant. Conservative design approaches were adopted to stay within the current material performance database. Detailed neutronics and thermal-fluidic analyses were performed to develop a reference core design. Region-dependent 33-group cross sections were generated based on the ENDF/B-VII.0 data using the MC2-3 code. Core and fuel cycle analyses were performed in theta-r-z geometries using the DIF3D and REBUS-3 codes. Reactivity coefficients and kinetics parameters were calculated using the VARI3D perturbation theory code. Thermo-fluidic analyses were performed using the ANSYS FLUENT computational fluid dynamics (CFD) code. Figure 0.1 shows a schematic radial layout of the reference 1000 MWt SLFFR core, and Table 0.1 summarizes the main design parameters of SLFFR-1000 loop plant. The fuel container is a 2.5 cm thick cylinder with an inner radius of 87.5 cm. The fuel

  16. Stationary Liquid Fuel Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Won Sik [Purdue Univ., West Lafayette, IN (United States); Grandy, Andrew [Argonne National Lab. (ANL), Argonne, IL (United States); Boroski, Andrew [Argonne National Lab. (ANL), Argonne, IL (United States); Krajtl, Lubomir [Argonne National Lab. (ANL), Argonne, IL (United States); Johnson, Terry [Argonne National Lab. (ANL), Argonne, IL (United States)

    2015-09-30

    For effective burning of hazardous transuranic (TRU) elements of used nuclear fuel, a transformational advanced reactor concept named SLFFR (Stationary Liquid Fuel Fast Reactor) was proposed based on stationary molten metallic fuel. The fuel enters the reactor vessel in a solid form, and then it is heated to molten temperature in a small melting heater. The fuel is contained within a closed, thick container with penetrating coolant channels, and thus it is not mixed with coolant nor flow through the primary heat transfer circuit. The makeup fuel is semi- continuously added to the system, and thus a very small excess reactivity is required. Gaseous fission products are also removed continuously, and a fraction of the fuel is periodically drawn off from the fuel container to a processing facility where non-gaseous mixed fission products and other impurities are removed and then the cleaned fuel is recycled into the fuel container. A reference core design and a preliminary plant system design of a 1000 MWt TRU- burning SLFFR concept were developed using TRU-Ce-Co fuel, Ta-10W fuel container, and sodium coolant. Conservative design approaches were adopted to stay within the current material performance database. Detailed neutronics and thermal-fluidic analyses were performed to develop a reference core design. Region-dependent 33-group cross sections were generated based on the ENDF/B-VII.0 data using the MC2-3 code. Core and fuel cycle analyses were performed in theta-r-z geometries using the DIF3D and REBUS-3 codes. Reactivity coefficients and kinetics parameters were calculated using the VARI3D perturbation theory code. Thermo-fluidic analyses were performed using the ANSYS FLUENT computational fluid dynamics (CFD) code. Figure 0.1 shows a schematic radial layout of the reference 1000 MWt SLFFR core, and Table 0.1 summarizes the main design parameters of SLFFR-1000 loop plant. The fuel container is a 2.5 cm thick cylinder with an inner radius of 87.5 cm. The fuel

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

    International Nuclear Information System (INIS)

    Morman, J.A.; Froehle, P.H.; Holland, J.W.; Bennett, J.D.

    1990-01-01

    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

  18. Integral Fast Reactor fuel pin processor

    International Nuclear Information System (INIS)

    Levinskas, D.

    1993-01-01

    This report discusses the pin processor which receives metal alloy pins cast from recycled Integral Fast Reactor (IFR) fuel and prepares them for assembly into new IFR fuel elements. Either full length as-cast or precut pins are fed to the machine from a magazine, cut if necessary, and measured for length, weight, diameter and deviation from straightness. Accepted pins are loaded into cladding jackets located in a magazine, while rejects and cutting scraps are separated into trays. The magazines, trays, and the individual modules that perform the different machine functions are assembled and removed using remote manipulators and master-slaves

  19. Nuclear fuel assembly

    International Nuclear Information System (INIS)

    Okubo, Kazutoshi.

    1991-01-01

    A region increased with the hydrophilic property at the surface is formed in the downstream portion of a cladding tube. Then, the surface of the region is made coarse. With such a constitution, coolants form three streams generally divided into a bubble stream, a slab stream and an annular stream along the cladding tube from the upstream to the downstream during operation of a BWR type reactor, and if BWR power is increased, the annular stream is enlarged. In this region, since oolant membranes for covering the surface of the region improved with its hydrophile property are easily formed by downcoming coolants, and the thickness of the coolant membrane is increased remarkably, the surface of the cladding tube can be cooled certainly and efficiently. Further, since there is no definit downwarding water stream in the annular portion, frictional pressure loss of the BWR is not increased by incresing the hydrophilic property. (T.M.)

  20. Fuel assembly storing rack

    International Nuclear Information System (INIS)

    Kajimura, Haruhiko; Nakamura, Masaaki.

    1997-01-01

    In a nuclear fuel storage rack comprising a plurality of square-shaped rack cells arranged vertically in a lattice-like configuration, the square rack cell comprises a stainless steel having a boron content of 0.75% or less and a boron equivalent with respect to calculated thermal neutron absorbing performance of 1.0 or more, and a distance between each of the gap of the square rack cells in adjacent with each other satisfies a predetermined condition. One example of the content is that B is from 0.05 to 0.75%, Gd is from 0.05 to 1.50%, C is 0.03% or less, Si is 1.0% or less, Mn is from 0.1 to 2.0%, P is 0.03% or less, S is 0.01% or less, Cr is from 18 to 26%, Ni is from 7 to 22% and Al is 0.1% or less. The distance between the rack cells can be reduced by using a material improved with thermal neutron absorbing performance by determining the boron equivalent to a predetermined value or more and with less B-content and good fabricability. In addition, the size of the rack cell itself can be reduced. This can greatly reduce the nuclear fuel storage area. (I.S.)

  1. Two-dimensional calculation by finite element method of velocity field and temperature field development in fast reactor fuel assembly. II

    International Nuclear Information System (INIS)

    Schmid, J.

    1985-11-01

    A package of updated computer codes for velocity and temperature field calculations for a fast reactor fuel subassembly (or its part) by the finite element method is described. Isoparametric triangular elements of the second degree are used. (author)

  2. Experimental investigation of local hydrodynamic characteristics of the peripheral area of a fast reactor fuel assembly. 2

    International Nuclear Information System (INIS)

    Hejna, J.; Vosahlo, L.; Cervenka, J.; Smid, J.

    Aerodynamic model HEM-2 was used for the first stage of experiments aimed at obtaining data allowing to determine the effect of assembly configuration changes on hydrodynamic conditions. The results are given of measuring tangential stresses and velocity fields. (Ha)

  3. Spacer grid for fuel assembly of nuclear reactor comprising opposite support points made with elastic thin plates

    International Nuclear Information System (INIS)

    Feutrel, C.

    1983-01-01

    Two series of thin walls form square cells, each containing a fuel pencil. Support points are made in the cells walls. Splines obtained by two parallel slots in the length of the cells. The reaction of fuel pencil produce a deformation of the elastic splines made in the plate, for compensation of the tolerance allowed on the diameter of the pencils [fr

  4. United States Domestic Research Reactor Infrastructure - TRIGA Reactor Fuel Support

    International Nuclear Information System (INIS)

    Morrell, Douglas

    2008-01-01

    The purpose of the United State Domestic Research Reactor Infrastructure Program is to provide fresh nuclear reactor fuel to United States universities at no, or low, cost to the university. The title of the fuel remains with the United States government and when universities are finished with the fuel, the fuel is returned to the United States government. The program is funded by the United States Department of Energy - Nuclear Energy division, managed by Department of Energy - Idaho Field Office, and contracted to the Idaho National Laboratory's Management and Operations Contractor - Battelle Energy Alliance. Program has been at Idaho since 1977 and INL subcontracts with 26 United States domestic reactor facilities (13 TRIGA facilities, 9 plate fuel facilities, 2 AGN facilities, 1 Pulstar fuel facility, 1 Critical facility). University has not shipped fuel since 1968 and as such, we have no present procedures for shipping spent fuel. In addition: floor loading rate is unknown, many interferences must be removed to allow direct access to the reactor tank, floor space in the reactor cell is very limited, pavement ends inside our fence; some of the surface is not finished. The whole approach is narrow, curving and downhill. A truck large enough to transport the cask cannot pull into the lot and then back out (nearly impossible / refused by drivers); a large capacity (100 ton), long boom crane would have to be used due to loading dock obstructions. Access to the entrance door is on a sidewalk. The campus uses it as a road for construction equipment, deliveries and security response. Large trees are on both sides of sidewalk. Spent fuel shipments have never been done, no procedures approved or in place, no approved casks, no accident or safety analysis for spent fuel loading. Any cask assembly used in this facility will have to be removed from one crane, moved on the floor and then attached to another crane to get from the staging area to the reactor room. Reactor

  5. Plutonium assemblies in reload 1 of the Dodewaard Reactor

    International Nuclear Information System (INIS)

    Bairiot, H.; Deramaix, P.; Vandenberg, C.; Leenders, L.; Mostert, P.

    1977-01-01

    Since 1963, Belgonucleaire has been developing the design of plutonium assemblies of the island type (i.e., plutonium rods inserted in the control zone of the assembly and enriched uranium rods at the periphery) for light water reactors. The application to boiling water reactors (BWRs) led to the introduction, in April 1971, of two prototype plutonium island assemblies in the Dodewaard BWR (The Netherlands): Those assemblies incorporating plutonium in 42 percent of the rods are interchangeable with standard uranium assemblies of the same reload. Their design, which had to meet these criteria, was performed using the routine order in use at Belgonucleaire; experimental checks included a mock-up configuration simulated in the VENUS critical facility at Mol and open-vessel cold critical experiments performed in the Dodewaard core. The pelleted plutonium rods were fabricated and controlled by Belgonucleaire following the manufacturing procedures developed at the production plant. In one of the assemblies, three vibrated plutonium fuel rods with a lower fuel density were introduced in the three most highly rated positions to reduce the power rating. Those plutonium assemblies experienced peak pellet ratings up to 535 W/cm and were discharged in April 1974 after having reached a mean burnup of approximately 21,000 MWd/MT. In-core instrumentation during operation, visual examinations, and reactivity substitution experiments during reactor shutdown did not indicate any special feature for those assemblies compared to the standard uranium assemblies, thereby demonstrating their interchangeability

  6. Inlet for fuel assembly having finger control rods

    International Nuclear Information System (INIS)

    Berglund, A.; Suvanto, A.; Tornblom, L.

    1975-01-01

    A nuclear reactor with vertically arranged fuel assemblies positioned on supporting members and with control rods displaceably arranged in guide tubes between the fuel rods inside the fuel assemblies is described. The supporting plate is provided with a transverse end piece with throttling means for the liquid flow which passes from below up through the supporting member and past the fuel rods in the fuel assembly. The inlets for the guide tubes for the control rods are located below the end piece and the throttling means. In this way a higher pressure prevails at the inlet to the guide tubes than above the end piece, so that a stronger flow of coolant is produced through guide tubes than through the fuel assembly. (U.S.)

  7. Advanced Research Reactor Fuel Development

    Energy Technology Data Exchange (ETDEWEB)

    Kim, C. K.; Park, H. D.; Kim, K. H. (and others)

    2006-04-15

    RERTR program for non-proliferation has propelled to develop high-density U-Mo dispersion fuels, reprocessable and available as nuclear fuel for high performance research reactors in the world. As the centrifugal atomization technology, invented in KAERI, is optimum to fabricate high-density U-Mo fuel powders, it has a great possibility to be applied in commercialization if the atomized fuel shows an acceptable in-reactor performance in irradiation test for qualification. In addition, if rod-type U-Mo dispersion fuel is developed for qualification, it is a great possibility to export the HANARO technology and the U-Mo dispersion fuel to the research reactors supplied in foreign countries in future. In this project, reprocessable rod-type U-Mo test fuel was fabricated, and irradiated in HANARO. New U-Mo fuel to suppress the interaction between U-Mo and Al matrix was designed and evaluated for in-reactor irradiation test. The fabrication process of new U-Mo fuel developed, and the irradiation test fuel was fabricated. In-reactor irradiation data for practical use of U-Mo fuel was collected and evaluated. Application plan of atomized U-Mo powder to the commercialization of U-Mo fuel was investigated.

  8. 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.

  9. REACTOR FUEL ELEMENTS TESTING CONTAINER

    Science.gov (United States)

    Whitham, G.K.; Smith, R.R.

    1963-01-15

    This patent shows a method for detecting leaks in jacketed fuel elements. The element is placed in a sealed tank within a nuclear reactor, and, while the reactor operates, the element is sparged with gas. The gas is then led outside the reactor and monitored for radioactive Xe or Kr. (AEC)

  10. System for assembling nuclear fuel elements

    International Nuclear Information System (INIS)

    1980-01-01

    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.)

  11. Shock absorbing structure for nuclear fuel assemblies

    International Nuclear Information System (INIS)

    Jabsen, F.S.

    1981-01-01

    A hydraulic apparatus is described that absorbs shocks that may be applied to fuel assemblies. Spring pads mounted on the upper end fittings of the fuel assemblies have plungers that move within hollow guide posts attached to the upper grids of the fuel assemblies. (L.L.)

  12. Fuel assembly and fuel cladding tube

    International Nuclear Information System (INIS)

    Tsutsumi, Shinro; Ito, Ken-ichi; Inagaki, Masatoshi; Nakajima, Junjiro.

    1996-01-01

    A fuel cladding tube is a zirconium liner tube formed by lining a pure zirconium layer on the inner side of a zirconium alloy tube. The fuel cladding tube is formed by extrusion molding of a composite billet formed by inserting a pure zirconium billet into a zirconium alloy billet. Accordingly, the pure zirconium layer and the zirconium alloy tube are strongly joined by metal bond. The fuel cladding tube has an external oxide film on the outer surface of the zirconium alloy tube and an internal oxide film on the inner side of the pure zirconium layer. The external oxide film has a thickness preferably of about 1μm. The internal oxide film has a thickness of not more than 10μm, preferably, from 1 to 5μm. With such a constitution, flaws to be formed on both inner and outer surfaces of the cladding tube upon assembling a fuel assembly can be reduced thereby enabling to reduce the amount of hydrogen absorbed to the cladding tube. (I.N.)

  13. Fuel handling grapple for nuclear reactor plants

    International Nuclear Information System (INIS)

    Rousar, D.L.

    1992-01-01

    This patent describes a fuel handling system for nuclear reactor plants. It comprises: a reactor vessel having an openable top and removable cover and containing therein, submerged in water substantially filling the reactor vessel, a fuel core including a multiplicity of fuel bundles formed of groups of sealed tube elements enclosing fissionable fuel assembled into units, the fuel handling system consisting essentially of the combination of: a fuel bundle handling platform movable over the open top of the reactor vessel; a fuel bundle handling mast extendable downward from the platform with a lower end projecting into the open top reactor vessel to the fuel core submerged in water; a grapple head mounted on the lower end of the mast provided with grapple means comprising complementary hooks which pivot inward toward each other to securely grasp a bail handle of a nuclear reactor fuel bundle and pivot backward away from each other to release a bail handle; the grapple means having a hollow cylindrical support shaft fixed within the grapple head with hollow cylindrical sleeves rotatably mounted and fixed in longitudinal axial position on the support shaft and each sleeve having complementary hooks secured thereto whereby each hook pivots with the rotation of the sleeve secured thereto; and the hollow cylindrical support shaft being provided with complementary orifices on opposite sides of its hollow cylindrical and intermediate to the sleeves mounted thereon whereby the orifices on both sides of the hollow cylindrical support shaft are vertically aligned providing a direct in-line optical viewing path downward there-through and a remote operator positioned above the grapple means can observe from overhead the area immediately below the grapple hooks

  14. Canadian power reactor fuel

    International Nuclear Information System (INIS)

    Page, R.D.

    1976-03-01

    The following subjects are covered: the basic CANDU fuel design, the history of the bundle design, the significant differences between CANDU and LWR fuel, bundle manufacture, fissile and structural materials and coolants used in the CANDU fuel program, fuel and material behaviour, and performance under irradiation, fuel physics and management, booster rods and reactivity mechanisms, fuel procurement, organization and industry, and fuel costs. (author)

  15. Most advanced HTP fuel assembly design for EPR

    International Nuclear Information System (INIS)

    Francillon, Eric; Kiehlmann, Horst-Dieter

    2006-01-01

    End 2003, the Finnish electricity utility Teollisuuden Voima Oy (TVO) signed the contract for building an EPR in Olkiluoto (Finland). Mid 2004, the French electricity utility EDF selected an EPR to be built in France. In 2005, Framatome ANP, an AREVA and Siemens company, announced that they will be pursuing a design certification in the U.S. The EPR development is based on the latest PWR product lines of former Framatome (N4) and Siemens Nuklear (Konvoi). As an introductory part, different aspects of the EPR core characteristics connected to fuel assembly design are presented. It includes means of ensuring reactivity control like hybrid AIC/B4C control rod absorbers and gadolinium as burnable absorber integrated in fuel rods, and specific options for in-core instrumentation, such as Aeroball type instrumentation. Then the design requirements for the EPR fuel assembly are presented in term of very high burnup capacity, rod cladding and fuel assembly reliability. Framatome ANP fuel assembly product characteristics meeting these requirements are then described. EPR fuel assembly design characteristics benefit from the experience feedback of the latest fuel assembly products designed within Framatome ANP, leading to resistance to assembly deformation, high fuel rod restraint and prevention of handling hazards. EPR fuel assembly design features the best components composing the cornerstones of the upgraded family of fuel assemblies that FRAMATOME ANP proposes today. This family is based on a set of common characteristics and associated features, which include the HMP grid as bottom end spacer, the MONOBLOC guide tube and the Robust FUELGUARD as lower tie plate, the use of the M5 Alloy, as cladding and structure material. This fully re-crystallized, ternary Zr-Nb-O alloy produces radically improved in-reactor corrosion, very low hydrogen uptake and growth and an excellent creep behavior, which are described there. EPR fuel assembly description also includes fuel rod

  16. Results of TVSA fuel assemblies development and 10-years operation in WWER-1000 reactors cores. Development trends

    International Nuclear Information System (INIS)

    Molchanov, V.; Sharikov, A.; Samoilov, O.; Kaidalov, V.; Falkov, A.; Romanov, A.; Shishkin, A.

    2009-01-01

    The basic TVSA design is now operating successfully in 17 VVER-1000 power units of Russia, Ukraine and Bulgaria. TVSA design is characterized by reliability and validity of designs and resource characteristics. TVSA possesses light hydraulics and peak ratio for implementation of effective fuel cycles. Unique parameters on time of maintenance and burnup are achieved. The basic design - TVSA-5M - has been introduced into commercial operation. It is necessary to expand heading TVSA with debris filter and temperature control tubes. Two directions of evolution of a design - TVSA-ALPHA and TVSAPLUS are realized. Evolution is directed on increasing uranium content and realization of 5x1 and 3x1.5 fuel cycles with operation at raised power. The TVSA design offers a great optimization potential based on a series of proven (reference) technical solutions. TVSA is competitive and cost effective and ensures the market advantages of Russian nuclear fuel

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

    International Nuclear Information System (INIS)

    Dragusin, M.

    2009-01-01

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

  18. Preliminary concepts: safeguards for spent light-water reactor fuels

    International Nuclear Information System (INIS)

    Cobb, D.D.; Dayem, H.A.; Dietz, R.J.

    1979-06-01

    The technology available for safeguarding spent nuclear fuels from light-water power reactors is reviewed, and preliminary concepts for a spent-fuel safeguards system are presented. Essential elements of a spent-fuel safeguards system are infrequent on-site inspections, containment and surveillance systems to assure the integrity of stored fuel between inspections, and nondestructive measurements of the fuel assemblies. Key safeguards research and development activities necessary to implement such a system are identified. These activities include the development of tamper-indicating fuel-assembly identification systems and the design and development of nondestructive spent-fuel measurement systems

  19. Lateral restraint assembly in a nuclear reactor

    International Nuclear Information System (INIS)

    Brown, S.J.; Gorholt, W.

    1977-01-01

    A lateral restraint assembly is described for a reactor of, for example, the high temperature gas-cooled type which commonly includes a reactor core of relatively complex construction supported within a shell or vessel providing a shielded cavity for containing the reactor core. (U.K.)

  20. Determination of mixture coefficients in tests simulating channels of a fuel assembly in a water cooled reactor

    International Nuclear Information System (INIS)

    Ferreira, W.R.

    1983-09-01

    Here, the mixture coefficients are determined in tests which simulate two symmetric and two assymetric coolant channels of a water cooled reactor. It was studies the effects of coolant flow and of the distance among the coolant channels on these coefficients. The technique used to determine the mixture coefficients was to introduce a tracer (methylene blue) into the flow. The determination of the tracer concentration at the end of the channels was made by means of spectrophotometric analysis. (author)

  1. Results of trial operation of the WWER advanced fuel assemblies

    International Nuclear Information System (INIS)

    Vasilchenko, I.; Dragunov, Y.; Mikhalchuk, A.

    2001-01-01

    The paper describes results from experimental operation of advanced WWER-1000 fuel assemblies (AFA) at five units in Balakovo NPP. Advanced fuel is developed according to the concept of standard WWER-1000 fuel assembly (jacket-free). The new features includes: 1) zirconium guiding channels (alloy E-635 and E-110) and spacer grids (alloy E-110); 2) integrated burnable absorber gadolinium; 3) extended service life of fuel assemblies (FA) and absorber rods (possibility of repair of FA); 4) improved adoption to reactor conditions. Some results of AFA pilot operation of a three year operation are presented and analyses of effectiveness of improvements are made concerning application of zirconium channels and grids; application of integrated burnable absorbers; extension of FA and absorbing rods service life and FA repairability. These new features of WWER-1000 fuel design allow: 1) to reduce the average fuel enrichment to the 3.77% instead of 4.31% in U-235; 2) to reduce the FA axial load in reactor hot state by 40%,; 3) increasing of fuel operation in reactor to the 30000 effective days with possibility to have a 5-year residence time in the reactor. The design of new generation FA for WWER-440 reactors involves few key changes. Fuel inventory in new fuel design is increased due to elongation of fuel stack and reducing the diameter of the central hole. Vibration stability is enhanced as a result of: no-play junction of the fuel rod with the lower grid; change of SG arrangements; strengthening of the lower grid unit; secure of the central tube in the gap. Water-uranium ration is increased. Introduction of all these kinds of modernization in a 5-year fuel cycle reduces fuel component in the energy cost to the 7%

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

    International Nuclear Information System (INIS)

    Wu Zhongwang; Zhang Yajun

    2000-01-01

    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

  3. Reactor core fuel management

    International Nuclear Information System (INIS)

    Silvennoinen, P.

    1976-01-01

    The subject is covered in chapters, entitled: concepts of reactor physics; neutron diffusion; core heat transfer; reactivity; reactor operation; variables of core management; computer code modules; alternative reactor concepts; methods of optimization; general system aspects. (U.K.)

  4. Grid spacers for use in a nuclear fuel assembly

    International Nuclear Information System (INIS)

    Kuwako, Akira.

    1987-01-01

    Purpose: To obtain spacers capable of reducing the pressure loss by enlarging coolant flow channels when the fuel temperature is high, while capable of reliably maintaining the fuel pins with no vibrations when the fuel temperature is low. Constitution: This invention concerns grid spacers for constituting fuel assemblies for use in water cooled reactors. Memory shape alloys are disposed at least a portion of a spacer element that takes such a shape as urging the pin when the fuel temperature is low, while enlarging the coolant flow channel to reduce the pressure loss when the fuel temperature is high. (Ikeda, J.)

  5. Heat evaluation examination of fuel assembly

    International Nuclear Information System (INIS)

    Suto, Shinya; Nakabayashi, Hiroki; Yao, Kaoru

    2007-03-01

    The cooling examination was executed by using the simulated fuel assembly to obtain the basic data of the most effective cooling system in the lazer disassembling process of the spent fuel assembly of prototype fast breeder reactor 'Monju'. As a result, the following have been understood. (1) Before the laser disassembling (there is not any duct tube cutting), it is possible to cool enough by the amount of the wind of 20m 3 /h or more flowing from the handling head side. (2) After the laser disassembling begins (duct tube is cut), 1kW or more of the heat generation cannot be cooled by ventilation from the handling head side. (3) Cooling by the flow across fuel pin is required during lazer disassembling. The basic data of the cooling system was obtained from these examination results. However, for cooling across fuel pin during the laser disassembling, it is necessary to examine shape of the side cooling nozzle, spraying angle, and flow velocity at the nozzle exit, etc. enough. (author)

  6. Storage method for spent fuel assembly

    International Nuclear Information System (INIS)

    Tajiri, Hiroshi.

    1992-01-01

    In the present invention, spent fuel assemblies are arranged at a dense pitch in a storage rack by suppressing the reactivity of the assemblies, to increase storage capacity for the spent fuel assemblies. That is, neutron absorbers are filled in the cladding tube of an absorbing rod, and the diameter thereof is substantially equal with that of a fuel rod. A great amount of the absorbing rods are arranged at the outer circumference of the fuel assembly. Then, they are fixed integrally to the fuel assembly and stored in a storage rack. In this case, the storage rack may be constituted only with angle materials which are inexpensive and installed simply. With such a constitution, in the fuel assembly having absorbing rods wound therearound, neutrons are absorbed by absorbing rods and the reactivity is lowered. Accordingly, the assembly arrangement pitch in the storage rack can be made dense. As a result, the storage capacity for the assemblies is increased. (I.S.)

  7. Micro-Reactor Physics of MOX-Fueled Core

    International Nuclear Information System (INIS)

    Takeda, T.

    2001-01-01

    Recently, fuel assemblies of light water reactors have become complicated because of the extension of fuel burnup and the use of high-enriched Gd and mixed-oxide (MOX) fuel, etc. In conventional assembly calculations, the detailed flux distribution, spectrum distribution, and space dependence of self-shielding within a fuel pellet are not directly taken into account. The experimental and theoretical study of investigating these microscopic properties is named micro-reactor physics. The purpose of this work is to show the importance of micro-reactor physics in the analysis of MOX fuel assemblies. Several authors have done related studies; however, their studies are limited to fuel pin cells, and they are never mentioned with regard to burnup effect, which is important for actual core design

  8. FFTF reactor assembly system technology

    International Nuclear Information System (INIS)

    Mangelsdorf, T.A.

    1975-01-01

    An overview is presented of the FFTF reactor and plant together with descriptions of core components, core internals, core system, primary and secondary control rod system, reactor instrumentation, reactor vessel and closure head, and supporting test programs

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

    International Nuclear Information System (INIS)

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

    2008-06-01

    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

  10. 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.

  11. Power generation costs for alternate reactor fuel cycles

    International Nuclear Information System (INIS)

    Smolen, G.R.; Delene, J.G.

    1980-09-01

    The total electric generating costs at the power plant busbar are estimated for various nuclear reactor fuel cycles which may be considered for power generation in the future. The reactor systems include pressurized water reactors (PWR), heavy-water reactors (HWR), high-temperature gas cooled reactors (HTGR), liquid-metal fast breeder reactors (LMFBR), light-water pre-breeder and breeder reactors (LWPR, LWBR), and a fast mixed spectrum reactor (FMSR). Fuel cycles include once-through, uranium-only recycle, and full recycle of the uranium and plutonium in the spent fuel assemblies. The U 3 O 8 price for economic transition from once-through LWR fuel cycles to both PWR recycle and LMFBR systems is estimated. Electric power generation costs were determined both for a reference set of unit cost parameters and for a range of uncertainty in these parameters. In addition, cost sensitivity parameters are provided so that independent estimations can be made for alternate cost assumptions

  12. Bimetallic spacer means for a nuclear fuel assembly

    International Nuclear Information System (INIS)

    Anthony, A.J.

    1981-01-01

    A bimetallic spacer means designed to be cooperatively associated with a nuclear fuel assembly and operative to resist the occurrence of in-reactor bowing of the nuclear fuel assembly. The subject bimetallic spacer means in accord with one embodiment of the invention includes a member formed, at least principally, of Zircaloy to which are attached a plurality of stainless steel strips. The latter stainless steel strips are located on the external surface of the Zircaloy member and with the major axis of each of the plurality of stainless steel strips extending substantially perpendicular to the major axis of the Zircaloy member. In accord with another embodiment of the invention, the subject bimetallic spacer means includes a member formed at least principally of Zircaloy to which a plurality of stainless steel strips are attached so as to be positioned thereon externally thereof and with the major axis of each of the plurality of stainless steel strips extending substantially parallel to the major axis of the Zircaloy member. In accord with a further embodiment of the invention, the stainless steel strips are attached to preselected members, each embodying at least a cladding of Zircaloy, which are located in the rows of fuel rods that define the perimeter of the fuel matrix of the nuclear fuel assembly. In each of the embodiments, the stainless steel strips during power production expand outwardly to a greater extent than do the members to which the stainless steel strips are attached, thereby forming stiff springs which abut against like bimetallic spacer means with which the other nuclear fuel assemblies are provided in a given nuclear reactor core to thus prevent the occurrence of in-reactor bowing of the nuclear fuel assemblies. Namely, the stainless steel strips expand laterally relative to the fuel assembly and thereby occupy the space adjacent to the external surface of the fuel assembly

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

    International Nuclear Information System (INIS)

    Woolstenhulme, N.E.; Moore, G.A.; Perez, D.M.; Wachs, D.M.

    2010-01-01

    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)

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

    International Nuclear Information System (INIS)

    Archipov, A.; Kolochko, V.N.

    2001-01-01

    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)

  15. CANDU reactor experience: fuel performance

    International Nuclear Information System (INIS)

    Truant, P.T.; Hastings, I.J.

    1985-07-01

    Ontario Hydro has more than 126 reactor-years experience in operating CANDU reactors. Fuel performance has been excellent with 47 000 channel fuelling operations successfully completed and 99.9 percent of the more than 380 000 bundles irradiated operating as designed. Fuel performance limits and fuel defects have had a negligible effect on station safety, reliability, the environment and cost. The actual incapability charged to fuel is less than 0.1 percent over the stations' lifetimes, and more recently has been zero

  16. Ceramics as nuclear reactor fuels

    International Nuclear Information System (INIS)

    Reeve, K.D.

    1975-01-01

    Ceramics are widely accepted as nuclear reactor fuel materials, for both metal clad ceramic and all-ceramic fuel designs. Metal clad UO 2 is used commercially in large tonnages in five different power reactor designs. UO 2 pellets are made by familiar ceramic techniques but in a reactor they undergo complex thermal and chemical changes which must be thoroughly understood. Metal clad uranium-plutonium dioxide is used in present day fast breeder reactors, but may eventually be replaced by uranium-plutonium carbide or nitride. All-ceramic fuels, which are necessary for reactors operating above about 750 0 C, must incorporate one or more fission product retentive ceramic coatings. BeO-coated BeO matrix dispersion fuels and silicate glaze coated UO 2 -SiO 2 have been studied for specialised applications, but the only commercial high temperature fuel is based on graphite in which small fuel particles, each coated with vapour deposited carbon and silicon carbide, are dispersed. Ceramists have much to contribute to many aspects of fuel science and technology. (author)

  17. Method for inspecting nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    Jabsen, F.S.

    1979-01-01

    A technique for disassembling a nuclear reactor fuel element without destroying the individual fuel pins and other structural components from which the element is assembled is described. A traveling bridge and trolley span a water-filled spent fuel storage pool and support a strongback. The strongback is under water and provides a working surface on which the spent fuel element is placed for inspection and for the manipulation that is associated with disassembly and assembly. To remove, in a non-destructive manner, the grids that hold the fuel pins in the proper relative positions within the element, bars are inserted through apertures in the grids with the aid of special tools. These bars are rotated to flex the adjacent grid walls and, in this way relax the physical engagement between protruding portions of the grid walls and the associated fuel pins. With the grid structure so flexed to relax the physical grip on the individual fuel pins, these pins can be withdrawn for inspection or replacement as necessary without imposing a need to destroy fuel element components

  18. Performance tests for integral reactor nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Sohn, Dong-Seong; Yim, Jeong-Sik; Lee, Chong-Tak; Kim, Han-Soo; Koo, Yang-Hyun; Lee, Byung-Ho; Cheon, Jin-Sik; Oh, Je-Yong

    2006-02-15

    An integral type reactor SMART plans to utilize metallic Zr-U fuel which is Zr-based alloy with 34{approx}38 wt% U. In order to verify the technologies for the design and manufacturing of the fuel and get a license, performance tests were carried out. Experimental Fuel Assembly (EFA) manufactured in KAERI is being successfully irradiated in the MIR reactor of RIAR from September 4 2004, and it has achieved burnup of 0.21 g/cc as of January 25 2006. Thermal properties of irradiated Zr-U fuel were measured. Up to the phase transformation temperature, thermal diffusivity increased linearly in proportion to temperature. However its dependence on the burnup was not significant. RIA tests with 4 unirradiated Zr-U fuel rods were performed in Kurchatov Institute to establish a safety criterion. In the case of the un-irradiated Zr-U fuel, the energy deposition during the control rod ejection accident should be less than 172 cal/g to prevent the failure accompanying fuel fragmentation and dispersal. Finally the irradiation tests of fuel rods have been performed at HANARO. The HITE-2 test was successfully completed up to a burnup of 0.31 g/cc. The HITE-3 test began in February 2004 and will be continued up to a target burnup of 0.6 g/cc.

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

    International Nuclear Information System (INIS)

    Fujieda, Tadashi; Inagaki, Masatoshi; Takase, Iwao; Nishino, Yoshitaka; Yamashita, Jun-ichi; Yamanaka, Akihiro; Ito, Ken-ichi; Nakajima, Junjiro; Seto, Takehiro.

    1998-01-01

    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.)

  20. Main results of post-irradiation examinations of new-generation fuel assemblies VVER-1000

    International Nuclear Information System (INIS)

    Zvir, E.; Markov, D.; Polenok, V.; Zhitelev, V.; Kobylyansky, G.

    2009-01-01

    To increase the competitiveness of Russian nuclear fuel at the foreign market and to improve its technical and economic performance in order to provide a necessary level of safety, it is necessary to solve certain important tasks: Increase of fuel burn-up; Extension of operational lifetime of fuel assemblies and operational reliability of nuclear fuel; Introduction of cost-beneficial and flexible fuel cycles. Alternative fuel assemblies TVSA VVER-1000 and TVS-2 are used as a basis to optimize the nuclear fuel and develop advanced fuel cycles for nuclear power plants with VVER-1000 reactor types. Four fuel assemblies TVSA operated during 1 and up to 6 reactor cycles, reference fuel assembly TVS-2 operated during three reactor cycles and achieved an average fuel burnup of 48MW·day/kgU as well as failed fuel assembly TVS-2 operated during one cycle were examined at RIAR in recent years. The main objectives of these examinations were to obtain experimental data in support of operational integrity of products or to find out reasons of their failure. The performed post-irradiation examinations confirmed the operational integrity of alternative fuel assemblies TVSA including their geometrical stability up to the average fuel burnup of 55 MW·day/kgU over the fuel assembly (FA) (up to the maximal fuel burnup of ∼73 MW·day/kgU in fuel rods) and of TVS-2 up to the average fuel burnup of 48 MW·day/kgU over the fuel assembly. The changes introduced in the design of VVER-1000 fuel assembly during the development of alternative fuel assembly TVSA and TVS-2 did not make any negative effect on fuel rods. It was proved that causes of fuel rod failure were not related to design features of fuel assemblies. The design features and operating conditions of fuel assemblies under examinations are briefly described. Post-irradiation examinations proved the geometrical stability of fuel assemblies TVSA and TVS-2 under operation up to the fuel burnup of ∼50 MW day/kgU, as for the

  1. Caramel fuel for research reactors

    International Nuclear Information System (INIS)

    Bussy, P.

    1979-11-01

    This fuel for research reactors is made of UO 2 pellets in a zircaloy cladding to replace 93% enriched uranium. It is a cold fuel, non contaminating and non proliferating, enrichment is only 7 to 8%. Irradiation tests were performed until burn-up of 50000 MWD/t [fr

  2. Conditioning of nuclear reactor fuel

    International Nuclear Information System (INIS)

    1975-01-01

    A method of conditioning the fuel of a nuclear reactor core to minimize failure of the fuel cladding comprising increasing the fuel rod power to a desired maximum power level at a rate below a critical rate which would cause cladding damage is given. Such conditioning allows subsequent freedom of power changes below and up to said maximum power level with minimized danger of cladding damage. (Auth.)

  3. Reactor transients tests for SNR fuel elements in HFR reactor

    International Nuclear Information System (INIS)

    Plitz, H.

    1989-01-01

    In HFR reactor, fuel pins of LMFBR reactors are putted in irradiation specimen capsules cooled with sodium for reactor transients tests. These irradiation capsules are instrumented and the experiences realized until this day give results on: - Fuel pins subjected at a continual variation of power - melting fuel - axial differential elongation of fuel pins

  4. Fuels for Canadian research reactors

    International Nuclear Information System (INIS)

    Feraday, M.A.

    1993-01-01

    This paper includes some statements and remarks concerning the uranium silicide fuels for which there is significant fabrication in AECL, irradiation and defect performance experience; description of two Canadian high flux research reactors which use high enrichment uranium (HEU) and the fuels currently used in these reactors; limited fabrication work done on Al-U alloys to uranium contents as high as 40 wt%. The latter concerns work aimed at AECL fast neutron program. This experience in general terms is applied to the NRX and NRU designs of fuel

  5. Magnetic scanning of LWR fuel assemblies

    International Nuclear Information System (INIS)

    Fiarman, S.; Moodenbaugh, A.

    1980-01-01

    Nondestructive assay (NDA) techniques are available both for fresh and spent fuel, but generally are too time consuming and do not uniquely identify an assembly. A new method is reported to obtain a signature from a magnetic scan of each assembly. This scan is an NDA technique that detects magnetic inclusions. It is potentially fast (5 min/assembly), and may provide a unique signature from the magnetic properties of each fuel assembly

  6. Mixed Reload Design Using MOX and UOX Fuel Assemblies

    International Nuclear Information System (INIS)

    Ramon, Ramirez Sanchez J.; Perry, R.T.

    2002-01-01

    As part of the studies involved in plutonium utilization assessment for a Boiling Water Reactor, a conceptual design of MOX fuel was developed, this design is mechanically the same design of 10 X 10 BWR fuel assemblies but different fissile material. Several plutonium and gadolinium concentrations were tested to match the 18 months cycle length which is the current cycle length of LVNPP, a reference UO 2 assembly was modeled to have a full cycle length to compare results, an effective value of 0.97 for the multiplication factor was set as target for 470 Effective Full Power days for both cycles, here the gadolinium concentration was a key to find an average fissile plutonium content of 6.55% in the assembly. A reload of 124 fuel assemblies was assumed to simulate the complete core, several load fractions of MOX fuel mixed with UO 2 fresh fuel were tested to verify the shutdown margin, the UO 2 fuel meets the shutdown margin when 124 fuel assemblies are loaded into the core, but it does not happen when those 124 assemblies are replaced with MOX fuel assemblies, so the fraction of MOX was reduced step by step up to find a mixed load that meets both length cycle and shutdown margin. Finally the conclusion is that control rods losses some of their worth in presence of plutonium due to a more hardened neutron spectrum in MOX fuel and this fact limits the load of MOX fuel assemblies in the core, this results are shown in this paper. (authors)

  7. Top Nozzle Holddown Spring Optimization of KSNP Fuel Assembly

    International Nuclear Information System (INIS)

    Lee, Seong Ki; Park, Nam Kyu; Kim, Hyeong Koo; Lee, Joon Ro; Kim, Jae Won

    2002-01-01

    Nuclear fuel assembly for Korea Standard Nuclear Power (KSNP) Plant has 4 helical compression springs at the upper end of it. The springs, in conjunction with the fuel assembly weight, apply a holddown force against excess of buoyancy forces and the upward hydraulic forces due to the reactor coolant flow. Thus the holddown spring is to be designed such that the positive net downward force will be maintained for all normal and anticipated transient flow and temperature conditions in the nuclear reactor. With satisfying these in-reactor requirements of the fuel assembly holddown spring. Under the assumption that spring density is constant, the volume nozzle holddown spring. Under the assumption that spring density is constant, the volume minimization is executed by using the design variables, viz., wire diameter, mean coil diameter, minimization is executed by using the design variables, viz., wire diameter, mean coil diameter are within the compatible range of the fuel assembly structural components. Based on these conditions, the optimum design of the holddown spring is obtained considering the reactor operating condition and by using ANSYS code. The optimized spring has the properties that are a decreased volume and increased stiffness, compared with the existing one even if the absolute values are very similar each other. The holddown spring design features and the algorithm developed in this study could be directly applicable to the current commercial production. Therefore, it could be used to enhance the design efficiency and the functional performance of the spring, and to reduce a material cost a little

  8. Fast Neutron Emission Tomography of Used Nuclear Fuel Assemblies

    Science.gov (United States)

    Hausladen, Paul; Iyengar, Anagha; Fabris, Lorenzo; Yang, Jinan; Hu, Jianwei; Blackston, Matthew

    2017-09-01

    Oak Ridge National Laboratory is developing a new capability to perform passive fast neutron emission tomography of spent nuclear fuel assemblies for the purpose of verifying their integrity for international safeguards applications. Most of the world's plutonium is contained in spent nuclear fuel, so it is desirable to detect the diversion of irradiated fuel rods from an assembly prior to its transfer to ``difficult to access'' storage, such as a dry cask or permanent repository, where re-verification is practically impossible. Nuclear fuel assemblies typically consist of an array of fuel rods that, depending on exposure in the reactor and consequent ingrowth of 244Cm, are spontaneous sources of as many as 109 neutrons s-1. Neutron emission tomography uses collimation to isolate neutron activity along ``lines of response'' through the assembly and, by combining many collimated views through the object, mathematically extracts the neutron emission from each fuel rod. This technique, by combining the use of fast neutrons -which can penetrate the entire fuel assembly -and computed tomography, is capable of detecting vacancies or substitutions of individual fuel rods. This paper will report on the physics design and component testing of the imaging system. This material is based upon work supported by the U.S. Department of Energy, Office of Defense Nuclear Nonproliferation Research and Development within the National Nuclear Security Administration, under Contract Number DE-AC05-00OR22725.

  9. Framatome experience in fuel assembly repair and reconstitution

    International Nuclear Information System (INIS)

    Leroy, G.

    1998-01-01

    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 UO 2 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)

  10. Neutronic characteristics of linear-assembly breed-and-burn reactors

    International Nuclear Information System (INIS)

    Petroski, Robert; Forget, Benoit; Forsberg, Charles

    2012-01-01

    Highlights: ► Simple models used to characterize general behavior of linear-assembly B and B reactors. ► Diffusion theory model developed to explain axial distributions, height vs. reactivity. ► Neutron excess concept reformulated to include linear-assembly B and B reactors. ► Designed model of B and B reactor started using melt-refined B and B reactor used fuel. ► Computed doubling time of fuel cycle requiring no chemical separations. - Abstract: Linear-assembly breed-and-burn (B and B) reactors are B and B reactors that use axially connected assemblies similar to conventional LWR or fast reactor fuel assemblies. Methods for analyzing linear-assembly B and B reactors and their fuel cycles are developed and applied. General neutronic characteristics of linear-assembly B and B reactors are analyzed, including the effects that burnup, shuffling sequence, and radial and axial size have on equilibrium-cycle k-effective. The mechanisms that give rise to a highly peaked axial burnup distribution are explained, and a method for predicting peak burnup vs. k-effective based on infinite-medium depletion calculations is developed. Next, the neutron excess concept from previous studies of B and B reactors is extended to apply to linear-assembly B and B reactors, which allows the amount of starter fuel needed to establish a given equilibrium cycle to be calculated. Several example applications of the neutron excess formulation are given. First, an example model of a linear-assembly B and B reactor is analyzed to find the neutron excess cost of an equilibrium cycle. Second, simple one-dimensional models are used to predict the neutron excess value obtainable from different starter fuel configurations. Finally, these ideas are applied to design a fuel cycle consisting of linear-assembly B and B reactors and fuel recycling via a melt refining process. The neutron excess concept is used to design an appropriate starter fuel configuration made from melt refined fuel, which

  11. Caramel, uranium oxide fuel plates for water cooled reactors

    International Nuclear Information System (INIS)

    Bussy, Pierre; Delafosse, Jacques; Lestiboudois, Guy; Cerles, J.-M.; Schwartz, J.-P.

    1979-01-01

    The fuel is composed of thin plates assembled parallel to each other to form bundles or assemblies. Each plate is composed of a pavement of uranium oxide pellets, insulated from each other by a zircaloy cladding. The 235 U enrichment does not exceed 8%. The range of uses for this fuel extends from electric power generating reactors to irradiation reactors for research work. A parametric study in test loops has made it possible to determine the operating limits of this thick fuel, without bursting. The resulting diagram gives the permissible power densities, with and without cycling for specific burn-ups beyond 50,000 MWd/t. The thinnest plates were also irradiated in total in the form of advance assemblies irradiated in the core of the OSIRIS pile prior to its transformation. This transformation and the operation of this reactor with a core of 'Caramel' elements is the main trial experiment of this fuel [fr

  12. Thimble grip fuel assembly handling tool

    International Nuclear Information System (INIS)

    Salton, R.B.; Hornak, L.P.; Marshall, J.R.; Meuschke, R.E.

    1989-01-01

    This patent describes an apparatus for lifting a fuel assembly of a nuclear reactor. The fuel assembly consists of a top nozzle and control rod guide tubes. The apparatus having a gripping means comprised of: a life plate, an actuating plate having a plurality of apertures, the actuating plate disposed in spaced relationship below the lift plate and vertically movable relative thereto; gripping members operably associated with the lift and actuating plates, the gripping members comprising: (a) a vertical rod fixedly secured near its top end to the lift plate and projecting downward therefrom through an associated aperture in the actuating plate, the rod having a first frustoconical surface formed near its lower end, (b) a generally cylindrical, elastically deformable vertical sleeve having a bore therethrough with a first inner diameter, the sleeve having a first bevelled inside surface near the top end and a second bevelled inside surface at the bottom end of the sleeve, and (c) a vertical gripper actuator disposed about the rod

  13. Nuclear reactors and fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2014-07-01

    The Nuclear Fuel Center (CCN) of IPEN produces nuclear fuel for the continuous operation of the IEA-R1 research reactor of IPEN. The serial production started in 1988, when the first nuclear fuel element was delivered for IEA-R1. In 2011, CCN proudly presents the 100{sup th} nuclear fuel element produced. Besides routine production, development of new technologies is also a permanent concern at CCN. In 2005, U{sub 3}O{sub 8} were replaced by U{sub 3}Si{sub 2}-based fuels, and the research of U Mo is currently under investigation. Additionally, the Brazilian Multipurpose Research Reactor (RMB), whose project will rely on the CCN for supplying fuel and uranium targets. Evolving from an annual production from 10 to 70 nuclear fuel elements, plus a thousand uranium targets, is a huge and challenging task. To accomplish it, a new and modern Nuclear Fuel Factory is being concluded, and it will provide not only structure for scaling up, but also a safer and greener production. The Nuclear Engineering Center has shown, along several years, expertise in the field of nuclear, energy systems and correlated areas. Due to the experience obtained during decades in research and technological development at Brazilian Nuclear Program, personnel has been trained and started to actively participate in design of the main system that will compose the Brazilian Multipurpose Reactor (RMB) which will make Brazil self-sufficient in production of radiopharmaceuticals. The institution has participated in the monitoring and technical support concerning the safety, licensing and modernization of the research reactors IPEN/MB-01 and IEA-R1. Along the last two decades, numerous specialized services of engineering for the Brazilian nuclear power plants Angra 1 and Angra 2 have been carried out. The contribution in service, research, training, and teaching in addition to the development of many related technologies applied to nuclear engineering and correlated areas enable the institution to

  14. Nuclear reactors and fuel cycle

    International Nuclear Information System (INIS)

    2014-01-01

    The Nuclear Fuel Center (CCN) of IPEN produces nuclear fuel for the continuous operation of the IEA-R1 research reactor of IPEN. The serial production started in 1988, when the first nuclear fuel element was delivered for IEA-R1. In 2011, CCN proudly presents the 100 th nuclear fuel element produced. Besides routine production, development of new technologies is also a permanent concern at CCN. In 2005, U 3 O 8 were replaced by U 3 Si 2 -based fuels, and the research of U Mo is currently under investigation. Additionally, the Brazilian Multipurpose Research Reactor (RMB), whose project will rely on the CCN for supplying fuel and uranium targets. Evolving from an annual production from 10 to 70 nuclear fuel elements, plus a thousand uranium targets, is a huge and challenging task. To accomplish it, a new and modern Nuclear Fuel Factory is being concluded, and it will provide not only structure for scaling up, but also a safer and greener production. The Nuclear Engineering Center has shown, along several years, expertise in the field of nuclear, energy systems and correlated areas. Due to the experience obtained during decades in research and technological development at Brazilian Nuclear Program, personnel has been trained and started to actively participate in design of the main system that will compose the Brazilian Multipurpose Reactor (RMB) which will make Brazil self-sufficient in production of radiopharmaceuticals. The institution has participated in the monitoring and technical support concerning the safety, licensing and modernization of the research reactors IPEN/MB-01 and IEA-R1. Along the last two decades, numerous specialized services of engineering for the Brazilian nuclear power plants Angra 1 and Angra 2 have been carried out. The contribution in service, research, training, and teaching in addition to the development of many related technologies applied to nuclear engineering and correlated areas enable the institution to fulfill its mission that is

  15. Method for compacting spent nuclear reactor fuel rods

    International Nuclear Information System (INIS)

    Wachter, W.J.

    1988-01-01

    In a nuclear reactor system which requires periodic physical manipulation of spent fuel rods, the method of compacting fuel rods from a fuel rod assembly is described. The method consists of: (1) removing the top end from the fuel rod assembly; (2) passing each of multiple fuel rod pulling elements in sequence through a fuel rod container and thence through respective consolidating passages in a fuel rod directing chamber; (3) engaging one of the pulling elements to the top end of each of the fuel rods; (4) drawing each of the pulling elements axially to draw the respective engaged fuel rods in one axial direction through the respective the passages in the chamber to thereby consolidate the fuel rods into a compacted configuration of a cross-sectional area smaller than the cross-sectional area occupied thereby within the fuel rod assembly; and (5) drawing all of the engaged fuel rods concurrently and substantially parallel to one another in the one axial direction into the fuel rod container while maintaining the compacted configuration whereby the fuel rods are aligned within the container in a fuel rod density of the the fuel rod assembly

  16. Fuel transfer manipulator for liquid metal nuclear reactors

    International Nuclear Information System (INIS)

    Sturges, R.H.

    1983-01-01

    A manipulator for transferring fuel assemblies between inclined fuel chutes of a liquid metal nuclear reactor installation. Hoisting means are mounted on a mount supported by beams pivotably attached by pins to the mount and to the floor in such a manner that pivoting of the beams causes movement and tilting of a hoist tube between positions of alignment with the inclined chutes. (author)

  17. Scanning tunneling microscope assembly, reactor, and system

    Science.gov (United States)

    Tao, Feng; Salmeron, Miquel; Somorjai, Gabor A

    2014-11-18

    An embodiment of a scanning tunneling microscope (STM) reactor includes a pressure vessel, an STM assembly, and three spring coupling objects. The pressure vessel includes a sealable port, an interior, and an exterior. An embodiment of an STM system includes a vacuum chamber, an STM reactor, and three springs. The three springs couple the STM reactor to the vacuum chamber and are operable to suspend the scanning tunneling microscope reactor within the interior of the vacuum chamber during operation of the STM reactor. An embodiment of an STM assembly includes a coarse displacement arrangement, a piezoelectric fine displacement scanning tube coupled to the coarse displacement arrangement, and a receiver. The piezoelectric fine displacement scanning tube is coupled to the coarse displacement arrangement. The receiver is coupled to the piezoelectric scanning tube and is operable to receive a tip holder, and the tip holder is operable to receive a tip.

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  19. Nondestructive examination of Oconee 1 fuel assemblies after four cycles of irradiation

    International Nuclear Information System (INIS)

    Pyecha, T.D.; Mayer, J.T.; Guthrie, B.A. III; Riordan, J.E.

    1980-12-01

    Five B and W Mark B (15 x 15) pressurized water reactor fuel assemblies were nondestructively examined after four cycles of irradiation in the Oconee 1 reactor. Four of the five assemblies examined had a burnup of 40,000 MWd/mtU; the fifth assembly had a burnup of 36,800 MWd/mtU. This effort is part of a Department of Energy program to improve uranium utilization by extending the burnup of light water reactor fuel. The examinations were conducted in the Oconee 1 and 2 spent fuel storage pool. Data obtained included fuel assembly and fuel rod dimensions, water channel spacings, spacer grid and holddown spring forces, fuel column stack and axial gap lengths, and crud samples. The results indicate that the assemblies performed well through four cycles of operation; all of the data were within design limits

  20. Nuclear fuel assembly with improved spectral shift-producing rods

    International Nuclear Information System (INIS)

    Ferrari, H.M.

    1987-01-01

    This patent describes a nuclear reactor having fuel assemblies and a moderator-coolant liquid flowing through the fuel assemblies, each fuel assembly including an organized array of nuclear fuel rods wherein the moderator-coolant liquid flows along the fuel rods, at least one improved spectral shift-producing rod disposed among the fuel rods. The spectra shift-producing rod consists of: (a) an elongated hollow hermetically-sealed tubular member; (b) a weakened region formed in a portion of the member, the portion being subject to rupture at a given level of internal pressure; and (c) burnable poison material contained in the member which generates gas in the member as operation of the reactor proceeds normally, the material being soluble in the moderator-coolant liquid when brought into contact therewith; (d) the given level of internal pressure being less than the maximum level of internal pressure normally expected to be generated within the member by the poison material by normal operation of the reactor

  1. Fuel for new Russian reactor VVER-1200

    Energy Technology Data Exchange (ETDEWEB)

    Vasilchenko, Ivan Nikitovich [GRPress, 21, Ordzhonikidze Street, 142103 Podolsk, Moscow region (Russian Federation)

    2009-06-15

    A great program is accepted in Russia on increasing the nuclear power capacities. The basis of the program is commissioning of VVER-1200 Units of AES-2006 design. This is largely an evolutionary project of VVER-1000 reactor plant. It is referred also to reactor core. The plant electric power is increased due to increase in the reactor thermal power and forcing the main parameters and the efficiency increase. With this, reactor pressure increases from 15,7 to 16,2 MPa. The reactor inlet temperature increases from 290 deg. C to 298 deg. C, and outlet temperature from 319 deg. C to 329 deg. C. In a set of the design for four Units (2 Units at Novovoronezh NPP and 2 Units at Leningrad NPP) two base fuel cycles are developed: 5 year and 3 year. To provide such fuel cycles the fuel loading is increased by 8 tons, as compared to VVER-1000 base design, due to fuel column increase by 200 mm and change of fuel pellet sizes. In the mentioned fuel cycles the average burnup in the unloaded batch will be {approx}57 MW.day/kg U and 52 MW.day/kg U (maximum burnup over FAs is 64,5 MW.day/kg U and 60,3 MW.day/kg U), respectively. Specific consumption of natural uranium will be reduced by 5% as compared to that reached at VVER-1000 reactor. In spite of increase in Unit power the limiting permissible fuel rod linear heat rate is decreased from 448 W/cm to 420 W/cm. Refueling pattern is used with small neutron escape. The safety criteria are used that were established for VVER-1000, except for those that did not comply with EUR. For instance, the number of leaky fuel rods under accident is limited. The more stringent requirements are stated on efficiency margin of CPS rods for reactor shutdown that is ensured by the increased number of CPS rods. The well-proved design of fuel assembly TVS-2 and its close modification TVS-2M, operated at Balakovo NPP and Rostov NPP, is laid down in the basis of the core design. The load-carrying component of this structure is a rigid skeleton formed by

  2. PWR and BWR spent fuel assembly gamma spectra measurements

    Energy Technology Data Exchange (ETDEWEB)

    Vaccaro, S. [European Commission, DG Energy, Directorate EURATOM Safeguards Luxembourg (Luxembourg); Tobin, S.J.; Favalli, A. [Los Alamos National Laboratory, Los Alamos, NM (United States); Grogan, B. [Oak Ridge National Laboratory, Oak Ridge (United States); Jansson, P. [Uppsala University, Uppsala (Sweden); Liljenfeldt, H. [Oak Ridge National Laboratory, Oak Ridge (United States); Mozin, V. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Hu, J. [Oak Ridge National Laboratory, Oak Ridge (United States); Schwalbach, P. [European Commission, DG Energy, Directorate EURATOM Safeguards Luxembourg (Luxembourg); Sjöland, A. [Swedish Nuclear Fuel and Waste Management Company (SKB) (Sweden); Trellue, H.; Vo, D. [Los Alamos National Laboratory, Los Alamos, NM (United States)

    2016-10-11

    A project to research the application of nondestructive assay (NDA) to spent fuel assemblies is underway. The research team comprises the European Atomic Energy Community (EURATOM), embodied by the European Commission, DG Energy, Directorate EURATOM Safeguards; the Swedish Nuclear Fuel and Waste Management Company (SKB); two universities; and several United States national laboratories. The Next Generation of Safeguards Initiative–Spent Fuel project team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. This study focuses on spectrally resolved gamma-ray measurements performed on a diverse set of 50 assemblies [25 pressurized water reactor (PWR) assemblies and 25 boiling water reactor (BWR) assemblies]; these same 50 assemblies will be measured with neutron-based NDA instruments and a full-length calorimeter. Given that encapsulation/repository and dry storage safeguards are the primarily intended applications, the analysis focused on the dominant gamma-ray lines of {sup 137}Cs, {sup 154}Eu, and {sup 134}Cs because these isotopes will be the primary gamma-ray emitters during the time frames of interest to these applications. This study addresses the impact on the measured passive gamma-ray signals due to the following factors: burnup, initial enrichment, cooling time, assembly type (eight different PWR and six different BWR fuel designs), presence of gadolinium rods, and anomalies in operating history. To compare the measured results with theory, a limited number of ORIGEN-ARP simulations were performed.

  3. Stress analysis of fuel assemblies under seismic load

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  4. Experimental studies of resistance fretting-wear of fuel rods for VVER-1000 and TVS-KVADRAT fuel assemblies

    International Nuclear Information System (INIS)

    Makarov, V.; Afanasiev, A.; Egorov, Yu.; Matvienko, I.

    2015-01-01

    The paper covers the results of the studies performed to justify the wear resistance of fuel rods in contact with the spacer grids of TVS VVER-1000 fuel assembly and TVS-KVADRAT square fuel assembly of Russian design for PWR-900 reactor. The presented results of three testing stages comprise: Testing of mockup fuel rods of VVER TVS fuel assembly for fretting wear under the conditions of the water chemistry of VVER reactor; Testing models of different design embodiments of the fuel rods for VVER TVS fuel assembly for fretting wear in still cold water; Testing mockup fuel rods of TVS-KVADRAT square fuel assembly for PWR reactor for frettingwear under the conditions of PWR water chemistry. The effect of structural and operational factors was determined (amplitudes, fuel rod vibration frequencies, values of cladding-to-spacer grid cell gap for the depth of fuel rod cladding wear etc.), an assessment was made of the threshold values of fuel rod vibration parameters, which, if not exceeded, provide the absence of the fuel rod cladding fretting wear in the fuel rod-to spacer grid contact area. Key words: fretting wear, fuel rod, spacer grid, VVER, PWR (author)

  5. In-core fuel management for nuclear reactor

    International Nuclear Information System (INIS)

    Ross, M.F.; Visner, S.

    1986-01-01

    This patent describes in-core fuel management for nuclear reactor in which the first cycle of a pressurized water nuclear power reactor has a multiplicity of elongated, square fuel assemblies supported side-by-side to form a generally cylindrical, stationary core consisting entirely of fresh fuel assemblies. Each assembly of the first type has a substantially similar low average fissile enrichment of at least about 1.8 weight percent U-235, each assembly of the second type having a substantially similar intermediate average fissile enrichment at least about 0.4 weight percent greater than that of the first type, and each assembly of the third type having a substantially similar high average fissile enrichment at least about 0.4 weight percent greater than that of the intermediate type, the arrangement of the low, intermediate, and high enrichment assembly types which consists of: a generally cylindrical inner core region consisting of approximately two-thirds the total assemblies in the core and forming a figurative checkerboard array having a first checkerboard component at least two-thirds of which consists of high enrichment and intermediate enrichment assemblies, at least some of the high enrichment assemblies containing fixed burnable poison shims, and a second checkerboard component consisting of assemblies other than the high enrichment type; and a generally annular outer region consisting of the remaining assemblies and including at least some but less than two-thirds of the high enrichment type assemblies

  6. Development and use of the IVV-2M fuel assembly

    International Nuclear Information System (INIS)

    Aden, V.G.; Bulkin, Yu.M.; Vasenkov, V.I.

    1987-01-01

    The design and performance of a fuel assembly, intended for use in the water cooled and moderated IVV-2M research reactor and consisting of dioxide dispersed in an aluminium alloy matrix encased in an aluminium alloy can, is presented. Experimental and theoretical studies included neutron characteristics of the reactor core, thermohydraulic behavior, and thermal neutron flux distribution. Data are also presented on the specific charge in u 235, the metal to water volume ratio, the specific heat removal surface, and maximum power of the assembly

  7. Spent nuclear fuel discharges from US reactors 1993

    International Nuclear Information System (INIS)

    1995-02-01

    The Energy Information Administration (EIA) of the U.S. Department of Energy (DOE) administers the Nuclear Fuel Data Survey, Form RW-859. This form is used to collect data on fuel assemblies irradiated at commercial nuclear reactors operating in the United States, and the current inventories and storage capacities of those reactors. These data are important to the design and operation of the equipment and facilities that DOE will use for the future acceptance, transportation, and disposal of spent fuels. The data collected and presented identifies trends in burnup, enrichment, and spent nuclear fuel discharged form commercial light-water reactor as of December 31, 1993. The document covers not only spent nuclear fuel discharges; but also site capacities and inventories; canisters and nonfuel components; and assembly type characteristics

  8. Passive gamma analysis of the boiling-water-reactor assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Vo, D., E-mail: ducvo@lanl.gov [Los Alamos National Laboratory, Los Alamos, NM (United States); Favalli, A. [Los Alamos National Laboratory, Los Alamos, NM (United States); Grogan, B. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Jansson, P. [Uppsala University, Uppsala (Sweden); Liljenfeldt, H. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Mozin, V. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Schwalbach, P. [European Atomic Energy Community (EURATOM), Luxemburg (Luxembourg); Sjöland, A. [Swedish Nuclear Fuel and Waste Management Company, Stockholm (Sweden); Tobin, S.; Trellue, H. [Los Alamos National Laboratory, Los Alamos, NM (United States); Vaccaro, S. [European Atomic Energy Community (EURATOM), Luxemburg (Luxembourg)

    2016-09-11

    This research focused on the analysis of a set of stationary passive gamma measurements taken on the spent nuclear fuel assemblies from a boiling water reactor (BWR) using pulse height analysis data acquisition. The measurements were performed on 25 different BWR assemblies in 2014 at Sweden's Central Interim Storage Facility for Spent Nuclear Fuel (Clab). This study was performed as part of the Next Generation of Safeguards Initiative–Spent Fuel project to research the application of nondestructive assay (NDA) to spent fuel assemblies. The NGSI–SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay (NDA) measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. The final objective of this project is to quantify the capability of several integrated NDA instruments to meet the aforementioned goals using the combined signatures of neutrons, gamma rays, and heat. This report presents a selection of the measured data and summarizes an analysis of the results. Specifically, trends in the count rates measured for spectral lines from the following isotopes were analyzed as a function of the declared burnup and cooling time: {sup 137}Cs, {sup 154}Eu, {sup 134}Cs, and to a lesser extent, {sup 106}Ru and {sup 144}Ce. From these measured count rates, predictive algorithms were developed to enable the estimation of the burnup and cooling time. Furthermore, these algorithms were benchmarked on a set of assemblies not included in the standard assemblies set used by this research team.

  9. Parallel processing of neutron transport in fuel assembly calculation

    International Nuclear Information System (INIS)

    Song, Jae Seung

    1992-02-01

    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

  10. Detection of failed fuel rods in shrouded BWR fuel assemblies

    International Nuclear Information System (INIS)

    Baero, G.; Boehm, W.; Goor, B.; Donnelly, T.

    1988-01-01

    A manipulator and an ultrasonic testing (UT) technique were developed to identify defective fuel rods in shrouded BWR fuel assemblies. The manipulator drives a UT probe axially through the bottom tie plate into the water channels between the fuel rods. The rotating UT probe locates defective fuel rods by ingressed water which attenuates the UT-signal. (author)

  11. Discussion on the re-irradiated fuel assembly with damaged guide vanes

    International Nuclear Information System (INIS)

    Li Ligang

    2013-01-01

    In January 2011, during the second plant of CNNC Nuclear Power Operations Management Co., Ltd.(hereinafter referred to as the second plant) refueling outage, the visual inspection found the guide vanes of fuel assembly A had felling off. After the National Nuclear Safety Administration (NNSA) estimated and approved, the fuel assembly A was reloaded in the specified location of reactor core. During the refueling outage in March 2012, the fuel assembly A was removed again from the reactor core. Visual inspection confirmed that the fuel assembly A was complete and without abnormal changes. The practice provides reference for re-irradiated of fuel assembly with the same type of damaged guide vanes, and provides case support for standard development for the same type of re-irradiated fuel assembly with damaged guide vanes. (author)

  12. Spacer grid with mixing blades for nuclear fuel assembly

    International Nuclear Information System (INIS)

    Noailly, J.

    1986-01-01

    The spacer grid for nuclear fuel assembly has two sets of intersecting metal plates provided with blades and defining cells. The plates are fitted only with half-blades associated with a single grid opening. The half-blades of adjacent cells are arranged at 90deg C to each other and each plate has at most one half-blade at each corner of a cell. The invention concerns fuel assemblies of pressurized water reactors. The blades arranged on a single side of the plate provide a good hydraulic uniformity. The invention provides a uniform distribution of blades (and thus of absorbing material in each hydraulic cell) [fr

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

    International Nuclear Information System (INIS)

    Saad, M.; Broeskamp, M.; Hahn, H.; Bignan, G.; Boisset, M.; Silie, P.

    1995-01-01

    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.)

  14. The effect of the fuel rod friction force to the fuel assembly lateral mechanical characteristics

    International Nuclear Information System (INIS)

    Ha, Dong Geun; Jeon, Sang Youn; Suh, Jung Min

    2012-01-01

    The Fuel Assembly (FA) for light water reactor consists of hundreds of fuel rods, guide tubes, spacer grids, top/bottom nozzles. The guide tubes transmit vertical loads between the top and bottom nozzles, position the fuel rod support grids vertically, react the loads from the fuel rods that are applied to the grids, and provide some of the lateral load capability for the overall fuel assembly. The guide tubes are the structural members of the skeleton assembly. And the spacer grids maintain the fuel rod array by providing positive lateral restraint to the fuel rod but only frictional restraint in the axial direction. Figure 1 shows the outline of skeleton, FA and the location of guide tubes in the view of cross section. 17x17 FA has 24 guide tubes and one instrumentation tube. When the FA is in reactor, the lateral stiffness is one of very important factors from the view point of in reactor integrity of fuel assembly such as guarantee of the cool able geometry, the control rod insertion etc. The lateral stiffness of FA is mainly determined by skeleton lateral stiffness. And the fuel rods loaded in the spacer grids reinforce the FA lateral stiffness. Generally, fuel rods and spacer grids create the nonlinear friction force between fuel rod tube and grid spring/dimple against external lateral force of FA. Thus, it is necessary to study the contribution of the fuel rods friction force to the FA lateral stiffness. So, this paper is to show how much amount of the fuel rod grid interaction contributes to the FA lateral stiffness based on the test results

  15. Thermal-hydraulic and neutron-physical characteristics of a new SCWR fuel assembly

    International Nuclear Information System (INIS)

    Liu, X.J.; Cheng, X.

    2009-01-01

    A new fuel assembly design for a thermal supercritical water cooled reactor (SCWR) core is proposed. Compared to the existing fuel assemblies, the present fuel assembly has two-rows of fuel rods between the moderator channels, to achieve a more uniform moderation for all fuel rod cells, and subsequently, a more uniform radial power distribution. In addition, a neutron-kinetics/thermal-hydraulics coupling method is developed, to analyze the neutron-physical and thermal-hydraulic behavior of the fuel assembly designs. This coupling method is based on the sub-channel analysis code COBRA-IV for thermal-hydraulics and the neutron-kinetics code SKETCH-N for neutron-physics. Both the COBRA-IV code and the SKETCH-N code are accordingly modified. An interface is established for the data transfer between these two codes. This coupling method is applied to both the one-row fuel assemblies (previous design) and the two-row fuel assemblies (present design). The performance of the two types of fuel assemblies is compared. The results show clearly that the two-row fuel assembly has more favorable neutron-physical and thermal-hydraulic characteristics than the one-row fuel assembly. The effect of various parameters on the fuel assembly performance is discussed. The coupling method is proven to be well suitable for further applications to SCWR fuel assembly design analysis

  16. Improvements in or relating to cooling systems for nuclear fuel assemblies

    International Nuclear Information System (INIS)

    Ljubivy, A.G.; Batjukov, V.I.; Shkhian, T.G.; Fadeev, A.I.

    1980-01-01

    A cooling system is proposed which can be used to cool a set of nuclear fuel assemblies arranged in a reactor core or placed in a container for spent fuel assemblies. The object of the invention is to provide a system which would prevent leakage of coolant from the vessel in the event of a rupture of the coolant supply pipeline externally of the vessel. In the case of the reactor cooling system the level of the coolant is stopped from dropping below the level of the active portion of the fuel assemblies and thus prevents a breakdown of the reactor. (UK)

  17. Fuel elements of research reactors in China

    International Nuclear Information System (INIS)

    Zhou Yongmao; Chen Dianshan; Tan Jiaqiu

    1987-01-01

    This paper describes the current status of design, fabrication of fuel elements for research reactors in China, emphasis is placed on the technology of fuel elements for the High Flux Engineering Test Reactor (HFETR). (author)

  18. Design analysis of a new SCWR fuel assembly using a coupled method

    International Nuclear Information System (INIS)

    Liu Xiaojing; Yang Ting; Cheng Xu

    2011-01-01

    Among the six GEN-Ⅳ reactor concepts recommended by the Gen-Ⅳ International Forum (GIF), supercritical water-cooled reactor (SCWR) is the only reactor type with water as coolant. Compared to the existing reactors, it has economic advantage and technology continuity. Based on the newly developed coupling code, analysis on the square SCWR assembly is carried out in this paper. A new design concept of SCWR fuel assembly is proposed. The results achieved so far indicate favorable thermal-hydraulic performance and neutron-physical behavior of the new fuel assembly compared to the previous ones. (authors)

  19. Development of packagings for 'MONJU' blanket fuel assemblies

    International Nuclear Information System (INIS)

    Shibata, Kan; Ouchi, Yuichiro; Matsuzaki, Masaaki; Okuda, Yoshihisa

    1995-01-01

    Blanket assemblies for prototype Fast Breeder Reactor 'MONJU' are made at commercial fuel fabrication plants capable of handling deplete Uranium in Japan. For the purpose of transport the assemblies are inserted into a packaging that is set horizontally at the fabrication plants because of compatibility with equipment installed at the plants. On the other hand, the assemblies must be taken out from the packaging set vertically at 'MONJU' due to compatibility. For this reason development of a new packaging, which makes it possible to take assemblies in and out both horizontally and vertically, is needed to carry out transport of assemblies for reload. The development and fabrication of the packagings, taking about two years, were completed in March 1995. The packagings were used in transport of assemblies in June 1995 for the first change. This report introduces the outline of the packaging and confirmation tests done in the process of development. (author)

  20. Safeguards instrument to monitor spent reactor fuel

    International Nuclear Information System (INIS)

    Nicholson, N.; Dowdy, E.J.; Holt, D.M.; Stump, C.

    1981-01-01

    A hand-held instrument for monitoring irradiated nuclear fuel inventories located in water-filled storage ponds has been developed. This instrument provides sufficient precise qualitative and quantitative information to be useful as a confirmatory technique to International Atomic Energy Agency inspectors, and is believed to be of potential use to nuclear fuel managers and to operators of spent-fuel storage facilities, both at reactor and away-from-reactor, and to operators of nuclear fuel reprocessing plants. Because the Cerenkov radiation glow can barely be seen by the unaided eye under darkened conditions, a night vision device is incorporated to aid the operator in locating the fuel assembly to be measured. Beam splitting optics placed in front of the image intensifier and a preset aperture select a predetermined portion of the observed scene for measurement of the light intensity using a photomultiplier (PM) tube and digital readout. The PM tube gain is adjusted by use of an internal optical reference source, providing long term repeatability and instrument-to-instrument cnsistency. Interchangeable lenses accommodate various viewing and measuring conditions