WorldWideScience

Sample records for reactor peripheral assemblies

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

    International Nuclear Information System (INIS)

    Kobayashi, Kaoru.

    1998-01-01

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

  2. 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

  3. 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)

  4. 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.)

  5. 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.)

  6. 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

  7. 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.

  8. 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.)

  9. BWR type reactor

    International Nuclear Information System (INIS)

    Watanabe, Shoichi

    1983-01-01

    Purpose : To flatten the radial power distribution in the reactor core thereby improve the thermal performance of the reactor core by making the moderator-fuel ratio of fuel assemblies different depending on their position in the reactor core. Constitution : The volume of fuels disposed in the peripheral area of the reactor core is decreased by the increase of the volume of moderators in fuel assemblies disposed in the peripheral area of the reactor core to thereby make the moderator-fuel volume greater in the peripheral area than that in the central area. The moderator-fuel ratio adjustment is attained by making the number of water rods greater, decreasing the diameter of fuel pellets or decreasing the number of fuel pins in fuel assemblies disposed at the peripheral area of the reactor core as compared with fuel assemblies disposed at the central area of the reactor core. In this way, the infinite multiplication factors of fuels can be increased to thereby improve the reactor core performance. (Aizawa, K.)

  10. 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.)

  11. 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.)

  12. 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.)

  13. 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

  14. 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

  15. Circumferential nonuniformity of cladding radiation swelling of fast reactor peripheral fuel elements

    International Nuclear Information System (INIS)

    Reutov, V.F.; Farkhutdinov, K.G.

    1977-01-01

    The results are presented of the investigation into the perimeter radiation swelling of Kh18N10T stainless steel cladding in different cross sections of a peripheral fuel element of the BR-5 reactor. The fluence on the cladding is 1.8-2.9 x 10 22 fast neutr/cm 2 , the operating temperatures in different parts of the fuel element being 430 deg to 585 deg C. There has been observed circumferential non-uniformity of the distribution, concentration, and of the total volume of radiation cavities, which is due to temperature non-uniformity along the cladding perimeter. It is shown that such non-uniformity of radiation swelling of the cladding material may result in bending of the peripheral fuel element with regard to the fuel assembly sheath walls

  16. Reactor core of nuclear reactor

    International Nuclear Information System (INIS)

    Sasagawa, Masaru; Masuda, Hiroyuki; Mogi, Toshihiko; Kanazawa, Nobuhiro.

    1994-01-01

    In a reactor core, a fuel inventory at an outer peripheral region is made smaller than that at a central region. Fuel assemblies comprising a small number of large-diameter fuel rods are used at the central region and fuel assemblies comprising a great number of smalldiameter fuel rods are used at the outer peripheral region. Since a burning degradation rate of the fuels at the outer peripheral region can be increased, the burning degradation rate at the infinite multiplication factor of fuels at the outer region can substantially be made identical with that of the fuels in the inner region. As a result, the power distribution in the direction of the reactor core can be flattened throughout the entire period of the burning cycle. Further, it is also possible to make the degradation rate of fuels at the outer region substantially identical with that of fuels at the inner side. A power peak formed at the outer circumferential portion of the reactor core of advanced burning can be lowered to improve the fuel integrity, and also improve the reactor safety and operation efficiency. (N.H.)

  17. 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

  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. 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.)

  20. 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

  1. 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)

  2. 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

  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 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)

  5. Lateral restraint assembly for reactor core

    Science.gov (United States)

    Gorholt, Wilhelm; Luci, Raymond K.

    1986-01-01

    A restraint assembly for use in restraining lateral movement of a reactor core relative to a reactor vessel wherein a plurality of restraint assemblies are interposed between the reactor core and the reactor vessel in circumferentially spaced relation about the core. Each lateral restraint assembly includes a face plate urged against the outer periphery of the core by a plurality of compression springs which enable radial preloading of outer reflector blocks about the core and resist low-level lateral motion of the core. A fixed radial key member cooperates with each face plate in a manner enabling vertical movement of the face plate relative to the key member but restraining movement of the face plate transverse to the key member in a plane transverse to the center axis of the core. In this manner, the key members which have their axes transverse to or subtending acute angles with the direction of a high energy force tending to move the core laterally relative to the reactor vessel restrain such lateral movement.

  6. 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.)

  7. 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

  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. 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

  10. 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)

  11. 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.)

  12. Vibration analysis of reactor assembly internals for Prototype Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Chellapandi, P.; Jalaldeen, S.; Srinivasan, R.; Chetal, S.C.; Bhoje, S.B.

    2003-01-01

    Vibration analysis of the reactor assembly components of 500 MWe Prototype Fast Breeder Reactor (PFBR) is presented. The vibration response of primary pump as well as dynamic forces developed at its supports are predicted numerically. The stiffness properties of hydrostatic bearing are determined by formulating and solving governing fluid and structural mechanics equations. The dynamic forces exerted by pump are used as input data for the dynamic response of reactor assembly components, mainly inner vessel, thermal baffle and control plug. Dynamic response of reactor assembly components is also predicted for the pressure fluctuations caused by sodium free level oscillations. Thermal baffle (weir shell) which is subjected to fluid forces developed at the associated sodium free levels is analysed by formulating and solving a set of non-linear equations for fluids, structures and fluid structure interaction (FSI). The control rod drive mechanism is analysed for response under flow induced forces on the parts subjected to cross flow in the zone just above the core top, taking into account FSI between sheaths of control and safety rod and absorber pin bundle. Based on the analysis results, it is concluded that the reactor assembly internals are free from any risk of mechanical as well as flow induced vibrations. (author)

  13. Reactor core and initially loaded reactor core of nuclear reactor

    International Nuclear Information System (INIS)

    Koyama, Jun-ichi; Aoyama, Motoo.

    1989-01-01

    In BWR type reactors, improvement for the reactor shutdown margin is an important characteristic condition togehter with power distribution flattening . However, in the reactor core at high burnup degree, the reactor shutdown margin is different depending on the radial position of the reactor core. That is , the reactor shutdown margin is smaller in the outer peripheral region than in the central region of the reactor core. In view of the above, the reactor core is divided radially into a central region and as outer region. The amount of fissionable material of first fuel assemblies newly loaded in the outer region is made less than the amount of the fissionable material of second fuel assemblies newly loaded in the central region, to thereby improve the reactor shutdown margin in the outer region. Further, the ratio between the amount of the fissionable material in the upper region and that of the fissionable material in the lower portion of the first fuel assemblies is made smaller than the ratio between the amount of the fissionable material in the upper region and that of the fissionable material in the lower region of the second fuel assemblies, to thereby obtain a sufficient thermal margin in the central region. (K.M.)

  14. 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

  15. 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

  16. 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)

  17. 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

  18. 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

  19. Modular head assembly and method of retrofitting existing nuclear reactor facilities

    International Nuclear Information System (INIS)

    Malandra, L.J.; Ledue, R.J.; Hankinson, M.F.; Kowalski, E.F.

    1987-01-01

    A method is described of retrofitting existing nuclear reactor facilities so as to form a modular closure head assembly for a nuclear reactor pressure vessel, where the existing nuclear reactor facilities comprise control rod drive mechanism cooling systems which include vertically extending elbow air ducts inter-connecting vertically spaced upper and lower air manifolds. The elbow air ducts extend radially beyond the peripheral envelope of the closure head, comprising the steps of: removing the upper air manifold; removing the vertically extending elbow air ducts; capping the air ports of the lower air manifold which ports were previously fluidically connecting the lower air manifold to the vertically extending elbow air ducts; disposing vertically upwardly extending air exhaust ducts above the lower air manifold in such an manner that the air exhaust ducts are disposed within the peripheral envelope of the closure head; fluidically connecting exhaust fans to the upper regions of the air exhaust ducts; fluidically connecting the lower regions of the air exhaust ducts the lower air manifold; permanently securing lift rods to the closure head at positions disposed radially outwardly of the lower air manifold; attaching a seismic support platform to the lift rods; proving fluidic passage of the vertically extending air exhaust ducts through the seismic support platform; attaching a missile shield plate to the lift rods; and proving fluidic passage of the vertically extending air exhaust ducts through the missile shield plate

  20. Results of assembly test of HTTR reactor internals

    International Nuclear Information System (INIS)

    Maruyama, S.; Saikusa, A.; Shiozawa, S.; Tsuji, N.; Miki, T.

    1996-01-01

    The assembly test of the HTTR actual reactor internals had been carried out at the works, prior to their installation in the actual reactor pressure vessel(RPV) at the construction site. The assembly test consists of several items such as examining fabricating precision of each component and alignment of piled-up structures, measuring circumferential coolant velocity profile in the passage between the simulated RPV and the reactor internals as well as under the support plates, measuring by-pass flow rate through gaps between the reactor internals, and measuring the binding force of the core restraint mechanism. Results of the test showed good performance of the HTTR reactor internals. Installation of the reactor internals in the actual RPV was started at the construction site of HTTR in April, 1995. In the installation process, main items of the assembly test at the works were repeated to investigate the reproducibility of installation. (author). 5 refs, 11 figs

  1. 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.)

  2. 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.)

  3. 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

  4. 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.)

  5. 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.)

  6. Recirculation system for nuclear reactors

    International Nuclear Information System (INIS)

    Braun, H. E.; Dollard, W. J.; Tower, S. N.

    1980-01-01

    A recirculation system for use in pressurized water nuclear reactors to increase the output temperature of the reactor coolant, thereby achieving a significant improvement in plant efficiency without exceeding current core design limits. A portion of the hot outlet coolant is recirculated to the inlets of the peripheral fuel assemblies which operate at relatively low power levels. The outlet temperature from these peripheral fuel assemblies is increased to a temperature above that of the average core outlet. The recirculation system uses external pumps and introduces the hot recirculation coolant to the free space between the core barrel and the core baffle, where it flows downward and inward to the inlets of the peripheral fuel assemblies. In the unlikely event of a loss of coolant accident, the recirculation system flow path through the free space and to the inlets of the fuel assemblies is utilized for the injection of emergency coolant to the lower vessel and core. During emergency coolant injection, the emergency coolant is prevented from bypassing the core through the recirculation system by check valves inserted into the recirculation system piping

  7. 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.

  8. 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.)

  9. 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)

  10. 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

  11. 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.)

  12. 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

  13. 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

  14. 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.)

  15. 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

  16. 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.)

  17. 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

  18. 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

  19. 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.)

  20. 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.)

  1. 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

  2. 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)

  3. Transient bowing of core assemblies in advanced liquid metal fast reactors

    International Nuclear Information System (INIS)

    Kamal, S.A.; Orechwa, Y.

    1986-01-01

    Two alternative core restraint concepts are considered for a conceptual design of a 900 MWth liquid metal fast reactor core with a heterogeneous layout. The two concepts, known as limited free bowing and free flowering, are evaluated based on core bowing criteria that emphasize the enhancement of inherent reactor safety. The core reactivity change during a postulated loss of flow transient is calculated in terms of the lateral displacements and displacement-reactivity-worths of the individual assemblies. The NUBOW-3D computer code is utilized to determine the assembly deformations and interassembly forces that arise when the assemblies are subjected to temperature gradients and irradiation induced creep and swelling during the reactor operation. The assembly ducts are made of the ferritic steel HT-9 and remain in the reactor core for four-years at full power condition. Whereas both restraint systems meet the bowing criteria, a properly designed limited free bowing system appears to be more advantageous than a free flowering system from the point of view of enhancing the reactor inherent safety

  4. 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)

  5. 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

  6. 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.)

  7. 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.)

  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. 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

  10. Nuclear reactor shutdown control rod assembly

    International Nuclear Information System (INIS)

    Bilibin, K.

    1988-01-01

    This patent describes a nuclear reactor having a reactor core and a reactor coolant flowing therethrough, a temperature responsive, self-actuated nuclear reactor shutdown control rod assembly, comprising: an upper drive line terminating at its lower end with a substantially cylindrical wall member having inner and outer surfaces; a lower drive line having a lower end adapted to be attached to a neutron absorber; a ring movable disposed about the outer surface of the wall member of the upper drive line; thermal actuation means adapted to be in heat exchange relationship with coolant in an associated reactor core and in contact with the ring, and balls located within the openings in the upper drive line. When reactor coolant approaches a predetermined design temperature the actuation means moves the ring sufficiently so that the balls move radially out from the recess and into the space formed by the second portion of the ring thereby removing the vertical support for the lower drive line such that the lower drive line moves downwardly and inserts an associated neutron absorber into an associated reactor core resulting in automatic reduction of reactor power

  11. Reactor physics studies in the GCFR phase-II critical assembly

    International Nuclear Information System (INIS)

    Pond, R.B.

    1976-09-01

    The reactor physics studies performed in the gas cooled fast reactor (GCFR) mockup on ZPR-9 are covered. This critical assembly, designated Phase II in the GCFR program, had a single zone PuO 2 -UO 2 core composition and UO 2 radial and axial blankets. The assembly was built both with and without radial and axial stainless steel reflectors. The program included the following measurements: small-sample reactivity worths of reactor constituent materials (including helium); 238 U Doppler effect; uranium and plutonium reaction rate distributions; thorium, uranium, and plutonium α and reactor kinetics. Analysis of the measurements used ENDF/B-IV nuclear data; anisotropic diffusion coefficients were used to account for neutron streaming effects. Comparison of measurements and calculations to GCFR Phase I are also made

  12. Kinetic analysis of sub-prompt-critical reactor assemblies

    International Nuclear Information System (INIS)

    Das, S.

    1992-01-01

    Neutronic analysis of safety-related kinetics problems in experimental neutron multiplying assemblies has been carried out using a sub-prompt-critical reactor model. The model is based on the concept of a sub-prompt-critical nuclear reactor and the concept of instantaneous neutron multiplication in a reactor system. Computations of reactor power, period and reactivity using the model show excellent agreement with results obtained from exact kinetics method. Analytic expressions for the energy released in a controlled nuclear power excursion are derived. Application of the model to a Pulsed Fast Reactor gives its sensitivity between 4 and 5. (author). 6 refs., 4 figs., 1 tab

  13. 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.

  14. 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)

  15. 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)

  16. 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.)

  17. Absorbing rods for nuclear fast neutron reactor absorbing assembly

    International Nuclear Information System (INIS)

    Aji, M.; Ballagny, A.; Haze, R.

    1986-01-01

    The invention proposes a neutron absorber rod for neutron absorber assembly of a fast neutron reactor. The assembly comprises a bundle of vertical rods, each one comprising a stack of pellets made of a neutron absorber material contained in a long metallic casing with a certain radial play with regard to this casing; this casing includes traps for splinters from the pellets which may appear during reactor operation, at the level of contact between adjacent pellets. The present invention prevents the casing from rupture involved by the disintegration of the pellets producing pieces of boron carbide of high hardness [fr

  18. 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.)

  19. Experimental determination of the neutron source for the Argonauta reactor subcritical assembly

    Energy Technology Data Exchange (ETDEWEB)

    Renke, Carlos A.C.; Furieri, Rosanne C.A.A.; Pereira, Joao C.S.; Voi, Dante L.; Barbosa, Andre L.N., E-mail: renke@ien.gov.b [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2011-07-01

    The utilization of a subcritical assembly for the determination of nuclear parameters in a multiplier medium requires a well defined neutron source to carry out the experiments necessary for the acquisition of the desired data. The Argonauta research reactor installed at the Instituto de Engenharia Nuclear has a subcritical assembly, under development, to be coupled at the upper part of the reactor core that will provide the needed neutrons emerging from its internal thermal column made of graphite. In order to perform neutronic calculations to compare with the experimental results, it is necessary a precise knowledge of the emergent neutron flux that will be used as neutron source in the subcritical assembly. In this work, we present the thermal neutron flux profile determined experimentally via the technique of neutron activation analysis, using dysprosium wires uniformly distributed at the top of the internal thermal neutron column of the Argonauta reactor and later submitted to a detection system using Geiger-Mueller detector. These experimental data were then compared with those obtained through neutronic calculation using HAMMER and CITATION codes in order to validate this calculation system and to define a correct neutron source distribution to be used in the subcritical assembly. This procedure avoids a coupled neutronic calculation of the subcritical assembly and the reactor core. It has also been determined the dimension of the graphite pedestal to be used in the bottom of the subcritical assembly tank in order to smooth the emergent neutron flux at the reactor top. Finally, it is estimated the thermal neutron flux inside the assembly tank when filled with water. (author)

  20. 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)

  1. Reactor cell assembly for use in spectroscopy and microscopy applications

    Science.gov (United States)

    Grindstaff, Quirinus; Stowe, Ashley Clinton; Smyrl, Norm; Powell, Louis; McLane, Sam

    2015-08-04

    The present disclosure provides a reactor cell assembly that utilizes a novel design and that is wholly or partially manufactured from Aluminum, such that reactions involving Hydrogen, for example, including solid-gas reactions and thermal decomposition reactions, are not affected by any degree of Hydrogen outgassing. This reactor cell assembly can be utilized in a wide range of optical and laser spectroscopy applications, as well as optical microscopy applications, including high-temperature and high-pressure applications. The result is that the elucidation of the role of Hydrogen in the reactions studied can be achieved. Various window assemblies can be utilized, such that high temperatures and high pressures can be accommodated and the signals obtained can be optimized.

  2. Stackable Form-Factor Peripheral Component Interconnect Device and Assembly

    Science.gov (United States)

    Somervill, Kevin M. (Inventor); Ng, Tak-kwong (Inventor); Torres-Pomales, Wilfredo (Inventor); Malekpour, Mahyar R. (Inventor)

    2013-01-01

    A stackable form-factor Peripheral Component Interconnect (PCI) device can be configured as a host controller or a master/target for use on a PCI assembly. PCI device may comprise a multiple-input switch coupled to a PCI bus, a multiplexor coupled to the switch, and a reconfigurable device coupled to one of the switch and multiplexor. The PCI device is configured to support functionality from power-up, and either control function or add-in card function.

  3. 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

  4. 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)

  5. 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

  6. 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.

  7. 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)

  8. Development of conductor feedthrough module of LV electrical penetration assembly for research reactors

    International Nuclear Information System (INIS)

    Luo Zhiyuan; Wang Guangjin; Zhou Bin

    2007-01-01

    A LV electrical penetration assembly with perfusion sealing conductor feedthrough module was developed, which can be used for the connection of internal and external cables through the wall of the research reactor workshop. The LV electrical penetration assembly was combined with several independent modules. The maintenance and replacement of the assembly can be easily done in service. The sealing of conductor feedthrough module was achieved with the perfusion of self-extinguishing epoxy. The leakage between the conductor feedthrough module and the end plate module was blocked with rubber rings. The result of the leakage test and the electrical performance test for the samples of conductor feedthrough module satisfied the requirement of research reactor. The structure of the new electrical penetration assembly is simple and compact. It can be manufactured with mature technology and cost low price. The performance of the assembly is steady. It can be used widely in research reactors. (authors)

  9. Safe operation of critical assemblies and research reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1960-09-15

    Some countries have accumulated considerable experience in the operation of these reactors and have in the process developed safe practices. On the other hand, other countries which have recently acquired, or will soon acquire, such reactors do not have sufficient background of experience with them to have developed full knowledge regarding their safe operation. In this situation, the International Atomic Energy Agency has considered that it would be useful to make available to all its Member States a set of recommendations on the safe operation of these reactors, based on the accumulated experience and best practices. The Director General accordingly nominated a Pane Ion Safe Operation of Critical Assemblies and Research Reactors to assist the Agency's Secretariat in drafting such recommendations

  10. Method of monitoring fuel-rod vibrations in a nuclear fuel reactor

    International Nuclear Information System (INIS)

    Kawamura, Makoto; Takai, Katsuaki.

    1985-01-01

    Purpose: To monitor the vibration modes of fuel rods continuously and on real time during operation of a PWR type nuclear reactor. Method: Vibrations of fuel rods during reactor operation are mainly caused by the lateral flow of coolants flowing through the gaps at the joints of reactor core buffle plates into a reactor core and fretting damages may possibly be caused to the fuel rod support portions due to the vibrations. In view of the above, self-powered detectors are disposed at a plurality of axial positions for the respective peripheral fuel assemblies in adjacent with the buffle plates and the detection signals from neutron detectors, that is, the fluctuations in neutrons are subjected to a frequency analysis during the operation period. The neutron detectors are disposed at the periphery of the reactor core, because the fuel assemblies disposed at the peripheral portion directly undergo the lateral flow from the joints of the buffle plates and vibrates most violently. Thus, the vibration situations can be monitored continuously, in a three demensional manner and on real time. (Moriyama, K.)

  11. Spatial kinetics studies in liquid-metal fast breeder reactor critical assemblies

    International Nuclear Information System (INIS)

    Brumback, S.B.; Goin, R.W.; Carpenter, S.G.

    1988-01-01

    Recent measurements in the zero-power physics reactor have been used to study the effect of spatial decoupling in fast reactor critical assemblies of various sizes and compositions. Flux distributions in these assemblies had varying degrees of sensitivity to perturbations. Decoupling was investigated using rod-drop, boron-oscillator, and noise-coherence techniques, which emphasized different times following perturbations. Equilibrium flux distributions were also measured for subcritical configurations with inserted control rods. For most assemblies, accurate reactivity measurements were obtained by analyzing the power history from a single detector using inverse kinetics methods, assuming an instantaneous efficiency change for the detector. The instantaneous efficiency change assumption broke down, however, in assemblies with zones in which normal plutonium fuel was replaced by /sup 235/U fuel or fuel with a high /sup 240/Pu content. Flux redistributions caused by perturbations in these cores took several minutes to evolve

  12. Safe Operation of Critical Assemblies and Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1961-05-15

    This Manual is provided as a guide to the safe operation of critical assemblies and small research reactors. It is intended that it should be used by all authorities and persons concerned with, or responsible for, the use of such equipment, in addition to the scientists and technologists who are actually working with, or operating it. It is suggested that it will be of use to those wishing to design and manufacture, or purchase, critical assemblies or research reactors, as well as those already in possession of them, and that it will prove particularly helpful to those users who have no direct access to other collected sources of information. This Manual is not a set of rules or a code of practice, but a series of recommendations which must be interpreted with scientific judgement in their application to any particular problem. The guiding principles are given from which good operational procedures may be established and improved. The promulgation of rigid standards is both impossible and undesirable at the present time, since the topics discussed form part of a rapidly growing science and technology. Therefore, any recommendations made should not be used to restrict or inhibit future developments. The Manual is intended mainly for use in those Member States where there has been little experience in the operation of critical assemblies and research reactors. It has been compounded from the best practices which exist in Member States having a large amount of such experience, so that nothing in it should conflict with the best practices to be encountered in the field of safe operation.

  13. Safe Operation of Critical Assemblies and Research Reactors

    International Nuclear Information System (INIS)

    1961-01-01

    This Manual is provided as a guide to the safe operation of critical assemblies and small research reactors. It is intended that it should be used by all authorities and persons concerned with, or responsible for, the use of such equipment, in addition to the scientists and technologists who are actually working with, or operating it. It is suggested that it will be of use to those wishing to design and manufacture, or purchase, critical assemblies or research reactors, as well as those already in possession of them, and that it will prove particularly helpful to those users who have no direct access to other collected sources of information. This Manual is not a set of rules or a code of practice, but a series of recommendations which must be interpreted with scientific judgement in their application to any particular problem. The guiding principles are given from which good operational procedures may be established and improved. The promulgation of rigid standards is both impossible and undesirable at the present time, since the topics discussed form part of a rapidly growing science and technology. Therefore, any recommendations made should not be used to restrict or inhibit future developments. The Manual is intended mainly for use in those Member States where there has been little experience in the operation of critical assemblies and research reactors. It has been compounded from the best practices which exist in Member States having a large amount of such experience, so that nothing in it should conflict with the best practices to be encountered in the field of safe operation.

  14. Automatic coolant flow control device for a nuclear reactor assembly

    Science.gov (United States)

    Hutter, Ernest

    1986-01-01

    A device which controls coolant flow through a nuclear reactor assembly comprises a baffle means at the exit end of said assembly having a plurality of orifices, and a bimetallic member in operative relation to the baffle means such that at increased temperatures said bimetallic member deforms to unblock some of said orifices and allow increased coolant flow therethrough.

  15. 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)

  16. Safety considerations of new critical assembly for the Research Reactor Institute, Kyoto University

    International Nuclear Information System (INIS)

    Umeda, Iwao; Matsuoka, Naomi; Harada, Yoshihiko; Miyamoto, Keiji; Kanazawa, Takashi

    1975-01-01

    The new critical assembly type of nuclear reactor having three cores for the first time in the world was completed successfully at the Research Reactor Institute of Kyoto University in autumn of 1974. It is called KUCA (Kyoto University Critical Assembly). Safety of the critical assembly was considered sufficiently in consequence of discussions between the researchers of the institute and the design group of our company, and then many bright ideas were created through the discussions. This paper is described the new safety design of main equipments - oil pressure type center core drive mechanism, removable water overflow mechanism, core division mechanism, control rod drive mechansim, protection instrumentation system and interlock key system - for the critical assembly. (author)

  17. 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

  18. WWER-440 control assembly local power peaking investigation on LR-0 reactor

    International Nuclear Information System (INIS)

    Mikus, J.

    2002-01-01

    This paper presents information concerning the local power peaking problem induced by the WWER-440 control assembly and the investigation possibilities on the light water, zero power reactor LR-0 at the Nuclear Research Institute (NRI) Rez plc. A brief description is given about the disposable control assembly model, experimental arrangement and conditions on the LR-0 reactor with regard to the earlier performed investigations as well as to the relevant measurements to be realized in the near future.(abstract)

  19. 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

  20. 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)

  1. Intra-assembly flow redistribution in LMFBRs: a simple computational approach

    International Nuclear Information System (INIS)

    Khatib-Rahbar, M.; Cazzoli, E.G.

    1983-01-01

    The liquid metal fast breeder reactor (LMFBR) core consists of fuel, blanket, control, and shielding assemblies packed in a hexagonal configuration. Radial blanket assemblies occupy peripheral locations in the reactor core and are characterized by steep power gradients, while inner blanket assemblies are located within the fuel assembly region and have higher power levels but flatter distributions. It is due to the presence of this radial power gradient that large sodium temperature distributions exist at full power operation. However, at low power, low flow natural convection conditions, a significant flow redistribution takes place leading to considerable radial temperature flattening. The purpose of the present study is to formulate a simple flow-regime dependent model supported by experimental data for prediction of sodium temperature flattening due to buoyancy-induced flow redistribution in LMFBR subassemblies with significant radial power gradient

  2. A hybrid source-driven method to compute fast neutron fluence in reactor pressure vessel - 017

    International Nuclear Information System (INIS)

    Ren-Tai, Chiang

    2010-01-01

    A hybrid source-driven method is developed to compute fast neutron fluence with neutron energy greater than 1 MeV in nuclear reactor pressure vessel (RPV). The method determines neutron flux by solving a steady-state neutron transport equation with hybrid neutron sources composed of peripheral fixed fission neutron sources and interior chain-reacted fission neutron sources. The relative rod-by-rod power distribution of the peripheral assemblies in a nuclear reactor obtained from reactor core depletion calculations and subsequent rod-by-rod power reconstruction is employed as the relative rod-by-rod fixed fission neutron source distribution. All fissionable nuclides other than U-238 (such as U-234, U-235, U-236, Pu-239 etc) are replaced with U-238 to avoid counting the fission contribution twice and to preserve fast neutron attenuation for heavy nuclides in the peripheral assemblies. An example is provided to show the feasibility of the method. Since the interior fuels only have a marginal impact on RPV fluence results due to rapid attenuation of interior fast fission neutrons, a generic set or one of several generic sets of interior fuels can be used as the driver and only the neutron sources in the peripheral assemblies will be changed in subsequent hybrid source-driven fluence calculations. Consequently, this hybrid source-driven method can simplify and reduce cost for fast neutron fluence computations. This newly developed hybrid source-driven method should be a useful and simplified tool for computing fast neutron fluence at selected locations of interest in RPV of contemporary nuclear power reactors. (authors)

  3. Compactable control element assembly for a nuclear reactor

    International Nuclear Information System (INIS)

    Dupen, C.F.G.

    1976-01-01

    A description is given of a compactable control element assembly for a nuclear reactor in which the absorber pins of the assembly are compacted during downward movement of the pin and are returned to their uncompacted state when downward movement is stopped. The control element assembly comprises a support member longitudinally movable within a control assembly duct and a plurality of absorber pins supported laterally outward of the support member and within the duct by pairs of support arms. The absorber pins are pivotably mounted to the support arms and the support arms in turn are supported from the support member for upward pivotable movement in a longitudinal plane. As the support member is moved downward, the support arms pivot upwardly and the absorber pins move upwardly and inwardly towards the support member. When the support member is stopped the absorber pins return to their uncompacted position

  4. 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

  5. Application of assembly module to high-temperature gas-cooled reactor full-scope simulation system

    International Nuclear Information System (INIS)

    Li Sifeng; Li Fu; Ma Yuanle; Shi Lei

    2007-01-01

    According to the circumstances that exist in the reactor full-scope simulators development as long development cycle, very difficult upgrade and narrow range of applicability, a kind of new model was developed based on assembly module which root in Linux kernel and successfully applied to the design of high-temperature gas-cooled reactor full-scope simulator system. The simulation results are coincident with the experimental ones, and it indicates that the new model based on assembly module is feasible to design of high-temperature gas cooled reactor simulation system. (authors)

  6. 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

  7. Remote assembly and maintenance of fusion reactors

    International Nuclear Information System (INIS)

    Becquet, M.C.; Farfaletti-Casali, F.

    1991-01-01

    This paper intend to present the state of the art in the field of remote assembly and maintenance, including system analysis design and operation for controlled fusion device such as JET, and the next NET and ITER reactors. The operational constraints of fusion reactors with respect to temperature, radiations dose rates and cumulated doses are considered with the resulting design requirements. Concepts like articulated boom, in-vessel vehicle and blanket handling device are presented. The close relations between computer simulations and experimental validation of those concepts are emphasized to ensure reliability of the operational behavior. Mockups and prototypes in reduced and full scale, as operating machines are described to illustrate the progress in remote operations for fusion reactors. The developments achieved at the Institute for System Engineering and Informatics of the Joint Research Center, in the field of remote blanket maintenance, reliability assessment of RH systems and remote cut and welding of lips joints are considered. (author)

  8. WWER-440 type reactor core

    International Nuclear Information System (INIS)

    Mizov, J.; Svec, P.; Rajci, T.

    1987-01-01

    Assemblies with patly spent fuel of enrichment within 5 and 36 MWd/kg U or lower than the maximum enrichment of freshly charged fuel are placed in at least one of the peripheral positions of each hexagonal sector of the WWER-440 reactor type core. This increases fuel availability and reduces the integral neutron dose to the reactor vessel. The duration is extended of the reactor campaign and/or the mean fuel enrichment necessary for the required duration of the period between refuellings is reduced. Thus, fuel costs are reduced by 1 up to 3%. The results obtained in the experiment are tabulated. (J.B.). 1 fig., 3 tabs

  9. 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.)

  10. Reactor core

    International Nuclear Information System (INIS)

    Matsuura, Tetsuaki; Nomura, Teiji; Tokunaga, Kensuke; Okuda, Shin-ichi

    1990-01-01

    Fuel assemblies in the portions where the gradient of fast neutron fluxes between two opposing faces of a channel box is great are kept loaded at the outermost peripheral position of the reactor core also in the second operation cycle in the order to prevent interference between a control rod and the channel box due to bending deformation of the channel box. Further, the fuel assemblies in the second row from the outer most periphery in the first operation cycle are also kept loaded at the second row in the second operation cycle. Since the gradient of the fast neutrons in the reactor core is especially great at the outer circumference of the reactor core, the channel box at the outer circumference is bent such that the surface facing to the center of the reactor core is convexed and the channel box in the second row is also bent to the identical direction, the insertion of the control rod is not interfered. Further, if the positions for the fuels at the outermost periphery and the fuels in the second row are not altered in the second operation cycle, the gaps are not reduced to prevent the interference between the control rod and the channel box. (N.H.)

  11. Calculation of Savannah River K Reactor Mark-22 assembly LOCA/ECS power limits

    International Nuclear Information System (INIS)

    Fischer, S.R.; Farman, R.F.; Birdsell, S.A.

    1992-01-01

    This paper summarizes the results of TRAC-PF1/MOD3 calculations of Mark-22 fuel assembly of loss-of-coolant accident/emergency cooling system (LOCA/ECS) power limits for the Savannah River Site (SRS) K Reactor. This effort was part of a larger effort undertaken by the Los Alamos National Laboratory for the US Department of Energy to perform confirmatory power limits calculations for the SRS K Reactor. A method using a detailed three-dimensional (3D) TRAC model of the Mark-22 fuel assembly was developed to compute LOCA/ECS power limits. Assembly power was limited to ensure that no point on the fuel assembly walls would exceed the local saturation temperature. The detailed TRAC model for the Mark-22 assembly consisted of three concentric 3D vessel components which simulated the two targets, two fuel tubes, and three main flow channels of the fuel assembly. The model included 100% eccentricity between the assembly annuli and a 20% power tilt. Eccentricity in the radial alignment of the assembly annuli arises because axial spacer ribs that run the length of the fuel and targets are used. Wall-shear, interfacial-shear, and wall heat-transfer correlations were developed and implemented in TRAC-PF1/MOD3 specifically for modeling flow and heat transfer in the narrow ribbed annuli encountered in the Mark-22 fuel assembly design. We established the validity of these new constitutive models using separate-effects benchmarks. TRAC system calculations of K Reactor indicated that the limiting ECS-phase accident is a double-ended guillonite break in a process water line at the pump discharge (i.e., a PDLOCA). The fuel assembly with the minimum cooling potential is identified from this system calculation. Detailed assembly calculations then were performed using appropriate boundary conditions obtained from this limiting system LOCA. Coolant flow rates and pressure boundary conditions were obtained from this system calculation and applied to the detailed assembly model

  12. Analysis of assembly serial number usage in domestic light-water reactors

    International Nuclear Information System (INIS)

    Reich, W.J.; Moore, R.S.

    1991-05-01

    Domestic light-water reactor (LWR) fuel assemblies are identified by a serial number that is placed on each assembly. These serial numbers are used as identifiers throughout the life of the fuel. The uniqueness of assembly serial numbers is important in determining their effectiveness as unambiguous identifiers. The purpose of this study is to determine what serial numbering schemes are used, the effectiveness of these schemes, and to quantify how many duplicate serial numbers occur on domestic LWR fuel assemblies. The serial numbering scheme adopted by the American National Standards Institute (ANSI) ensures uniqueness of assembly serial numbers. The latest numbering scheme adopted by General Electric (GE), was also found to be unique. Analysis of 70,971 fuel assembly serial numbers from permanently discharged fuel identified 11,948 serial number duplicates. Three duplicate serial numbers were found when analysis focused on duplication within the individual fuel inventory at each reactor site, but these were traced back to data entry errors and will be corrected by the Energy Information Administration (EIA). There were also three instances where the serial numbers used to identify assemblies used for hot cell studies differed from the serial numbers reported to the EIA. It is recommended that fuel fabricators and utilities adhere to the ANSI serial numbering scheme to ensure serial number uniqueness. In addition, organizations collecting serial number information, should request that all known serial numbers physically attached or associated with each assembly be reported and identified by the corresponding number scheme. 10 refs., 5 tabs

  13. Correlation and flux tilt measurements of coupled-core reactor assemblies

    International Nuclear Information System (INIS)

    Harries, J.R.

    1976-01-01

    The systematics of coupling reactivity and time delay between cores have been investigated with a series of coupled-core assemblies on the AAEC Split-table Critical Facility. The assemblies were similar to the Universities' Training Reactor (UTR), but had graphite coupling region thickness of 450 mm, 600 mm and 800 mm. The coupling reactivity measured by both the cross-correlation of reactor noise and the flux tilt methods was stronger than for the UTRs, but showed a similar trend with core spacing. The cross-correlograms were analysed using the two-node model to derive the time delays between the cores. The time delays were compared with thermal neutron wave propagation, and found to be consistent when the time delays were added to the individual node response-function delays. (author)

  14. 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)

  15. 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

  16. Irradiation Testing Vehicles for Fast Reactors from Open Test Assemblies to Closed Loops

    Energy Technology Data Exchange (ETDEWEB)

    Sienicki, James J. [Argonne National Lab. (ANL), Argonne, IL (United States); Grandy, Christopher [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-12-15

    A review of irradiation testing vehicle approaches and designs that have been incorporated into past Sodium-Cooled Fast Reactors (SFRs) or envisioned for incorporation has been carried out. The objective is to understand the essential features of the approaches and designs so that they can inform test vehicle designs for a future U.S. Fast Test Reactor. Fast test reactor designs examined include EBR-II, FFTF, JOYO, BOR-60, PHÉNIX, JHR, and MBIR. Previous designers exhibited great ingenuity in overcoming design and operational challenges especially when the original reactor plant’s mission changed to an irradiation testing mission as in the EBRII reactor plant. The various irradiation testing vehicles can be categorized as: Uninstrumented open assemblies that fit into core locations; Instrumented open test assemblies that fit into special core locations; Self-contained closed loops; and External closed loops. A special emphasis is devoted to closed loops as they are regarded as a very desirable feature of a future U.S. Fast Test Reactor. Closed loops are an important technology for irradiation of fuels and materials in separate controlled environments. The impact of closed loops on the design of fast reactors is also discussed in this report.

  17. 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

  18. 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.)

  19. 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.)

  20. 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.)

  1. 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

  2. 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

  3. Seismic Response Analysis of Assembled Reactor Vessel Internals

    International Nuclear Information System (INIS)

    Je, Sang-Yun; Chang, Yoon-Suk; Kang, Sung-Sik

    2015-01-01

    RVIs (Reactor Vessel Internals) perform important safe-related functions such as upholding the nuclear fuel assembly as well as providing the coolant passage of the reactor core and supporting the control rod drive mechanism. Therefore, the components including RVIs have to be designed and constructed taking into account the structural integrity under various accident scenarios. The reliable seismic analysis is essentially demanded to maintain the safe-related functions of RVIs. In this study, a modal analysis was performed based on the previous researches to examine values of frequencies, mode shapes and participation factors. Subsequently, the structural integrity respecting to the earthquake was evaluated through a response spectrum analysis by using the output variables of modal analysis. In this study, the structural integrity of the assembled RVIs was carried out against the seismic event via the modal and response spectrum analyses. Even though 287MPa of the maximum stress value occurred at the connected region between UGS and CSA, which was lower than its allowable value. At present, fluid-structure interaction effects are being examined for further realistic simulation, which will be reported in the near future

  4. 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

  5. Reactor Dynamics Experiments with a Sub-Critical Assembly

    International Nuclear Information System (INIS)

    Miley, G.H.; Yang, Y.; Wu, L.; Momota, H.

    2004-01-01

    A resurgence in use of nuclear power is now underway worldwide. However due to the shutdown of many university research reactors , student laboratories must rely more heavily on use of sub-critical assemblies. Here a driven sub-critical is described that uses a cylindrical Inertial Electrostatic Confinement (IEC) device to provide a fusion neutron source. The small IEC neutron source would be inserted in a fuel element position, with its power input controlled externally at a control panel. This feature opens the way to use of the critical assembly for a number of transient experiments such as sub-critical pulsing and neutron wave propagation. That in turn adds important new insights and excitement for the student teaching laboratory

  6. 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

  7. Television alignment of mast assembly in refueling of nuclear reactor

    International Nuclear Information System (INIS)

    Kaufmann, J.W.; Swidwa, K.J.; Hornak, L.P.

    1990-01-01

    This patent describes the refueling of a nuclear reactor having component assemblies of at least one type and being disposed in a pit in a containment under water, the refueling being carried out with a mast movable axially and circumferentially for raising and lowering the component assemblies, a mechanism, connected to an end of the mast, cooperative with the mast, for engaging a component assembly to be raised by the mast, a television camera, and a television monitor having an image-reference indication, the mechanism being connected to the mast movable with the mast; the method of positioning the mechanism to engage the component assembly appropriately for raising and lowering. It comprises: mounting the camera on the mechanism movable therewith, suspending the mast in the water of the pit with the mechanism extending from the end of the mast in the pit in position to engage the component assembly

  8. 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.

  9. Drop performance test of conceptually designed control rod assembly for prototype generation IV sodium-cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Young Kyu; Lee, Jae Han; Kim, Hoe Woong; KIm, Sung Kyun; Kim, Jong Bum [Sodium-cooled Fast Reactor NSSS Design Division, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2017-06-15

    The control rod assembly controls reactor power by adjusting its position during normal operation and shuts down chain reactions by its free drop under scram conditions. Therefore, the drop performance of the control rod assembly is important for the safety of a nuclear reactor. In this study, the drop performance of the conceptually designed control rod assembly for the prototype generation IV sodium-cooled fast reactor that is being developed at the Korea Atomic Energy Research Institute as a next-generation nuclear reactor was experimentally investigated. For the performance test, the test facility and test procedure were established first, and several free drop performance tests of the control rod assembly under different flow rate conditions were then carried out. Moreover, performance tests under several types and magnitudes of seismic loading conditions were also conducted to investigate the effects of seismic loading on the drop performance of the control rod assembly. The drop time of the conceptually designed control rod assembly for 0% of the tentatively designed flow rate was measured to be 1.527 seconds, and this agrees well with the analytically calculated drop time. It was also observed that the effect of seismic loading on the drop time was not significant.

  10. 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.

  11. Data base of reactor physics experimental results in Kyoto University critical assembly experimental facilities

    International Nuclear Information System (INIS)

    Ichihara, Chihiro; Fujine, Shigenori; Hayashi, Masatoshi

    1986-01-01

    The Kyoto University critical assembly experimental facilities belong to the Kyoto University Research Reactor Institute, and are the versatile critical assembly constructed for experimentally studying reactor physics and reactor engineering. The facilities are those for common utilization by universities in whole Japan. During more than ten years since the initial criticality in 1974, various experiments on reactor physics and reactor engineering have been carried out using many experimental facilities such as two solidmoderated cores, a light water-moderated core and a neutron generator. The kinds of the experiment carried out were diverse, and to find out the required data from them is very troublesome, accordingly it has become necessary to make a data base which can be processed by a computer with the data accumulated during the past more than ten years. The outline of the data base, the data base CAEX using personal computers, the data base supported by a large computer and so on are reported. (Kako, I.)

  12. Manufacturing requirements of reactor assembly components for PFBR (Paper No. 041)

    International Nuclear Information System (INIS)

    Murty, C.G.K.; Bhoje, S.B.

    1987-02-01

    This paper enumerates the requirements of 500 MWe Prototype Fast Breeder Reactor (PFBR) components and considering the present state of art of Indian industry an analysis is made on the challenges to be faced in manufacture highlighting the areas needing development. The large sizes and weights of the components coupled with the limitations on shop facilities and ODC transport, demand part of the fabrication to be done at shop and balance assembly work as well as certain assembly machining operations to be done at site work shop. The stringent geometrical tolerances coupled with extensive destructive and non-destructive examinations call for balanced and low heat input welding techniques and special inspection equipment like electronic co-ordinate determination system. The present paper deals with the specific manufacturing problems of the main reactor components. (author)

  13. 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

  14. 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)

  15. 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

  16. 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)

  17. 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.)

  18. 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)

  19. 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.)

  20. 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

  1. Sensitivity coefficients of reactor parameters in fast critical assemblies and uncertainty analysis

    International Nuclear Information System (INIS)

    Aoyama, Takafumi; Suzuki, Takayuki; Takeda, Toshikazu; Hasegawa, Akira; Kikuchi, Yasuyuki.

    1986-02-01

    Sensitivity coefficients of reactor parameters in several fast critical assemblies to various cross sections were calculated in 16 group by means of SAGEP code based on the generalized perturbation theory. The sensitivity coefficients were tabulated and the difference of sensitivity coefficients was discussed. Furthermore, the uncertainty of calculated reactor parameters due to cross section uncertainty were estimated using the sensitivity coefficients and cross section covariance data. (author)

  2. Guide tube insert assembly for use in a nuclear reactor

    International Nuclear Information System (INIS)

    Hopkins, R.J.; Land, J.T.

    1992-01-01

    This patent describes an internals assembly for a nuclear reactor of the type including an upper support plate and an upper core plate, each having apertures for conducting control rod assemblies into an out of fuel assemblies with the apertures of the upper support plate being aligned with the apertures of the upper core plate, a guide tube insert assembly comprising: an elongated tubular body extending between at least one of the aligned apertures formed in the upper support plate and the upper core plate; guide plates within the elongated tubular body, each of the guide plates having a planar surface extending substantially perpendicular to an axial direction of the tubular body; at least one interconnecting means for interconnecting the guide plates into a guide tube insert assembly such that the guide plates are simultaneously mountable within and removable from the elongated body, and the periphery of each of the guide plates is spaced apart from the inner walls of the elongated tubular body at every point when the insert assembly is mounted within the tubular body, and a stabilizing means for securing the lowermost guide plate of the guide tube insert assembly within the elongated tubular body to prevent rotational and lateral movement between the guide tube insert assembly and the tubular body

  3. Methods of assembling and disassembling spider and burnable poison rod structures for nuclear reactors

    International Nuclear Information System (INIS)

    Walton, L.A.

    1981-01-01

    A method is described of joining burnable poison rods to the spider arms of a pressurised water power reactor fuel assembly which is proof against the reactor core environment but permits these rods to be removed from the spider simply, swiftly and delicately. (U.K.)

  4. Transient Analysis of a Gas-cooled Fast Reactor for Single Control Assembly Withdrawal

    International Nuclear Information System (INIS)

    Choi, Hangbok

    2014-01-01

    The Energy Multiplier Module (EMZ) system response has been evaluated for control assembly withdrawal transients. Currently the EM2 core is equipped with six cylindrical drum-type control assemblies in the reflector zone for excess reactivity control and power maneuvering during the operating core life. This study investigates the system response to the control assembly withdrawal accident with various rotational speeds and reactivity worth to determine feasible control assembly design requirements from the physics viewpoint. The simulations have been conducted for single control assembly withdrawal transients without scram by a gas-cooled reactor plant simulator, which is based on a simplified plant nodal model, including the point reactor kinetics, single channel core thermal-fluid model, and a turbo-machinery performance model. Simulations were conducted for the middle-of- cycle core, when the excess reactivity of the core is the highest. Control assembly withdrawal times were varied from 1 (runaway) to 180 sec and reactivity worth was varied from 100 to 400 pcm. For a single control assembly withdrawal, the simulation has shown that the peak fuel temperature is expected to be ~1820°C when the assembly worth is 200 pcm and the runaway time is 1 sec per 180 degree rotation. The peak temperature could be reduced to ~1780°C if the assembly is rotated out in a moderate speed such as 1 degree/sec. These peak temperatures give a thermal margin of 22 to 24% to the melting point of uranium carbide fuel. The results also indicate that the current design with a single control assembly worth of 314 pcm may need adjustments in the future design. (author)

  5. 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

  6. Assembly homogenization techniques for light water reactor analysis

    International Nuclear Information System (INIS)

    Smith, K.S.

    1986-01-01

    Recent progress in development and application of advanced assembly homogenization methods for light water reactor analysis is reviewed. Practical difficulties arising from conventional flux-weighting approximations are discussed and numerical examples given. The mathematical foundations for homogenization methods are outlined. Two methods, Equivalence Theory and Generalized Equivalence Theory which are theoretically capable of eliminating homogenization error are reviewed. Practical means of obtaining approximate homogenized parameters are presented and numerical examples are used to contrast the two methods. Applications of these techniques to PWR baffle/reflector homogenization and BWR bundle homogenization are discussed. Nodal solutions to realistic reactor problems are compared to fine-mesh PDQ calculations, and the accuracy of the advanced homogenization methods is established. Remaining problem areas are investigated, and directions for future research are suggested. (author)

  7. 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)

  8. 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

  9. Measurement and analysis on dynamic behaviour of parallel-plate assembly in nuclear reactors

    International Nuclear Information System (INIS)

    Chen Junjie; Guo Changqing; Zou Changchuan

    1997-01-01

    Measurement and analysis on dynamic behaviour of parallel-plate assembly in nuclear reactors have been explored. The electromagnetic method, a new method of measuring and analysing dynamic behaviour with the parallel-plate assembly as the structure of multi-parallel-beams joining with single-beam, has been presented. Theoretical analysis and computation results of dry-modal natural frequencies show good agreement with experimental measurements

  10. 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

  11. Safe operation of research reactors and critical assemblies. Code of practice and annexes. 1984 ed

    International Nuclear Information System (INIS)

    1984-01-01

    The safe operation of research reactors and critical assemblies (hereafter termed 'reactors') requires proper design, construction, management and supervision. This Code of Practice deals mainly with management and supervision. The provisions of the Code apply to the whole life of the reactor, including modification, updating and upgrading. The Code may be subject to revision in the light of experience and the state of technology. The Code is aimed at defining minimum requirements for the safe operation of reactors. Emphasis is placed on which safety requirements should be met rather than on specifying how these requirements may be met. The Code also provides guidance and information to persons and authorities responsible for the operation of reactors. The Code recommends that documents dealing with the operation of reactors and including safety analyses be prepared and submitted for review and approval to a regulatory body. Operation would be authorized on the understanding that it would comply with limits and conditions designed to ensure safety. The Code covers a wide range of reactor types, which gives rise to a variety of safety issues. Safety issues applicable to specific reactor types only (e.g. fast reactors) are not necessarily covered in this Code. Some of the recommendations in the Code are not directly applicable to critical assemblies. A recommendation may therefore be interpreted according to the type of reactor concerned. In such cases the words 'adequate' and 'appropriate' are used to mean 'adequate' or 'appropriate' for the type of reactor under consideration.

  12. Educational use of research reactor (KUR) and critical assembly (KUCA) at Kyoto University

    International Nuclear Information System (INIS)

    Misawa, Tsuyoshi; Unesaki, Hironobu; Ichihara, Chihiro; Pyeon, Cheol Ho; Shiroya, Seiji

    2005-01-01

    At Kyoto University Research Reactor Institute, a research reactor of 5MW (KUR) and a critical assembly (KUCA) have been used for educational purpose to train undergraduate or graduate students. Using KUR, basic experiments for neutron applications have been carried out, and KUCA has been used for the education of nuclear engineering and technology. Especially, using KUCA, a joint reactor laboratory course of graduate level is offered every summer since 1975 by nine associated Japanese universities, and more than 2200 students attended this course

  13. Japanese contributions to containment structure, assembly and maintenance and reactor building for ITER

    International Nuclear Information System (INIS)

    Shibanuma, Kiyoshi; Honda, Tsutomu; Kanamori, Naokazu

    1991-06-01

    Joint design work on Conceptual Design Activity of International Thermonuclear Experimental Reactor (ITER) with four parties, Japan, the United States, the Soviet Union and the European Community began in April 1988 and was successfully completed in December 1990. In Japan, the home team was established in wide range of collaboration between JAERI and national institute, universities and heavy industries. The Fusion Experimental Reactor (FER) Team at JAERI is assigned as a core of the Japanese home team to support the joint Team activity and mainly conducted the design and R and D in the area of containment structure, remote handling and plant system. This report mainly describes the Japanese contribution on the ITER containment structure, remote handling and reactor building design. Main areas of contributions are vacuum vessel, attaching locks, electromagnetic analysis, cryostat, port and service line layout for containment structure, in-vessel handling equipment design and analysis, blanket handling equipment design and related short term R and D for assembly and maintenance, and finally reactor building design and analysis based on the equipment and service line layout and components flow during assembly and maintenance. (author)

  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. 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)

  16. 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

  17. The approximate thermal-model-testing method for non-stationary temperature fields in central zones of fast reactor assemblies

    International Nuclear Information System (INIS)

    Mikhin, V.I.; Matukhin, N.M.

    2000-01-01

    The approach to generalization of the non-stationary heat exchange data for the central zones of the nuclear reactor fuel assemblies and the approximate thermal-model-testing criteria are proposed. The fuel assemblies of fast and water-cooled reactors with different fuel compositions have been investigated. The reason of the non-stationary heat exchange is the fuel-energy-release time dependence. (author)

  18. 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.)

  19. 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

  20. 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

  1. 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

  2. Closed Loop In-Reactor Assembly (CLIRA): a fast flux test facility test vehicle

    International Nuclear Information System (INIS)

    Oakley, D.J.

    1978-01-01

    The Closed Loop In-Reactor Assembly (CLIRA) is a test vehicle for in-core material and fuel experiments in the Fast Flux Test Facility (FFTF). The FFTF is a fast flux nuclear test reactor operated for the Department of Energy (DOE) by Westinghouse Hanford Company in Richland, Washington. The CLIRA is a removable/replaceable part of the Closed Loop System (CLS) which is a sodium coolant system providing flow and temperature control independent of the reactor coolant system. The primary purpose of the CLIRA is to provide a test vehicle which will permit testing of nuclear fuels and materials at conditions more severe than exist in the FTR core, and to isolate these materials from the reactor core

  3. Analysis of the rotation accident of assemblies in boiling water reactors

    International Nuclear Information System (INIS)

    Becerril-Gonzalez M, J. J.; Fuentes M, L.; Castillo M, J. A.; Ortiz S, J. J.; Perusquia de Cueto, R.

    2012-10-01

    For this work was analyzed the impact that would cause the load of a rotated fuel assembly in the behaviour of the core in the Cycle 14 of the Unit 1 of the nuclear power plant of Laguna Verde. To carry out this analysis the code Simulate-3 was used, with which was possible to analyze the behavior of the effective multiplication factor and the thermal limits (MAPRAT, MFLPD and MFLCPR). The rotation of fuel assemblies to 90, 180 and 270 grades was analyzed with regard to the design position, with 0, 1, 2 and 3 burnt cycles for these assemblies. The results show that the thermal limits remain inside the allowed values, therefore if this accident type happened the reactor could continue operating in a sure way. (Author)

  4. 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

  5. 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

  6. Stabilization of magnet assemblies of permanent magnet sodium flowmeters used in fast breeder reactors

    Energy Technology Data Exchange (ETDEWEB)

    Rajan, K.K., E-mail: kkrajan@igcar.gov.in; Vijayakumar, G.

    2014-08-15

    Highlights: • Stabilization procedure for ALNICO-5 permanent magnet material is evolved. • Effect of time and temperature on ALNICO-5 assembly is determined. • Suitability of ALNICO-5 flowmeters at high temperatures is established. • Temperature coefficient of flux density is determined. - Abstract: Permanent magnet flow meters (PMFMs) are used to measure the sodium flow in sodium cooled Fast Breeder Reactor Circuits. Prototype fast breeder reactor (PFBR) which is under construction at Kalpakkam is a 500 MWe, sodium cooled, pool type reactor. Sodium flow measurement in various loops of the reactor is of prime importance from operational and safety point of view. To measure the flow of electrically conducting liquid sodium, in primary and secondary circuit pipe lines of PFBR, permanent magnet flow meters are used. PMFM is a non-invasive device, which works on the principle of generation of motional EMF by magnetic forces exerted on the charges in a moving conductor. Flowmeters of different pipe sizes ranging from 10 mm to 200 mm pipe diameter are required for PFBR. Long term performance of the flowmeters mainly depends on stability of permanent magnets used in flowmeters to generate constant magnetic field in stainless steel (SS) pipes. This paper describes the effects of time and temperature on permanent magnet assemblies made of ALNICO-V used in PFBR flowmeters. The stabilization methodology for ALNICO-V permanent magnet assemblies is evolved and established. Loss of magnetic field strength with respect to time and temperatures is determined by experiments and found negligible.

  7. Reactor core for FBR type reactor

    International Nuclear Information System (INIS)

    Fujita, Tomoko; Watanabe, Hisao; Kasai, Shigeo; Yokoyama, Tsugio; Matsumoto, Hiroshi.

    1996-01-01

    In a gas-sealed assembly for a FBR type reactor, two or more kinds of assemblies having different eigen frequency and a structure for suppressing oscillation of liquid surface are disposed in a reactor core. Coolant introduction channels for introducing coolants from inside and outside are disposed in the inside of structural members of an upper shielding member to form a shielding member-cooling structure in the reactor core. A structure for promoting heat conduction between a sealed gas in the assembly and coolants at the inner side or the outside of the assembly is disposed in the reactor core. A material which generates heat by neutron irradiation is disposed in the assembly to heat the sealed gases positively by radiation heat from the heat generation member also upon occurrence of power elevation-type event to cause temperature expansion. Namely, the coolants flown out from or into the gas sealed-assemblies cause differential fluctuation on the liquid surface, and the change of the capacity of a gas region is also different on every gas-sealed assemblies thereby enabling to suppress fluctuation of the reactor power. Pressure loss is increased by a baffle plate or the like to lower the liquid surface of the sodium coolants or decrease the elevating speed thereof thereby suppressing fluctuation of the reactor power. (N.H.)

  8. 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)

  9. 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)

  10. 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

  11. 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)

  12. 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.

  13. 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

  14. Mechanical Analysis of an Innovative Assembly Box with Honeycomb Structures Designed for a High Performance Light Water Reactor

    International Nuclear Information System (INIS)

    Herbell, Heiko; Himmel, Steffen; Schulenberg, Thomas

    2008-01-01

    The High Performance Light Water Reactor (HPLWR) is a water cooled reactor concept of the 4. generation, operated at a pressure beyond the critical point of water. Assemblies of this innovative reactor concept need to be built with assembly and moderator boxes, like boiling water reactors, to provide enough moderator water between them to compensate the low coolant density in the core. Hot, superheated steam conditions, on the other hand, require thermally insulated box walls rather than solid box walls to reduce the heat up of the moderator water. As a new an innovative approach, this paper describes moderator- and assembly boxes built from stainless steel honeycomb sandwich structures, in which the honeycomb cells are filled with alumina for thermal insulation. In comparison to solid box walls, the use of the presented design can provide the same stiffness but allows a drastic reduction of structural material and thus less neutron absorption. Finite element analyses are used to verify the required stiffness, to identify stress concentrations and to optimize the design. (authors)

  15. 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)

  16. 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

  17. Inspection and replacement of baffle assembly screws inside American reactor vessels

    International Nuclear Information System (INIS)

    Neal, K.; Chaumont, J.C.

    1999-01-01

    The baffle assembly inside the vessel of a 900 MWe reactor designed by Framatome, is made up of 44 plates fixed on 8 horizontal supports by a system of about 1000 screws. These plates undergo high neutron flux and the problem of screw cracking appeared at the end of the eighties in the first-generation reactors. The first operation on a large scale concerning the screws of a Westinghouse type reactor, was performed on the Tihange-1 power plant where Framatome controlled 960 screws and replaced 91. In 1997 as a consequence of the Belgian and French feedback experience, American plant operators launched a vast program of preventive actions: material analysis, inspection of baffle plate screws and replacement of defective screws. This program was held in cooperation with EPRI (electric power research institute) and under the control of NRC (nuclear regulatory commission). Framatome Technologies Inc (FTI) was in charge of the in-situ inspection and replacement of the screws. FTI designed special tools and equipment adapted to the 2-loop American reactors but the basis ideas were those applied on the Tihange reactor. The successful experience of FTI has allowed the firm to be commissioned for 6 2-loops American reactors. (A.C.)

  18. Electric arc apparatus for severing split-pin assemblies of guide tubes of nuclear reactors

    International Nuclear Information System (INIS)

    Burns, D.C.; Kauric, C.E.; Persang, J.C.

    1987-01-01

    This patent describes an apparatus for use in the replacement of an old split-pin assembly of a guide tube of a nuclear reactor by a new split-pin assembly, the old split-pin assembly including an old split pin and an old nut securing the old split pin to the guide tube, the old split-pin assembly and the guide tube being radioactive. The apparatus includes a metal disintegration machining tool, the tool having an electrode, means for mounting the tool submerged in a pool of water in engagement with the guide tube and with the old split-pin assembly secured to the guide tube, the tool being so mounted with the electrode in position to coact electrically with the last-named old split-pin assembly but not with the guide tube, and means, connected to the tool, for firing a disintegrating arc between the electrode and the assembly to disintegrate the assembly into readily removable fragments

  19. 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.

  20. 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

  1. 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

  2. Pump/heat exchanger assembly for pool-type reactor

    International Nuclear Information System (INIS)

    Nathenson, R.D.; Slepian, R.M.

    1987-01-01

    A heat exchanger and pump assembly comprising a heat exchanger including a housing for defining an annularly shaped cavity and supporting therein a plurality of heat transfer tubes. A pump is disposed beneath the heat exchanger and is comprised of a plurality of flow couplers disposed in a circular array. Each flow coupler is comprised of a pump duct for receiving a first electrically conductive fluid, i.e. the primary liquid metal, from a pool thereof, and a generator duct for receiving a second electrically conductive fluid, i.e. the intermediate liquid metal. The primary liquid metal is introduced from the reactor pool into the top, inlet ends of the tubes, flowing downward therethrough to be discharged from the tubes' bottom ends directly into the reactor pool. The primary liquid metal is variously introduced into the pump ducts directly from the reactor pool, either from the bottom or top end of the flow coupler. The intermediate fluid introduced into the generator ducts via the inlet duct and inlet plenum and after leaving the generator ducts passes through the annular cavity of the exchanger to cool the primary liquid in the tubes. The annular magnetic field of the pump is produced by a circular array of electromagnets having hollow windings cooled by a flow of the intermediate metal. (author)

  3. Design of a mixed recharge with MOX assemblies of greater relation of moderation for a BWR reactor

    International Nuclear Information System (INIS)

    Ramirez S, J.R.; Alonso V, G.; Palacios H, J.

    2004-01-01

    The study of the fuel of mixed oxides of uranium and plutonium (MOX) it has been topic of investigation in many countries of the world and those are even discussed in many places the benefits of reprocessing the spent fuel to extract the plutonium created during the irradiation of the fuel in the nuclear power reactors. At the moment those reactors that have been loaded partially with MOX fuel, are mainly of the type PWR where a mature technology has been achieved in some countries like they are France, Belgium and England, however the experience with reactors of the type BWR is more limited and it is continued studying the best way to introduce this type of fuel in BWRs, one of the main problems to introduce MOX in reactors BWR is the neutronic design of the same one, existing different concepts to introduce the plutonium in the assemblies of fuel and one of them is the one of increasing the relationship of moderation of the assemble. In this work a MOX fuel assemble design is presented and the obtained results so far in the ININ. These results indicate that the investigated concept has some exploitable advantages in the use of the MOX fuel. (Author)

  4. 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

  5. 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)

  6. Feasible reactor power cutback logic development for an integral reactor

    International Nuclear Information System (INIS)

    Han, Soon-Kyoo; Lee, Chung-Chan; Choi, Suhn; Kang, Han-Ok

    2013-01-01

    Major features of integral reactors that have been developed around the world recently are simplified operating systems and passive safety systems. Even though highly simplified control system and very reliable components are utilized in the integral reactor, the possibility of major component malfunction cannot be ruled out. So, feasible reactor power cutback logic is required to cope with the malfunction of components without inducing reactor trip. Simplified reactor power cutback logic has been developed on the basis of the real component data and operational parameters of plant in this study. Due to the relatively high rod worth of the integral reactor the control rod assembly drop method which had been adapted for large nuclear power plants was not desirable for reactor power cutback of the integral reactor. Instead another method, the control rod assembly control logic of reactor regulating system controls the control rod assembly movements, was chosen as an alternative. Sensitivity analyses and feasibility evaluations were performed for the selected method by varying the control rod assembly driving speed. In the results, sensitivity study showed that the performance goal of reactor power cutback system could be achieved with the limited range of control rod assembly driving speed. (orig.)

  7. 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

  8. Control assembly materials for water reactors: Experience, performance and perspectives. Proceedings of a technical committee meeting

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-02-01

    The safe, reliable and economic operation of water cooled nuclear power reactors depends to a large extent upon the reliable operation of control assemblies for the regulation and shutdown of the reactors. These consist of neutron absorbing materials clad in stainless steel or zirconium based alloys, guide tubes and guide cards, and other structural components. Current designs have worked extremely well in normal conditions, but less than ideal behaviour limits the lifetimes of control materials, imposing an economic penalty which acts as a strong incentive to produce improved materials and designs that are more reliable. Neutron absorbing materials currently in use include the ceramic boron carbide, the high melting point metal hafnium and the low melting point complex alloy Ag-In-Cd. Other promising neutron absorbing materials, such as dysprosium titanate, are being evaluated in the Russian Federation. These control materials exhibit widely differing mechanical, physical and chemical properties, which must be understood in order to be able to predict the behaviour of control rod assemblies. Identification of existing failure mechanisms, end of life criteria and the implications of the gradual introduction of extended burnup, mixed oxide (MOX) fuels and more complex fuel cycles constitutes the first step in a search for improved materials and designs. In the early part of this decade, it was recognized by the International Working Group on Fuel Performance and Technology (IWGFPT) that international conferences, symposia and published reviews on the materials science aspects of control assemblies were few and far between. Consequently, the IWGFPT recommended that the IAEA should rectify this situation with a series of Technical Committee meetings (TCMs) devoted entirely to the materials aspects of reactor control assemblies. The first was held in 1993 and in the intervening five years considerable progress has been made. In bringing together experts in the

  9. Control assembly materials for water reactors: Experience, performance and perspectives. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    2000-02-01

    The safe, reliable and economic operation of water cooled nuclear power reactors depends to a large extent upon the reliable operation of control assemblies for the regulation and shutdown of the reactors. These consist of neutron absorbing materials clad in stainless steel or zirconium based alloys, guide tubes and guide cards, and other structural components. Current designs have worked extremely well in normal conditions, but less than ideal behaviour limits the lifetimes of control materials, imposing an economic penalty which acts as a strong incentive to produce improved materials and designs that are more reliable. Neutron absorbing materials currently in use include the ceramic boron carbide, the high melting point metal hafnium and the low melting point complex alloy Ag-In-Cd. Other promising neutron absorbing materials, such as dysprosium titanate, are being evaluated in the Russian Federation. These control materials exhibit widely differing mechanical, physical and chemical properties, which must be understood in order to be able to predict the behaviour of control rod assemblies. Identification of existing failure mechanisms, end of life criteria and the implications of the gradual introduction of extended burnup, mixed oxide (MOX) fuels and more complex fuel cycles constitutes the first step in a search for improved materials and designs. In the early part of this decade, it was recognized by the International Working Group on Fuel Performance and Technology (IWGFPT) that international conferences, symposia and published reviews on the materials science aspects of control assemblies were few and far between. Consequently, the IWGFPT recommended that the IAEA should rectify this situation with a series of Technical Committee meetings (TCMs) devoted entirely to the materials aspects of reactor control assemblies. The first was held in 1993 and in the intervening five years considerable progress has been made. In bringing together experts in the

  10. Completion of the fabrication and assembly of the internal parts and pressure vessel of the LABGENE reactor

    International Nuclear Information System (INIS)

    Guimaraes, Leonam dos Santos

    2005-01-01

    The Navy's Technological Center in Sao Paulo (CTMSP) has successfully concluded in 2005 the final assembly of the internals of the Laboratory of Energy Generation's Reactor (LABGENE). This structure together with the fuel elements and the control rods drives mechanisms are part of a PWR type Nuclear Reactor. (author)

  11. 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

  12. 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

  13. 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)

  14. Research on the reactor physics using the Kyoto University Critical Assembly (KUCA)

    International Nuclear Information System (INIS)

    1986-10-01

    The Kyoto University Critical Assembly [KUCA] is a multi-core type critical assembly established in 1974, as a facility for the joint use study by researchers of all universities in Japan. Thereafter, many reactor physics experiments have been carried out using three cores (A-, B-, and C-cores) in the KUCA. In the A- and B-cores, solid moderator such as polyethylene or graphite is used, whereas light-water is utilized as moderator in the C-core. The A-core has been employed mainly in connection with the Cockcroft-Walton type accelerator installed in the KUCA, to measure (1) the subcriticality by the pulsed neutron technique for the critical safety research and (2) the neutron spectrum by the time-of-flight technique. Recently, a basic study on the tight lattice core has also launched using the A-core. The B-core has been employed for the research on the thorium fuel cycle ever since. The C-core has been employed (1) for the basic studies on the nuclear characteristics of light-water moderated high-flux research reactors, including coupled-cores, and (2) for a research related to reducing enrichment of uranium fuel used in research reactors. The C-core is being utilized in the reactor laboratory course experiment for students of ten universities in Japan. The data base of the KUCA critical experiments is generated so far on the basis of approximately 350 experimental reports accumulated in the KUCA. Besides, the assessed KUCA code system has been established through analyses on the various KUCA experiments. In addition to the KUCA itself, both of them are provided for the joint use study by researchers of all universities in Japan. (author)

  15. BWR type nuclear reactors

    International Nuclear Information System (INIS)

    Yamamoto, Toru.

    1987-01-01

    Purpose: To obtain reactor core characteristics with less changes in the excess reactivity due to fuel burnup even when the operation period varies. Constitution: In a BWR type reactor where fuel assemblies containing fuel rods incorporated with burnable poisons are arranged, the fuel assemblies are grouped into first fuel assemblies and second fuel assemblies. Then, the number of fuel rods incorporated with burnable poisons within the first fuel assemblies is made greater than that of the second fuel rods, while the concentration of the burnable poisons in the fuel rods incorporated with the burnable poisons in the first fuel assemblies is made lower than that of the fuel rods incorporated with the burnable poisons in the second fuel assemblies. In the BWR type reactor constituted in this way, the reactor core characteristics can be improved by changing the ratio between the first fuel assemblies and the second fuel assemblies charged to the reactor core, thereby decreasing the changes in the burnup of the excess reactivity. (Kamimura, M.)

  16. Recovery of reactor electrical assemblies using differential de-encapsulation to remove dielectric insulation systems

    International Nuclear Information System (INIS)

    Hubrig, J.G.; Hammerstone, E.B.

    1986-01-01

    State-of-the-art de-encapsulation technologies associated with the conventional dielectric insulation systems employed in the construction of electrical coils and power distribution systems do not allow for accurate fatigue/failure analysis or reliable recovery of costly assembly components. Differential de-encapsulation allows for the selective removal of contemporary thermoset resin based insulation systems to allow non-destructive penetration of insulation wall thicknesses to both examine critical areas and recover high performance metallic and non-metallic inserts for remanufacture; significantly reducing replacement costs and reactor downtime. The authors' analysis describes how the availability of engineering data from the selective and non-destructive removal of insulation materials will aid in the evaluation of original manufacture, materials and procedures; enabling redesign to enhance subsequent on line performance. They also discuss why the ability to recover coil and core assemblies for remanufacture will have a major economic impact on reactor management costs

  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. 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.)

  19. A Mixed-Oxide Assembly Design for Rapid Disposition of Weapons Plutonium in Pressurized Water Reactors

    International Nuclear Information System (INIS)

    Alonso, Gustavo; Adams, Marvin L.

    2002-01-01

    We have created a new mixed-oxide (MOX) fuel assembly design for standard pressurized water reactors (PWRs). Design goals were to maximize the plutonium throughput while introducing the lowest perturbation possible to the control and safety systems of the reactor. Our assembly design, which we call MIX-33, offers some advantages for the disposition of weapons-grade plutonium; it increases the disposition rate by 8% while increasing the worth of control material, compared to a previous Westinghouse design. The MIX-33 design is based upon two ideas: the use of both uranium and plutonium fuel pins in the same assembly, and the addition of water holes in the assembly. The main result of this paper is that both of these ideas are effective at increasing Pu throughput and increasing the worth of control material. With this new design, according to our analyses, we can transition smoothly from a full low-enriched-uranium (LEU) core to a full MIX-33 core while meeting the operational and safety requirements of a standard PWR. Given an interruption of the MOX supply, we can transition smoothly back to full LEU while meeting safety margins and using standard LEU assemblies with uniform pinwise enrichment distribution. If the MOX supply is interrupted for only one cycle, the transition back to a full MIX-33 core is not as smooth; high peaking could cause power to be derated by a few percent for a few weeks at the beginning of one transition cycle

  20. A device for supporting a pin bundle in a nuclear reactor assembly casing

    International Nuclear Information System (INIS)

    Marmonier, Pierre; Mesnage, Bernard; Teulon, Jean; Vayra, Jean; Venobre, Henri.

    1974-01-01

    Description is given of a device for supporting a pin-bundle in a nuclear reactor assembly casing. That device comprises a member coaxially mounted at the bottom of the vertically mounted casing, adapted to support a plurality of parallel rails along whose edges slide grooves made in the pin-plugs. It is characterized in that said supporting member is provided with a lateral groove open toward its periphery, cooperating with clamping-lugs that form extensions of the rail-sides and comprise an inwardly directed portion adapted to be engaged in the groove. This can be applied to fast neutron nuclear reactors [fr

  1. 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)

  2. 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.

  3. Reactor Vessel Surveillance Program for Advanced Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Kyeong-Hoon; Kim, Tae-Wan; Lee, Gyu-Mahn; Kim, Jong-Wook; Park, Keun-Bae; Kim, Keung-Koo

    2008-10-15

    This report provides the design requirements of an integral type reactor vessel surveillance program for an integral type reactor in accordance with the requirements of Korean MEST (Ministry of Education, Science and Technology Development) Notice 2008-18. This report covers the requirements for the design of surveillance capsule assemblies including their test specimens, test block materials, handling tools, and monitors of the surveillance capsule neutron fluence and temperature. In addition, this report provides design requirements for the program for irradiation surveillance of reactor vessel materials, a layout of specimens and monitors in the surveillance capsule, procedures of installation and retrieval of the surveillance capsule assemblies, and the layout of the surveillance capsule assemblies in the reactor.

  4. Vessel supporting structure for liquid metal cooled nuclear reactors

    International Nuclear Information System (INIS)

    Mahe, Armel; Jullien, Georges

    1974-01-01

    The supporting structure described is for a liquid metal cooled nuclear reactor, the vessel being of the type suspended to the end slab of the reactor. It includes a ring connected at one of its two ends to a single shell and at the other end to two shells. One of these three shells connected to the lower end of the ring forms the upper part of the vessel to be supported. The two other shells are embedded in two sperate parts of the slab. The ring and shell assembly is housed in an annular space provided in the end slab and separating it into two parts, namely a central part and a peripheral part [fr

  5. 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

  6. 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

  7. Vectorization and parallelization of a production reactor assembly code

    International Nuclear Information System (INIS)

    Vujic, J.L.; Martin, W.R.; Michigan Univ., Ann Arbor, MI

    1991-01-01

    In order to use efficiently the new features of supercomputers, production codes, usually written 10 -20 years ago, must be tailored for modern computer architectures. We have chosen to optimize the CPM-2 code, a production reactor assembly code based on the collision probability transport method. Substantial speedup in the execution times was obtained with the parallel/vector version of the CPM-2 code. In addition, we have developed a new transfer probability method, which removes some of the modelling limitations of the collision probability method encoded in the CPM-2 code, and can fully utilize the parallel/vector architecture of a multiprocessor IBM 3090. (author)

  8. Vectorization and parallelization of a production reactor assembly code

    International Nuclear Information System (INIS)

    Vujic, J.L.; Martin, W.R.

    1991-01-01

    In order to efficiently use new features of supercomputers, production codes, usually written 10 - 20 years ago, must be tailored for modern computer architectures. We have chosen to optimize the CPM-2 code, a production reactor assembly code based on the collision probability transport method. Substantial speedups in the execution times were obtained with the parallel/vector version of the CPM-2 code. In addition, we have developed a new transfer probability method, which removes some of the modelling limitations of the collision probability method encoded in the CPM-2 code, and can fully utilize parallel/vector architecture of a multiprocessor IBM 3090. (author)

  9. Numerical Analysis on the Free Fall Motion of the Control Rod Assembly for the Sodium Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Se-Hong; Choi, Choengryul; Son, Sung-Man [ELSOLTEC, Yongin (Korea, Republic of); Kim, Jae-Yong; Yoon, Kyung-Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    On receiving the scram signal, the control rod assemblies are released to fall into the reactor core by its weight. Thus drop time and falling velocity of the control rod assembly must be estimated for the safety evaluation. However, because of its complex shape, it is difficult to estimate the drop time by theoretical method. In this study, numerical analysis has been carried out in order to estimate drop time and falling velocity of the control rod assembly to provide the underlying data for the design optimization. Numerical analysis has been carried out to estimate the drop time and falling velocity of the control rod assembly for sodium-cooled fast reactor. Before performing the numerical analysis for the control rod assembly, sphere dropping experiment has been carried out for verification of the CFD methodology. The result of the numerical analysis for the method verification is almost same as the result of the experiment. Falling velocity and drag force increase rapidly in the beginning. And then it goes to the stable state. When the piston head of the control rod assembly is inserted into the damper, the drag force increases instantaneously and the falling velocity decreases quickly. The falling velocity is reduced about 14 % by damper. The total drop time of the control rod assembly is about 1.47s. In the next study, the experiment for the control rod assembly will be carried out, and its result is going to be compared with the CFD analysis result.

  10. 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)

  11. Multi-resolution and multi-scale simulation of the thermal hydraulics in fast neutron reactor assemblies

    International Nuclear Information System (INIS)

    Angeli, P.-E.

    2011-01-01

    The present work is devoted to a multi-scale numerical simulation of an assembly of fast neutron reactor. In spite of the rapid growth of the computer power, the fine complete CFD of a such system remains out of reach in a context of research and development. After the determination of the thermalhydraulic behaviour of the assembly at the macroscopic scale, we propose to carry out a local reconstruction of the fine scale information. The complete approach will require a much lower CPU time than the CFD of the entire structure. The macro-scale description is obtained using either the volume averaging formalism in porous media, or an alternative modeling historically developed for the study of fast neutron reactor assemblies. It provides some information used as constraint of a down-scaling problem, through a penalization technique of the local conservation equations. This problem lean on the periodic nature of the structure by integrating periodic boundary conditions for the required microscale fields or their spatial deviation. After validating the methodologies on some model applications, we undertake to perform them on 'industrial' configurations which demonstrate the viability of this multi-scale approach. (author) [fr

  12. Examination of a failed reactor coolant pump rotating assembly from Crystal River Unit 3

    International Nuclear Information System (INIS)

    Hayner, G.O.; Lubnow, T.; Clary, M.

    1990-01-01

    On January 18, 1989, the A reactor coolant pump rotating assembly at the Crystal River Unit 3 Nuclear Power Plant failed during operation. A rotating assembly from this pump had previously failed in 1986. The reactor coolant pump was fabricated by Byron Jackson Pump Division of Borg-Warner Ind. Products, Inc. from UNS S66286 superalloy (Alloy A286). A root cause failure analysis examination was performed on the pump shaft and other components. The failure analysis included shaft vibrational mode and stress analyses, pump clearance and alignment analyses, and detailed destructive examination of the shaft and hydrostatic bearing assemblies. Based on the detailed physical examination of the shaft it was concluded that cracks initiated in the pump shaft at two sites approximately 180 0 apart in a band of shallow, thermally induced fatigue cracks. The cracks initiated at the bottom edge of the motor end shrink fit pad under the shrink fit sleeve supporting the hydrostatic bearing journal. The band of thermally induced fatigue cracks was apparently caused by mixing of cold seal injection water and hot reactor coolant in gaps between the pump shaft and sleeve. The motor end shrink fit was apparently not effective in preventing introduction of the seal injection water to this area. Initial crack propagation occurred by fatigue due to lateral vibration; however, the majority of crack propagation occurred by abnormal torsional fatigue loading induced by contact and sticking between the rotating and stationary portions of the hydrostatic bearing. Final fracture of the shaft occurred by torsional overload. Metallurgical characteristics and mechanical properties of the shaft were within design specification and probably did not significantly influence the cracking process

  13. Central Reactivity Measurements on Assemblies 1 and 3 of the Fast Reactor FR0

    International Nuclear Information System (INIS)

    Londen, S.O.

    1966-01-01

    The reactivity effects of small samples of various materials have been measured, by the period method at the core centre of Assemblies 1 and 3 of the fast zero power reactor FR0. For some materials the reactivity change as a function of sample size has also been determined experimentally. The core of Assembly 1 consisted only of uranium enriched to 20 % whereas the core of Assembly 3 was diluted with 30 % graphite. The results have been compared with calculated values obtained with a second-order transport-theoretical perturbation model and using differently shielded cross sections depending upon sample size. Qualitative agreement has generally been found, although discrepancies still exist. The spectrum perturbation caused by the experimental arrangement has been analyzed and found to be rather important

  14. Central Reactivity Measurements on Assemblies 1 and 3 of the Fast Reactor FR0

    Energy Technology Data Exchange (ETDEWEB)

    Londen, S O

    1966-01-15

    The reactivity effects of small samples of various materials have been measured, by the period method at the core centre of Assemblies 1 and 3 of the fast zero power reactor FR0. For some materials the reactivity change as a function of sample size has also been determined experimentally. The core of Assembly 1 consisted only of uranium enriched to 20 % whereas the core of Assembly 3 was diluted with 30 % graphite. The results have been compared with calculated values obtained with a second-order transport-theoretical perturbation model and using differently shielded cross sections depending upon sample size. Qualitative agreement has generally been found, although discrepancies still exist. The spectrum perturbation caused by the experimental arrangement has been analyzed and found to be rather important.

  15. Primary circuit and reactor core T-H characteristics determination of WWER 440 reactors

    International Nuclear Information System (INIS)

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

    2010-01-01

    The WWER-440 nuclear fuel vendor permanently improves the assortment of produced nuclear fuel assemblies for achieving better fuel cycle economy and reactor operation safety. During unit refuelling there also could be made some other changes in hydraulic parameters of primary circuit (change of impeller wheels, hydraulic resistance coefficient changes of internal parts of primary circuit, etc.). Therefore it is necessary to determine real coolant flow rate through the reactor during units start-up after their refuelling, and also to have the skilled methodology and computing code for analyzing factors, which affecting the inaccuracy of coolant flow redistribution determination through reactor on flows through separate parts of reactor core in any case of parallel operation of different assembly types. Computing code TH-VCR and CORFLO are used for reactor core characteristics determination for one type of fuel and control assemblies and also in case of parallel operation of different assembly types. The code TH-VCR is able to calculate coolant flow rate for different combinations of three different fuel assembly channel types and three different control assembly channel types. The CORFLO code deals the area of the reactor 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 taken into account. Computing code CORFLO is verified and validated at this time. Paper presents some results from measurements of coolant flow rate through reactors during start-up after unit refuelling and short description of computing code TH-VCR and CORFLO with some calculated results. (Authors)

  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. 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

  18. Numerical verification of the theory of coupled reactors for a deuterium critical assembly using MCNP5

    International Nuclear Information System (INIS)

    Hussein, M.S.; Bonin, H.W.; Lewis, B.J.

    2013-01-01

    The theory of multipoint coupled reactors developed by multi-group transport is verified by using the probabilistic transport code MCNP5. The verification was performed by calculating the multiplication factors (or criticality factors) and coupling coefficients for a two-region test reactor known as Deuterium Critical Assembly, (DCA). The variations of the criticality factors and the coupling coefficients were investigated by changing of the water levels in the inner and outer cores. The numerical results of the model developed with MCNP5 code were validated and verified against published results and the mathematical model based on coupled reactor theory. (author)

  19. Numerical verification of the theory of coupled reactors for a deuterium critical assembly using MCNP5

    Energy Technology Data Exchange (ETDEWEB)

    Hussein, M.S.; Bonin, H.W.; Lewis, B.J., E-mail: mohamed.hussein@rmc.ca, E-mail: bonin-h@rmc.ca, E-mail: lewis-b@rmc.ca [Royal Military College of Canada, Dept. of Chemistry and Chemical Engineering, Kingston, Ontario (Canada)

    2013-07-01

    The theory of multipoint coupled reactors developed by multi-group transport is verified by using the probabilistic transport code MCNP5. The verification was performed by calculating the multiplication factors (or criticality factors) and coupling coefficients for a two-region test reactor known as Deuterium Critical Assembly, (DCA). The variations of the criticality factors and the coupling coefficients were investigated by changing of the water levels in the inner and outer cores. The numerical results of the model developed with MCNP5 code were validated and verified against published results and the mathematical model based on coupled reactor theory. (author)

  20. 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

  1. Tower Shielding Reactor II design and operation report. Vol. 3. Assembling and testing of the control mechanism assembly

    International Nuclear Information System (INIS)

    Ward, D.R.; Holland, L.B.

    1979-09-01

    The mechanisms that are operated to control the reactivity of the Tower Shielding Reactor II(TSR-II) are mounted on a Control Mechanism Housing (CMH) that is centered inside the reactor core. The information required to procure, fabricate, inspect, and assemble a CMH is contained in the ORNL engineering drawings listed in the appropriate sections. The components are fabricated and inspected from these drawings in accordance with a Quality Assurance Plan and a Manufacturing Plan. The material in this report describes the acceptance and performance tests of CMH subassemblies used ty the Tower Shielding Facility (TSF) staff but it can also be used by personnel fabricating the components. This information which was developed and used before the advent of the formalized QA Program and Manufacturing Plans evolved during the fabrication and testing of the first five CMHs

  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. PLGA nanofibers blended with designer self-assembling peptides for peripheral neural regeneration

    Energy Technology Data Exchange (ETDEWEB)

    Nune, Manasa; Krishnan, Uma Maheswari; Sethuraman, Swaminathan, E-mail: swami@sastra.edu

    2016-05-01

    Electrospun nanofibers are attractive candidates for neural regeneration due to similarity to the extracellular matrix. Several synthetic polymers have been used but they lack in providing the essential biorecognition motifs on their surfaces. Self-assembling peptide nanofiber scaffolds (SAPNFs) like RADA16 and recently, designer SAPs with functional motifs RADA16-I-BMHP1 areexamples, which showed successful spinal cord regeneration. But these peptide nanofiber scaffolds have poor mechanical properties and faster degradation rates that limit their use for larger nerve defects. Hence, we have developed a novel hybrid nanofiber scaffold of polymer poly(L-lactide-co-glycolide) (PLGA) and RADA16-I-BMHP1. The scaffolds were characterized for the presence of peptides both qualitatively and quantitatively using several techniques like SEM, EDX, FTIR, CHN analysis, Circular Dichroism analysis, Confocal and thermal analysis. Peptide self-assembly was retained post-electrospinning and formed rod-like nanostructures on PLGA nanofibers. In vitro cell compatibility was studied using rat Schwann cells and their adhesion, proliferation and gene expression levels on the designed scaffolds were evaluated. Our results have revealed the significant effects of the peptide blended scaffolds on promoting Schwann cell adhesion, extension and phenotypic expression. Neural development markers (SEM3F, NRP2 & PLX1) gene expression levels were significantly upregulated in peptide blended scaffolds compared to the PLGA scaffolds. Thus the hybrid blended novel designer scaffolds seem to be promising candidates for successful and functional regeneration of the peripheral nerve. - Highlights: • A novel blended scaffold of polymer PLGA and designer self-assembling peptide RADA16-I-BMPH1 was designed • The peptide retained the self-assembling features and formed rod like nanostructures on top of PLGA nanofibers • PLGA-peptide scaffolds have promoted the Schwann cell bipolar extension and

  4. FBR type reactor

    International Nuclear Information System (INIS)

    Inoue, Kotaro; Kawashima, Katsuyuki; Zuketen, Atsushi.

    1982-01-01

    Purpose: To flatten the power distribution of a reactor core and shorten the breeding time. Constitution: The reactor core comprises a core region having fission products, an outer blanket region surrounding the outer side of the core region and having fertile material and an inner blanket region disposed within the core region and having fertile material. The axial thickness of the inner blanket region is made greater at the central portion and smaller at the peripheral portion of the inner blanket region, and the outermost peripheral end at the peripheral portion of the inner blanket region is opposed by way of the core region to the outer blanket region. In such an arrangement, the power decrease in the peripheral portion of the core region can be suppressed to thereby flatten the power distribution in the reactor core and shorten the breeding time. (Moriyama, K.)

  5. Basic experiments of reactor physics using the critical assembly TCA

    International Nuclear Information System (INIS)

    Obara, Toru; Igashira, Masayuki; Sekimoto, Hiroshi; Nakajima, Ken; Suzaki, Takenori.

    1994-02-01

    This report is based on lectures given to graduate students of Tokyo Institute of Technology. It covers educational experiments conducted with the Tank-Type Critical Assembly (TCA) at Japan Atomic Energy Research Institute in July, 1993. During this period, the following basic experiments on reactor physics were performed: (1) Critical approach experiment, (2) Measurement of neutron flux distribution, (3) Measurement of power distribution, (4) Measurement of fuel rod worth distribution, (5) Measurement of safety sheet worth by the rod drop method. The principle of experiments, experimental procedure, and analysis of results are described in this report. (author)

  6. Advanced gray rod control assembly

    Science.gov (United States)

    Drudy, Keith J; Carlson, William R; Conner, Michael E; Goldenfield, Mark; Hone, Michael J; Long, Jr., Carroll J; Parkinson, Jerod; Pomirleanu, Radu O

    2013-09-17

    An advanced gray rod control assembly (GRCA) for a nuclear reactor. The GRCA provides controlled insertion of gray rod assemblies into the reactor, thereby controlling the rate of power produced by the reactor and providing reactivity control at full power. Each gray rod assembly includes an elongated tubular member, a primary neutron-absorber disposed within the tubular member said neutron-absorber comprising an absorber material, preferably tungsten, having a 2200 m/s neutron absorption microscopic capture cross-section of from 10 to 30 barns. An internal support tube can be positioned between the primary absorber and the tubular member as a secondary absorber to enhance neutron absorption, absorber depletion, assembly weight, and assembly heat transfer characteristics.

  7. Educational reactor-physics experiments with the critical assemble TCA

    Energy Technology Data Exchange (ETDEWEB)

    Tsutsui, Hiroaki; Okubo, Masaaki; Igashira, Masayuki [Tokyo Inst. of Tech. (Japan); Horiki, Oichiro; Suzaki, Takenori

    1997-10-01

    The Tank-Type Critical Assembly (TCA) of Japan Atomic Energy Research Institute is research equipment for light water reactor physics. In the present report, the lectures given to the graduate students of Tokyo Institute of Technology who participated in the educational experiment course held on 26-30 August at TCA are rearranged to provide useful information for those who will implement educational basic experiments with TCA in the future. This report describes the principles, procedures, and data analyses for (1) Critical approach and Exponential experiment, (2) Measurement of neutron flux distribution, (3) Measurement of power distribution, (4) Measurement of fuel rod worth distribution, and (5) Measurement of safety plate worth by the rod drop method. (author)

  8. Educational reactor-physics experiments with the critical assembly TCA

    International Nuclear Information System (INIS)

    Tsutsui, Hiroaki; Okubo, Masaaki; Igashira, Masayuki; Horiki, Oichiro; Suzaki, Takenori.

    1997-10-01

    The Tank-Type Critical Assembly (TCA) of Japan Atomic Energy Research Institute is research equipment for light water reactor physics. In the present report, the lectures given to the graduate students of Tokyo Institute of Technology who participated in the educational experiment course held on 26-30 August at TCA are rearranged to provide useful information for those who will implement educational basic experiments with TCA in the future. This report describes the principles, procedures, and data analyses for 1) Critical approach and Exponential experiment, 2) Measurement of neutron flux distribution, 3) Measurement of power distribution, 4) Measurement of fuel rod worth distribution, and 5) Measurement of safety plate worth by the rod drop method. (author)

  9. 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

  10. 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)

  11. Permanent seal ring for a nuclear reactor cavity

    International Nuclear Information System (INIS)

    Hankinson, M.F.; Marshall, J.R.

    1988-01-01

    A nuclear reactor containment arrangement is described including: a. a reactor vessel which thermally expands and contracts during cyclic operation of the reactor and which has a peripheral wall; b. a containment wall spaced apart from and surrounding the peripheral wall of the reactor vessel and defining an annular thermal expansion gap therebetween for accommodating thermal expansion; and c. an annular ring seal which sealingly engages and is affixed to and extends between the peripheral wall of the reactor vessel and the containment wall

  12. 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.)

  13. 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.)

  14. 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

  15. Full scale seismic simulation of a nuclear reactor with parallel finite element analysis code for assembled structure

    International Nuclear Information System (INIS)

    Yamada, Tomonori

    2010-01-01

    The safety requirement of nuclear power plant attracts much attention nowadays. With the growing computing power, numerical simulation is one of key technologies to meet this safety requirement. Center for Computational Science and e-Systems of Japan Atomic Energy Agency has been developing a finite element analysis code for assembled structure to accurately evaluate the structural integrity of nuclear power plant in its entirety under seismic events. Because nuclear power plant is very huge assembled structure with tens of millions of mechanical components, the finite element model of each component is assembled into one structure and non-conforming meshes of mechanical components are bonded together inside the code. The main technique to bond these mechanical components is triple sparse matrix multiplication with multiple point constrains and global stiffness matrix. In our code, this procedure is conducted in a component by component manner, so that the working memory size and computing time for this multiplication are available on the current computing environment. As an illustrative example, seismic simulation of a real nuclear reactor of High Temperature engineering Test Reactor, which is located at the O-arai research and development center of JAEA, with 80 major mechanical components was conducted. Consequently, our code successfully simulated detailed elasto-plastic deformation of nuclear reactor and its computational performance was investigated. (author)

  16. 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.

  17. 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.)

  18. 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.)

  19. Reactor performances and microbial communities of biogas reactors: effects of inoculum sources.

    Science.gov (United States)

    Han, Sheng; Liu, Yafeng; Zhang, Shicheng; Luo, Gang

    2016-01-01

    Anaerobic digestion is a very complex process that is mediated by various microorganisms, and the understanding of the microbial community assembly and its corresponding function is critical in order to better control the anaerobic process. The present study investigated the effect of different inocula on the microbial community assembly in biogas reactors treating cellulose with various inocula, and three parallel biogas reactors with the same inoculum were also operated in order to reveal the reproducibility of both microbial communities and functions of the biogas reactors. The results showed that the biogas production, volatile fatty acid (VFA) concentrations, and pH were different for the biogas reactors with different inocula, and different steady-state microbial community patterns were also obtained in different biogas reactors as reflected by Bray-Curtis similarity matrices and taxonomic classification. It indicated that inoculum played an important role in shaping the microbial communities of biogas reactor in the present study, and the microbial community assembly in biogas reactor did not follow the niche-based ecology theory. Furthermore, it was found that the microbial communities and reactor performances of parallel biogas reactors with the same inoculum were different, which could be explained by the neutral-based ecology theory and stochastic factors should played important roles in the microbial community assembly in the biogas reactors. The Bray-Curtis similarity matrices analysis suggested that inoculum affected more on the microbial community assembly compared to stochastic factors, since the samples with different inocula had lower similarity (10-20 %) compared to the samples from the parallel biogas reactors (30 %).

  20. 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

  1. Design of an integral missile shield in integrated head assembly for pressurized water reactor at commercial nuclear plants

    International Nuclear Information System (INIS)

    Baliga, Ravi; Watts, Tom Neal; Kamath, Harish

    2015-01-01

    In ICONE22, the authors presented the Integrated Head Assembly (IHA) design concept implemented at Callaway Nuclear Power Plant in Missouri, USA. The IHA concept is implemented to reduce the outage duration and the associated radiation exposure to the workers by reducing critical path time during Plant Refueling Outage. One of the head area components in the IHA is a steel missile shield designed to protect the Control Rod Drive Mechanism (CRDM) assembly from damaging other safety-related components in the vicinity in the Containment. Per Federally implemented General Design Criteria for commercial nuclear plants in the USA, the design of Nuclear Steam Supply System (NSSS) must provide protection from the damages caused by a postulated event of CRDM housing units and their associated parts disengaging from the reactor vessel assembly. This event is considered as a Loss of Coolant Accident (LOCA) and assumes that once the CRDM housing unit and their associated parts disengage from the reactor vessel internals assembly, they travel upward by the water jet with the following sequence of events: Per Reference 1, the drive shaft and control rod cluster are forced out of the reactor core by the differential pressure across the drive shaft with the assumption that the drive shaft and control rod cluster, latched together, are fully inserted when the accident occurs. After the travel, the rod cluster control spider will impact the lower side of the upper support plate inside the reactor vessel fracturing the flexure arms in the joint freeing the drive shaft from the control rod cluster. The control rod cluster is stopped by the upper support plate and will remain below the upper support plate during this accident. However, the drive shaft will continue to accelerate in the upward direction until it is stopped by a safety feature in the IHA. The integral missile shield as a safety feature in the IHA is designed to stop the CRDM drive shaft from moving further up in the

  2. 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)

  3. Shock buffer for nuclear control assembly

    International Nuclear Information System (INIS)

    Bevilacqua, F.

    1977-01-01

    A shock buffer is provided for the gradual deceleration of a rapidly descending control element assembly in a nuclear reactor. The interactive buffer components are associated respectively with the movable control element assembly and part of the upper guide structure independent of and spaced from the fuel assemblies of the reactor

  4. Reactor laboratory course for Korean under-graduate students in Kyoto University Critical Assembly (KUGSiKUCA)

    International Nuclear Information System (INIS)

    Pyeon, Cheol Ho; Misawa, Tsuyoshi; Unesaki, Hironobu; Ichihara, Chihiro; Shiroya, Seiji; Whang, Joo Ho; Kim, Myung Hyun

    2005-01-01

    The Reactor Laboratory Course for Korean Under-Graduate Students has been carried out at Kyoto University Critical Assembly of Japan. This course has been launched from fiscal year 2003 and has been founded by Ministry of Science and Technology of Korean Government. Since then, the total number of 43 Korean under-graduate students, who have majored in nuclear engineering of 6 universities in all over the Korea, has been taken part in this course. The reactor physics experiments have been performed in this course, such as Approach to criticality, Control rod calibration, Measurement of neutron flux and power calibration, and Educational reactor operation. As technical tour of Japan, nuclear site tour has been taken during their stay in Japan, such as PWR, FBR, nuclear fuel company and some institutes

  5. 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.

  6. 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...

  7. Assembly tool design

    International Nuclear Information System (INIS)

    Kanamori, Naokazu; Nakahira, Masataka; Ohkawa, Yoshinao; Tada, Eisuke; Seki, Masahiro

    1996-06-01

    The reactor core of the International Thermonuclear Experimental Reactor (ITER) is assembled with a number of large and asymmetric components within a tight tolerance in order to assure the structural integrity for various loads and to provide the tritium confinement. In addition, the assembly procedure should be compatible with remote operation since the core structures will be activated by 14-MeV neutrons once it starts operation and thus personal access will be prohibited. Accordingly, the assembly procedure and tool design are quite essential and should be designed from the beginning to facilitate remote operation. According to the ITER Design Task Agreement, the Japan Atomic Energy Research Institute (JAERI) has performed design study to develop the assembly procedures and associated tool design for the ITER tokamak assembly. This report describes outlines of the assembly tools and the remaining issues obtained in this design study. (author)

  8. 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)

  9. 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. )

  10. 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

  11. 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.)

  12. Peripheral myelin protein 22 alters membrane architecture

    Science.gov (United States)

    Mittendorf, Kathleen F.; Marinko, Justin T.; Hampton, Cheri M.; Ke, Zunlong; Hadziselimovic, Arina; Schlebach, Jonathan P.; Law, Cheryl L.; Li, Jun; Wright, Elizabeth R.; Sanders, Charles R.; Ohi, Melanie D.

    2017-01-01

    Peripheral myelin protein 22 (PMP22) is highly expressed in myelinating Schwann cells of the peripheral nervous system. PMP22 genetic alterations cause the most common forms of Charcot-Marie-Tooth disease (CMTD), which is characterized by severe dysmyelination in the peripheral nerves. However, the functions of PMP22 in Schwann cell membranes remain unclear. We demonstrate that reconstitution of purified PMP22 into lipid vesicles results in the formation of compressed and cylindrically wrapped protein-lipid vesicles that share common organizational traits with compact myelin of peripheral nerves in vivo. The formation of these myelin-like assemblies depends on the lipid-to-PMP22 ratio, as well as on the PMP22 extracellular loops. Formation of the myelin-like assemblies is disrupted by a CMTD-causing mutation. This study provides both a biochemical assay for PMP22 function and evidence that PMP22 directly contributes to membrane organization in compact myelin. PMID:28695207

  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. 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)

  15. Zero energy reactor 'RB'

    Energy Technology Data Exchange (ETDEWEB)

    Popovic, D; Takac, S; Markovic, H; Raisic, N; Zdravkovic, Z; Radanovic, Lj [Boris Kidric Institute of Nuclear Sciences, Vinca, Beograd (Yugoslavia)

    1959-03-15

    In 1958 the zero energy reactor RB was built with the purpose of enabling critical experiments with various reactor systems to be carried out. The first core assembly built in this reactor consists of heavy water as moderator and natural uranium metal as fuel. In order to be able to obtain very accurate results when measuring the main characteristics of the assembly the reactor was built as a completely bare system. (author)

  16. Contribution at the turbulence study in tridimensional flow: application at nuclear reactors

    International Nuclear Information System (INIS)

    Rodet, J.C.

    1985-01-01

    Inside hexagonal assemblies of 3 or 19 pins of nuclear reactor type Super Phenix parietal pressure ranges, mean kinematical ranges and turbulent ranges has been measured. For these latest a measurement methodology taking into account the orientation constraints binded at the geometry of assembly is developed for a simple probe having two crossed hot wires. The mean kinematic range obtained is tridimensional; it puts in clearness periodicities, a peripheral flow along the case and render an account of kink values. Local analysis of Reynold tensors permits to return localy and in selected references at tangential bidimensional situations. Evaluations of mixing lenghts are executed in subchannels, then integrated in a numerical code of thermo hydraulic prediction with object to improve it [fr

  17. Modular reactor head shielding system

    International Nuclear Information System (INIS)

    Jacobson, E. B.

    1985-01-01

    An improved modular reactor head shielding system is provided that includes a frame which is removably assembled on a reactor head such that no structural or mechanical alteration of the head is required. The shielding system also includes hanging assemblies to mount flexible shielding pads on trolleys which can be moved along the frame. The assemblies allow individual pivoting movement of the pads. The pivoting movement along with the movement allowed by the trolleys provides ease of access to any point on the reactor head. The assemblies also facilitate safe and efficient mounting of the pads directly to and from storage containers such that workers have additional shielding throughout virtually the entire installation and removal process. The flexible shielding pads are designed to interleave with one another when assembled around the reactor head for substantially improved containment of radiation leakage

  18. 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)

  19. Reactor physics studies in the steam flooded GCFR-Phase II critical assembly

    International Nuclear Information System (INIS)

    Bhattacharyya, S.K.

    1978-08-01

    A possible accident scenario in a Gas-Cooled Fast Reactor (GCFR) is the leakage of secondary steam into the core. Considerable analytical effort has gone into the study of the effects of such an accidental steam entry. The work described represents the first full scale experimental study of the steam-entry phenomenon in GCFRs. The reference GCFR model used for the study was the benchmark GCFR Phase II assembly, and polyethylene foam was used to provide a very homogeneous steam simulation. The reactivity worth of steam entry was measured for three different steam densities. In addition, a set of integral physics parameters were measured in the largest steam density (0.008 g/cm 3 ) configuration. The corresponding parameters were also measured in dry reference GCFR critical assembly for comparison. The experiments were analyzed using ENDF/B-IV data and two-dimensional diffusion theory methods. As in earlier GCFR critical experiments analysis, the Benoist method was used to treat the problem of neutron streaming

  20. Numerical verification/validation of the theory of coupled reactors for deuterium critical assembly, using MCNP5 and Serpent codes

    International Nuclear Information System (INIS)

    Hussein, M.S; Lewis, B.J.; Bonin, H.W.

    2013-01-01

    The theory of multipoint coupled reactors developed by multi-group transport is verified by using the probabilistic transport code MCNP5 and the continuous-energy Monte Carlo reactor physics burnup calculation Serpent code. The verification was performed by calculating the multiplication factors (or criticality factors) and coupling coefficients for a two-region test reactor known as the Deuterium Critical Assembly, DCA. The multiplication factors k eff calculated numerically and independently from simulations of the DCA by MCNP5 and Serpent codes are compared with the multiplication factors k eff calculated based on the coupled reactor theory. Excellent agreement was obtained between the multiplication factors k eff calculated with the Serpent code, with MCNP5, and from the coupled reactor theory. This analysis demonstrates that the Serpent code is valid for the multipoint coupled reactor calculations. (author)

  1. Numerical verification/validation of the theory of coupled reactors for deuterium critical assembly, using MCNP5 and Serpent codes

    Energy Technology Data Exchange (ETDEWEB)

    Hussein, M.S, E-mail: mohamed.hussein@rmc.ca [Royal Military College of Canada, Dept. of Chemistry and Chemical Engineering, Kingston, Ontario (Canada); Lewis, B.J., E-mail: Brent.Lewis@uoit.ca [Univ. of Ontario Inst. of Technology, Faculty of Energy Systems and Nuclear Science, Oshawa, Ontario (Canada); Bonin, H.W., E-mail: bonin-h@rmc.ca [Royal Military College of Canada, Dept. of Chemistry and Chemical Engineering, Kingston, Ontario (Canada)

    2013-07-01

    The theory of multipoint coupled reactors developed by multi-group transport is verified by using the probabilistic transport code MCNP5 and the continuous-energy Monte Carlo reactor physics burnup calculation Serpent code. The verification was performed by calculating the multiplication factors (or criticality factors) and coupling coefficients for a two-region test reactor known as the Deuterium Critical Assembly, DCA. The multiplication factors k{sub eff} calculated numerically and independently from simulations of the DCA by MCNP5 and Serpent codes are compared with the multiplication factors k{sub eff} calculated based on the coupled reactor theory. Excellent agreement was obtained between the multiplication factors k{sub eff} calculated with the Serpent code, with MCNP5, and from the coupled reactor theory. This analysis demonstrates that the Serpent code is valid for the multipoint coupled reactor calculations. (author)

  2. Design of a mixed recharge with MOX assemblies of greater relation of moderation for a BWR reactor; Diseno de una recarga mixta con ensambles MOX de mayor relacion de moderacion para un reactor BWR

    Energy Technology Data Exchange (ETDEWEB)

    Ramirez S, J.R.; Alonso V, G.; Palacios H, J. [ININ, Carretera Mexico-Toluca Km. 36.5, 52045 Estado de Mexico (Mexico)]. e-mail: jrrs@nuclear.inin.mx

    2004-07-01

    The study of the fuel of mixed oxides of uranium and plutonium (MOX) it has been topic of investigation in many countries of the world and those are even discussed in many places the benefits of reprocessing the spent fuel to extract the plutonium created during the irradiation of the fuel in the nuclear power reactors. At the moment those reactors that have been loaded partially with MOX fuel, are mainly of the type PWR where a mature technology has been achieved in some countries like they are France, Belgium and England, however the experience with reactors of the type BWR is more limited and it is continued studying the best way to introduce this type of fuel in BWRs, one of the main problems to introduce MOX in reactors BWR is the neutronic design of the same one, existing different concepts to introduce the plutonium in the assemblies of fuel and one of them is the one of increasing the relationship of moderation of the assemble. In this work a MOX fuel assemble design is presented and the obtained results so far in the ININ. These results indicate that the investigated concept has some exploitable advantages in the use of the MOX fuel. (Author)

  3. SP-100 nuclear assembly test: Test assembly functional requirements and system arrangement

    International Nuclear Information System (INIS)

    Fallas, T.T.; Gluck, R.; Motwani, K.; Clay, H.; O'Neill, G.

    1991-01-01

    This paper describes the functional requirements and the system that will be tested to validate the reactor, flight shield, and flight controller of the SP-100 Generic Flight System (GFS). The Nuclear Assembly Test (NAT) consists of the test article (SP-100 reactor with control devices and the flight shield) and its supporting systems. The NAT test assembly is being designed by GE. Westinghouse Hanford Company (WHC) is designing the test cell and vacuum vessel system that will contain the NAT test assembly (Renkey et al. 1989). Preliminary design reviews have been completed and the final design is under way

  4. 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)

  5. Development of SiC Neutron Detector Assembly to Measure the Neutron Flux of the Reactor Core

    Energy Technology Data Exchange (ETDEWEB)

    Park, Se Hwan; Park, June Sic; Shin, Hee Sung; Kim, Ho Dong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Yong Kyun [Hanyang University, Seoul (Korea, Republic of)

    2012-05-15

    At present, the conventional detector to measure the neutron at harsh environment is a Self Powered Neutron Detector (SPND). Rhodium(Rh)-103 is in the SPND. When neutron is incident on the Rhodium, the neutron capture reaction occurs, and the Rh-103 is converted to Rh-104. The Rh-104 is decayed to Pd-104 by {beta}-decay, and electrons are generated as the decay products. Because of the half life of Rh-104, approximately 5 minutes are required for the SPND output to reach the equilibrium condition. Therefore the on-line monitoring of the nuclear reactor state is limited if the neutron flux in the reactor core is monitored with the SPND. Silicon carbide (SiC) has the possibility to be developed as neutron detector at harsh environment, because the SiC can be operative at high temperature and high neutron flux conditions. Previously, the basic operation properties of the SiC detector were studied. Also, the radiation response of the SiC detector was studied at high neutron and gamma dose rate. The measurement results for an ex-core neutron flux monitor or a neutron flux monitor of the spent fuel were published. The SiC detector was also developed as neutron detector to measure the fissile material with active interrogation method. However, the studies about the development of SiC detector are still limited. In the present work, the radiation damage effect of the SiC detector was studied. The detector structure was determined based on the study, and a neutron detector assembly was made with the SiC detectors. The neutron and gamma-ray response of the detector assembly is presented in this paper. The detector assembly was positioned in the HANARO research reactor core, the performance test was done. The preliminary results are also included in this paper

  6. 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

  7. 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

  8. 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

  9. Improvements in or relating to nuclear reactors

    International Nuclear Information System (INIS)

    Timofeev, A.V.; Batjukov, V.I.; Fadeev, A.I.; Shapkin, A.F.; Shikhiyan, T.G.; Ordynsky, G.V.; Drachev, V.P.; Pogodin, E.N.

    1980-01-01

    A refuelling installation for nuclear reactor complexes is described for recharging the reactor vessels of such complexes with new fuel assemblies and for removing spent fuel assemblies from the reactor vessel. (U.K.)

  10. File list: Unc.Bld.20.AllAg.Peripheral_blood [Chip-atlas[Archive

    Lifescience Database Archive (English)

    Full Text Available Unc.Bld.20.AllAg.Peripheral_blood hg19 Unclassified Blood Peripheral blood SRX10800...66 http://dbarchive.biosciencedbc.jp/kyushu-u/hg19/assembled/Unc.Bld.20.AllAg.Peripheral_blood.bed ...

  11. File list: Unc.Bld.50.AllAg.Peripheral_blood [Chip-atlas[Archive

    Lifescience Database Archive (English)

    Full Text Available Unc.Bld.50.AllAg.Peripheral_blood hg19 Unclassified Blood Peripheral blood SRX10800...66 http://dbarchive.biosciencedbc.jp/kyushu-u/hg19/assembled/Unc.Bld.50.AllAg.Peripheral_blood.bed ...

  12. 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

  13. 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)

  14. Design configuration of GCFR core assemblies

    International Nuclear Information System (INIS)

    LaBar, M.P.; Lee, G.E.; Meyer, R.J.

    1980-05-01

    The current design configurations of the core assemblies for the gas-cooled fast reactor (GCFR) demonstration plant reactor core conceptual design are described. Primary emphasis is placed upon the design innovations that have been incorporated in the design of the core assemblies since the establishment of the initial design of an upflow GCFR core. A major feature of the design configurations is that they are prototypical of core assemblies for use in commercial plants; a larger number of the same assemblies would be used in a commercial plant

  15. 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)

  16. Nuclear reactors built, being built, or planned, 1991

    Energy Technology Data Exchange (ETDEWEB)

    Simpson, B.

    1992-07-01

    This document contains unclassified information about facilities built, being built, or planned in the United States for domestic use or export as of December 31, 1991. The book is divided into three major sections: Section 1 consists of a reactor locator map and reactor tables; Section 2 includes nuclear reactors that are operating, being built, or planned; and Section 3 includes reactors that have been shut down permanently or dismantled. Sections 2 and 3 contain the following classification of reactors: Civilian, Production, Military, Export, and Critical Assembly. Export reactor refers to a reactor for which the principal nuclear contractor is an American company -- working either independently or in cooperation with a foreign company (Part 4, in each section). Critical assembly refers to an assembly of fuel and assembly of fuel and moderator that requires an external source of neutrons to initiate and maintain fission. A critical assembly is used for experimental measurements (Part 5).

  17. Nuclear reactors built, being built, or planned, 1991

    International Nuclear Information System (INIS)

    Simpson, B.

    1992-07-01

    This document contains unclassified information about facilities built, being built, or planned in the United States for domestic use or export as of December 31, 1991. The book is divided into three major sections: Section 1 consists of a reactor locator map and reactor tables; Section 2 includes nuclear reactors that are operating, being built, or planned; and Section 3 includes reactors that have been shut down permanently or dismantled. Sections 2 and 3 contain the following classification of reactors: Civilian, Production, Military, Export, and Critical Assembly. Export reactor refers to a reactor for which the principal nuclear contractor is an American company -- working either independently or in cooperation with a foreign company (Part 4, in each section). Critical assembly refers to an assembly of fuel and assembly of fuel and moderator that requires an external source of neutrons to initiate and maintain fission. A critical assembly is used for experimental measurements (Part 5)

  18. File list: Unc.Bld.10.AllAg.Peripheral_blood [Chip-atlas[Archive

    Lifescience Database Archive (English)

    Full Text Available Unc.Bld.10.AllAg.Peripheral_blood hg19 Unclassified Blood Peripheral blood SRX10800...66,SRX1080067 http://dbarchive.biosciencedbc.jp/kyushu-u/hg19/assembled/Unc.Bld.10.AllAg.Peripheral_blood.bed ...

  19. File list: Unc.Bld.05.AllAg.Peripheral_blood [Chip-atlas[Archive

    Lifescience Database Archive (English)

    Full Text Available Unc.Bld.05.AllAg.Peripheral_blood hg19 Unclassified Blood Peripheral blood SRX10800...66,SRX1080067 http://dbarchive.biosciencedbc.jp/kyushu-u/hg19/assembled/Unc.Bld.05.AllAg.Peripheral_blood.bed ...

  20. 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

  1. Fertile assembly for a fast neutron nuclear reactor cooled by liquid sodium, with regulation of the cooling rate

    International Nuclear Information System (INIS)

    Pradal, L.; Berte, M.; Chiarelli, C.

    1985-01-01

    The assembly has a casing in which are arranged the fertile elements, the liquid sodium flowing through the casing along these elements. It includes several apertured diaphragms transverse to the rods to regulate the liquid sodium flow rate. At least one diaphragm, in its central part around its aperture, of a material soluble in liquid sodium, such as copper. The invention applies, more particularly, to fast neutron nuclear reactor having a heterogeneous core. The coolant flow can increase with time to match the increased power generated by the fertile assembly along its life [fr

  2. Reactor container

    International Nuclear Information System (INIS)

    Oyamada, Osamu; Furukawa, Hideyasu; Uozumi, Hiroto.

    1979-01-01

    Purpose: To lower the position of an intermediate slab within a reactor container and fitting a heat insulating material to the inner wall of said intermediate slab, whereby a space for a control rod exchanging device and thermal stresses of the inner peripheral wall are lowered. Constitution: In the pedestal at the lower part of a reactor pressure vessel there is formed an intermediate slab at a position lower than diaphragm floor slab of the outer periphery of the pedestal thereby to secure a space for providing automatic exchanging device of a control rod driving device. Futhermore, a heat insulating material is fitted to the inner peripheral wall at the upper side of the intermediate slab part, and the temperature gradient in the wall thickness direction at the time of a piping rupture trouble is made gentle, and thermal stresses at the inner peripheral wall are lowered. (Sekiya, K.)

  3. 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)

  4. 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

  5. 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.)

  6. 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)

  7. Modeling the spatial distribution of the parameters of the coolant in the reactor volume

    International Nuclear Information System (INIS)

    Nikonov, S.P.

    2011-01-01

    In this paper the approach to the question about the spatial distribution of the parameters of the coolant in-reactor volume. To describe the in-core space is used specially developed preprocessor. When the work of the preprocessor in the first place, is recreated on the basis of available information (mostly-the original drawings) with high accuracy three-dimensional description of the structures of the reactor volume and, secondly, are prepared on this basis blocks input to the nodal system code improved estimate ATHLET, allows to take into account the hydrodynamic interaction between the spatial control volumes. As an example the special case of solutions of international standard problem on the reconstruction of the transition process in the third unit of the Kalinin nuclear power plant, due to the shutdown of one of the four Main Coolant Pumps in operation at the rated capacity (first download). Model-core area consists of approximately 58 000 control volumes and spatial relationships. It shows the influence of certain structural units of the core to the distribution of the mass floe rate of its height. It is detected a strong cross-flow coolant in the area over the baffle. Moreover, we study the distribution of the coolant temperature at the assembly head of WWER-1000 reactor. It is shown that in the region of the top of the assembly head, where we have installation of thermocouples, the flow coolant for internal assemblies core is formed by only from guide channel Reactor control and protected system Control rod flow, or a mixture of the guide channel flow and flow from the area in front of top grid head assembly (the peripheral assemblies). It is shown that the magnitude of the flow guide channels affects not only the position of control rods, but also the presence of a particular type of measuring channels (Self powered neutron detector sensors or Temperature control sensors) in the cassette. (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. 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

  10. Initial charge reactor core

    International Nuclear Information System (INIS)

    Kiyono, Takeshi

    1984-01-01

    Purpose: To effectivity burn fuels and improve the economical performance in an inital charge reactor core of BWR type reactors or the likes. Constitution: In a reactor core constituted with a plurality of fuel assemblies which are to be partially replaced upon fuel replacement, the density of the fissionable materials and the moderator - fuel ratio of a fuel assembly is set corresponding to the period till that fuel assembly is replaced, in which the density of the nuclear fissionable materials is lowered and the moderator - fuel ratio is increased for the fuel assembly with a shorter period from the fueling to the fuel exchange and, while on the other hand, the density of the fissionable materials is increased and the moderator - fuel ratio is decreased for the fuel assembly with a longer period from the fueling to the replacement. Accordingly, since the moderator - fuel ratio is increased for the fuel assembly to be replaced in a shorter period, the neutrons moderating effect is increased to increase the reactivity. (Horiuchi, T.)

  11. 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

  12. 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.)

  13. 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

  14. 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)

  15. 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.)

  16. Experimental Studies on Assemblies 1 and 2 of the Fast Reactor FR-0. Part 1

    Energy Technology Data Exchange (ETDEWEB)

    Andersson, T L; Hellstrand, E; Londen, S O; Tiren, L I

    1965-08-15

    FR0 is a fast zero power reactor built for experiments in reactor physics. It is a split table machine containing vertical fuel elements. 120 kg of U{sup 235} are available as fuel, which is fabricated into metallic plates of 20 % enrichment. The control system comprises 5 spring-loaded safety elements and 3 + 1 elements for startup operations and power control. The reactor went critical in February 1964. The first assemblies studied were made up of undiluted fuel into a cylindrical and a spherical core, respectively, surrounded by a reflector made of copper. The present report describes some experiments made on these systems. Primarily, critical mass determinations, flux distribution measurements and studies of the conversion ratio are dealt with. The measured quantities have been compared with theoretical predictions using various transport theory programmes (DSN, TDC) and cross section sets. The experimental results show that the neutron spectrum in the copper reflector is softer than predicted, but apart from this discrepancy agreement with theory has generally been obtained.

  17. 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

  18. Nuclear reactor spring strip grid spacer

    International Nuclear Information System (INIS)

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

    1980-01-01

    An improved and novel grid spacer was developed for use in nuclear reactor fuel assemblies. It is comprised of a series of intersecting support strips and a peripheral support band attached to the ends of the support strips. Each of the openings into which the fuel element is inserted has a number of protruding dimples and springs extending in different directions. The dimples coact with the springs to secure the fuel rods in the openings. Compared with previous designs, this design gives more positive alignment of the support stips while allowing greater flexibility to counterbalance the effects of thermal expansion. The springs are arranged in alternating directions so that the reaction forces tend to counterbalance each other, which in turn minimizes the reaction loads on the supporting structure. (D.N.)

  19. 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.)

  20. 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.)

  1. 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)

  2. 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)

  3. Reactor laboratory course for students majoring in nuclear engineering with the Kyoto University Critical Assembly (KUCA)

    International Nuclear Information System (INIS)

    Nishihara, H.; Shiroya, S.; Kanda, K.

    1996-01-01

    With the use of the Kyoto University Critical Assembly (KUCA), a joint reactor laboratory course of graduate level is offered every summer since 1975 by nine associated Japanese universities (Hokkaido University, Tohoku University, Tokyo Institute of Technology, Musashi Institute of Technology, Tokai University, Nagoya University, Osaka University, Kobe University of Mercantile Marine and Kyushu University) in addition to a reactor laboratory course of undergraduate level for Kyoto University. These courses are opened for three weeks (two weeks for the joint course and one week for the undergraduate course) to students majoring in nuclear engineering and a total of 1,360 students have taken the course in the last 21 years. The joint course has been institutionalized with the background that it is extremely difficult for a single university in Japan to have her own research or training reactor. By their effort, the united faculty team of the joint course have succeeded in giving an effective, unique one-week course, taking advantage of their collaboration. Last year, an enquete (questionnaire survey) was conducted to survey the needs for the educational experiments of graduate level and precious data have been obtained for promoting reactor laboratory courses. (author)

  4. Methods of assembling and disassembling spider and burnable poison rod structures for nuclear reactors

    International Nuclear Information System (INIS)

    Edwards, G.T.; Schluderberg, D.C.

    1981-01-01

    A technique is provided for engaging and disengaging burnable poison rods from a spider in a nuclear reactor fuel assembly. A cap on the end of each of the burnable poison rods is provided with a shank or stem that is received in a respective bore formed in the spider. A frangible flange secures the shank and rod to the spider. Pressing the shank in the direction of the bore axis by means, e.g., of a plate ruptures the frangible flange to release the rod from the spider. (author)

  5. Variable mounting assembly for transducers employed in nuclear reactor vessel inspection apparatus

    International Nuclear Information System (INIS)

    Elsner, H.J.; Antol, R.F.; Castner, R.P.

    1979-01-01

    A positionally variable mounting assembly for transducers used to interrogate a nuclear reactor vessel is disclosed. Means are provided for clamping each transducer of an array about its flange in a central restraining block. The central restraining block is, in turn, pivotally mounted in a yoke. The yoke is movable secured to bars or rails bolted to the transducer plate and, by loosening appropriate bolts, can be moved along the ways or pivoted about one of them. Further, the restraining block can be removed from the yoke and pivotally clamped in a different orientation to upstanding brackets attached to the transducer array plate, or rotated through 90 0 and then secured again in the yoke

  6. 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.

  7. 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

  8. 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)

  9. PARR-2: reactor description and experiments

    International Nuclear Information System (INIS)

    Wyne, M.F.; Meghji, J.H.

    1990-12-01

    PARR-2 is a miniature neutron source reactor (MNSR) research reactor has been designed at the rate of 27 kW. Reactor assembly comprises of peaking characteristics with a self limiting flux. In this report reactor description with its assembly and instrumentation control system has been explained. The reactor engineering and physics experiments which can be performed on this reactor are explained in this report. PARR-2 is fueled with HEU fuel pins which are about 90% enriched in U-235. Specific requirements for the safety of the reactor, its building and the personnel, normal instrumentation as required in an industrial environment is sufficient. (A.B.)

  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. Stress analysis for CANDU reactor structure assembly following a postulated p/t, c/t rupture after flow blockage

    Energy Technology Data Exchange (ETDEWEB)

    Soliman, S A; Lee, T; Ibrahim, A M; Hodgson, S [Atomic Energy of Canada Ltd., Saskatoon, SK (Canada)

    1996-12-31

    This paper describes the collapse load calculations for the reactor structure assembly under the postulated fuel channel flow blockage Level D (faulted) loading condition. Under the flow blockage condition, the primary coolant flow path is obstructed between the inlet and outlet feeder connections to the headers. This, in turn, is postulated to cause the pressure tube and calandria tube to rupture and release hot molten fuel into the moderator, producing a hydrodynamic transient within the calandria shell. The most severe hydrodynamic loads occur within a fraction of a second (0.14 second). The peak pressure for the limiting case scenario for Level D condition is 120 psig, due to a single channel failure event. Under this accident condition, it is shown that the reactor structure assembly can withstand the pressure transient and the structural integrity of the core is assured. A finite element model is generated and used to calculate the minimum collapse load. The ANSYS code is used with element type Stif-43 for elastic/plastic, large deformation and small strain analysis. (author). 1 ref., 3 tabs., 9 figs.

  12. Stress analysis for CANDU reactor structure assembly following a postulated p/t, c/t rupture after flow blockage

    International Nuclear Information System (INIS)

    Soliman, S.A.; Lee, T.; Ibrahim, A.M.; Hodgson, S.

    1995-01-01

    This paper describes the collapse load calculations for the reactor structure assembly under the postulated fuel channel flow blockage Level D (faulted) loading condition. Under the flow blockage condition, the primary coolant flow path is obstructed between the inlet and outlet feeder connections to the headers. This, in turn, is postulated to cause the pressure tube and calandria tube to rupture and release hot molten fuel into the moderator, producing a hydrodynamic transient within the calandria shell. The most severe hydrodynamic loads occur within a fraction of a second (0.14 second). The peak pressure for the limiting case scenario for Level D condition is 120 psig, due to a single channel failure event. Under this accident condition, it is shown that the reactor structure assembly can withstand the pressure transient and the structural integrity of the core is assured. A finite element model is generated and used to calculate the minimum collapse load. The ANSYS code is used with element type Stif-43 for elastic/plastic, large deformation and small strain analysis. (author). 1 ref., 3 tabs., 9 figs

  13. 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)

  14. Development and characterization of the control assembly system for the large 2400 MWth Generation IV gas-cooled fast reactor

    International Nuclear Information System (INIS)

    Girardin, G.; Rimpault, G.; Morin, F.; Bosq, J.C.; Coddington, P.; Mikityuk, K.; Chawla, R.

    2008-01-01

    The present paper is related to the design and neutronic characterization of the principal control assembly system for the reference large (2400 MWth) Generation IV gas-cooled fast reactor (GFR), which makes use of ceramic-ceramic (CERCER) plate-type fuel-elements with (U-Pu) carbide fuel contained within a SiC inert matrix. For the neutronic calculations, the deterministic code system ERANOS-2.0 has been used, in association with a full core model including a European fast reactor (EFR)-type pattern for the control assemblies as a starting point. More specifically, the core contains a total of 33 control (control system device: CSD) and safety (diverse safety device: DSD) assemblies implemented in three banks. In the design of the new control assembly system, particular attention was given to the heat generation within the assemblies, so that both neutronic and thermal-hydraulic constraints could be appropriately accounted for. The thermal-hydraulic calculations have been performed with the code COPERNIC, significant coolant mass flow rates being found necessary to maintain acceptable cladding temperatures of the absorber pins. Complementary to the design study, neutronic investigations have been performed to assess the impact of the control assemblies in the GFR core in greater detail (rod interactions, shift of the flux, peaking factors, etc.). Thus, considerable shadowing effects have been observed between the first bank and the safety bank, as also between individual assemblies within the first bank. Large anti-shadowing effects also occur, the most prominent being that between the two CSD banks, where the total assembly worth is almost doubled in comparison to the sum of the individual values. Additional investigations have been performed and, in this context, it has been found that computation of the first-order eigenvalue and the eigenvalue separation is a robust tool to anticipate control assembly interactions in a large fast-spectrum core. One interesting

  15. File list: ALL.Bld.50.AllAg.Peripheral_blood [Chip-atlas[Archive

    Lifescience Database Archive (English)

    Full Text Available ALL.Bld.50.AllAg.Peripheral_blood hg19 All antigens Blood Peripheral blood SRX10033...4075,SRX1034080,SRX1034076,SRX1034079,SRX1034072,SRX1034078,SRX848890,SRX1034067 http://dbarchive.biosciencedbc.jp/kyushu-u/hg19/assembled/ALL.Bld.50.AllAg.Peripheral_blood.bed ...

  16. File list: ALL.Bld.20.AllAg.Peripheral_blood [Chip-atlas[Archive

    Lifescience Database Archive (English)

    Full Text Available ALL.Bld.20.AllAg.Peripheral_blood hg19 All antigens Blood Peripheral blood SRX10033...4075,SRX1034080,SRX1034076,SRX1034079,SRX1034072,SRX1034078,SRX848890,SRX1034067 http://dbarchive.biosciencedbc.jp/kyushu-u/hg19/assembled/ALL.Bld.20.AllAg.Peripheral_blood.bed ...

  17. Multiple lead seal assembly for a liquid-metal-cooled fast-breeder nuclear reactor

    International Nuclear Information System (INIS)

    Hutter, E.; Pardini, J.A.

    1977-01-01

    A reusable multiple lead seal assembly provides leak-free passage of stainless-steel-clad instrument leads through the cover on the primary tank of a liquid-metal-cooled fast-breeder nuclear reactor. The seal isolates radioactive argon cover gas and sodium vapor within the primary tank from the exterior atmosphere and permits reuse of the assembly and the stainless-steel-clad instrument leads. Leads are placed in flutes in a seal body, and a seal shell is then placed around the seal body. Circumferential channels in the body and inner surface of the shell are contiguous and together form a conduit which intersects each of the flutes, placing them in communication with a port through the wall of the seal shell. Liquid silicone rubber sealant is injected into the flutes through the port and conduit; the sealant fills the space in the flutes not occupied by the leads themselves and dries to a rubbery hardness. A nut, threaded onto a portion of the seal body not covered by the seal shell, jacks the body out of the shell and shears the sealant without damage to the body, shell, or leads. The leads may then be removed from the body. The sheared sealant is cleaned from the body, leads, and shell and the assembly may then be reused with the same or different leads. 3 claims, 6 figures

  18. Evaluation of wrapper tube temperatures of fast neutron reactors using the TRANSCOEUR-2 code

    Energy Technology Data Exchange (ETDEWEB)

    Valentin, B.; Brun P. [CEA/DRN/DEC/SECA/LHC CEN, St Paul Lez Durance (France); Chaigne, G. [FRAMATOME/NOVATOME, Lyon (France)

    1995-09-01

    This paper deals with the thermal loading estimation of wrapper tubes using the TRANSCOEUR-2 code. This estimation requires a knowledge of two temperature fields: the first involves the peripheral sub-channel temperatures of each sub-assembly calculated by the design code CADET, and the second, outside the sub-assemblies, is the inter-wrapper flow temperature field calculated by the thermal-hydraulic code TRIO-VF with boundary conditions taken from CADET. Theoretical models of the three codes are presented as well as the first TRANSCOEUR-2 wrapper tube temperature calculation performed on the European Fast Reactor (EFR) Core Design 6/91 (CD 6/91) under nominal power conditions. The results show a temperature variation of 115{degrees}C between the bottom of the lower blanket and the top of the upper blanket fuel sub-assemblies in the center of the core and 95{degrees}C at the core periphery. The wrapper tube temperatures are higher in the center than in the external core.

  19. The Bare Critical Assembly of Natural Uranium and Heavy Water

    Energy Technology Data Exchange (ETDEWEB)

    Popovic, D [Boris Kidric Institute of Nuclear Sciences, Vinca, Belgrade (Yugoslavia)

    1958-07-01

    The first reactor built in Yugoslavia was the bare zero energy heavy water and natural uranium assembly at the Boris Kidric Institute of Nuclear Sciences, Belgrade. The reactor went critical on April 29, 1958. The possession of four tons of natural uranium metal and the temporary availability of seven tons of heavy water encouraged the staff of the Institute to build a critical assembly. A critical assembly was chosen, rather than high flux reactor, because the heavy water was available only temporarily. Besides, a 10 MW, enriched uranium, research reactor is being built at the same Institute and should be ready for operation late this year. It was supposed that the zero energy reactor would provide experience in carrying out critical experiments, operational experience with nuclear reactors, and the possibility for an extensive program in reactor physics. (author)

  20. 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.

  1. 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)

  2. 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)

  3. 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

  4. Nuclear reactor vessel decontamination systems

    International Nuclear Information System (INIS)

    McGuire, P. J.

    1985-01-01

    There is disclosed in the present application, a decontamination system for reactor vessels. The system is operatable without entry by personnel into the contaminated vessel before the decontamination operation is carried out and comprises an assembly which is introduced into the vertical cylindrical vessel of the typical boiling water reactor through the open top. The assembly includes a circular track which is centered by guideways permanently installed in the reactor vessel and the track guides opposed pairs of nozzles through which water under very high pressure is directed at the wall for progressively cutting and sweeping a tenacious radioactive coating as the nozzles are driven around the track in close proximity to the vessel wall. The whole assembly is hoisted to a level above the top of the vessel by a crane, outboard slides on the assembly brought into engagement with the permanent guideways and the assembly progressively lowered in the vessel as the decontamination operation progresses. The assembly also includes a low pressure nozzle which forms a spray umbrella above the high pressure nozzles to contain radioactive particles dislodged during the decontamination

  5. 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

  6. Reactor vessel sealing plug

    International Nuclear Information System (INIS)

    Dooley, R.A.

    1986-01-01

    An apparatus is described for sealing a cold leg nozzle of a nuclear reactor pressure vessel from a remote location comprising: at least one sealing plug for mechanically sealing the nozzle from the inside of the reactor pressure vessel. The sealing plug includes a plate and a cone assembly having an end part receptive in the nozzle, the plate being axially moveable relative to the cone assembly. The plate and cone assembly have confronting bevelled edges defining an opening therebetween. A primary O-ring is disposed about the opening and is supported on the bevelled edges, the plate being guidably mounted to the cone assembly for movement toward the cone assembly to radially expand the primary O-ring into sealing engagement with the nozzle. A means is included for providing relative movement between the outer plate and the cone assembly

  7. File list: Pol.Bld.50.AllAg.Peripheral_blood_stem_cells [Chip-atlas[Archive

    Lifescience Database Archive (English)

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  8. File list: Pol.Bld.20.AllAg.Peripheral_blood_stem_cells [Chip-atlas[Archive

    Lifescience Database Archive (English)

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  9. File list: Unc.Bld.50.AllAg.Peripheral_blood_stem_cells [Chip-atlas[Archive

    Lifescience Database Archive (English)

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  10. File list: Pol.Bld.05.AllAg.Peripheral_blood_stem_cells [Chip-atlas[Archive

    Lifescience Database Archive (English)

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  11. File list: Unc.Bld.05.AllAg.Peripheral_blood_stem_cells [Chip-atlas[Archive

    Lifescience Database Archive (English)

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  12. File list: Unc.Bld.10.AllAg.Peripheral_blood_stem_cells [Chip-atlas[Archive

    Lifescience Database Archive (English)

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  13. File list: Unc.Bld.20.AllAg.Peripheral_blood_stem_cells [Chip-atlas[Archive

    Lifescience Database Archive (English)

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  14. File list: Pol.Bld.10.AllAg.Peripheral_blood_stem_cells [Chip-atlas[Archive

    Lifescience Database Archive (English)

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  15. Permanent cavity seal ring for a nuclear reactor containment arrangement

    International Nuclear Information System (INIS)

    Swidwa, K.J.; Salton, R.B.; Marshall, J.R.

    1990-01-01

    This patent describes a nuclear reactor containment arrangement. It comprises: a reactor pressure vessel which thermally expands and contracts during cyclic operation of the reactor, the vessel having a peripheral wall and a horizontally outwardly extending flange thereon; a containment wall having a shelf, the wall spaced from and surrounding the peripheral wall of the reactor pressure vessel defining an annular expansion gap therebetween, and an annular ring seal extending across the annular expansion gap to provide a water-tight seal therebetween

  16. Plasma core reactor applications

    International Nuclear Information System (INIS)

    Latham, T.S.; Rodgers, R.J.

    1976-01-01

    Analytical and experimental investigations are being conducted to demonstrate the feasibility of fissioning uranium plasma core reactors and to characterize space and terrestrial applications for such reactors. Uranium hexafluoride (UF 6 ) fuel is injected into core cavities and confined away from the surface by argon buffer gas injected tangentially from the peripheral walls. Power, in the form of thermal radiation emitted from the high-temperature nuclear fuel, is transmitted through fused-silica transparent walls to working fluids which flow in axial channels embedded in segments of the cavity walls. Radiant heat transfer calculations were performed for a six-cavity reactor configuration; each cavity is approximately 1 m in diameter by 4.35 m in length. Axial working fluid channels are located along a fraction of each cavity peripheral wall

  17. Welding facilities for NPP assembling

    International Nuclear Information System (INIS)

    Rojtenberg, S.S.

    1987-01-01

    Recommendations concerning the choice of equipment for welding in pre-assembling work shops, in the enlarging assembling shops and at the assembling site, are given. Advanced production automatic welders and semiautomatic machines, applied during the NPP equipment assembling as well as automatic machines specially produced for welding the main reactor components and pipelines are described. Automatic and semiautomatic machine and manual welding post supply sources are considered

  18. Study of graphite reactivity worth on well-defined cores assembled on LR-0 reactor

    International Nuclear Information System (INIS)

    Košťál, Michal; Rypar, Vojtěch; Milčák, Ján; Juříček, Vlastimil; Losa, Evžen; Forget, Benoit; Harper, Sterling

    2016-01-01

    Highlights: • A light water critical facility for graphite reactivity worth measurements. • Comparison of calculated and measured k eff . • Effect of graphite description on k eff . - Abstract: Graphite is an often-used moderating material on the basis of its good moderating power and very low absorption cross section. This small absorption cross section permits the use of natural or low-enriched uranium in graphite moderated reactors. Graphite is now being considered as the moderator for Fluoride-salt-cooled High Temperature Reactors (FHR). The critical moderator level was measured for various graphite block configurations in an experimental dry assembly of the LR-0 reactor. Comparisons with experiments were performed between Monte Carlo simulation tools for which satisfactory agreement was obtained with the exception of some systematic discrepancies. The larger discrepancies were observed when using the ENDF/B-VII.0 library. To decrease the uncertainties, based on conservative assumptions, relative comparisons were done. The results provided by the different nuclear data libraries are within 3 sigma interval of experimental uncertainties. It has been determined that differences between the results of calculations are caused by variations in the (n,n), (n,n′), (n,g) reactions and also by various angular distributions, while the (n,g) cross section variations play only a minor role for these configurations.

  19. Training reactor deployment. Advanced experimental course on designing new reactor cores

    International Nuclear Information System (INIS)

    Skoda, Radek

    2009-01-01

    Czech Technical University in Prague (CTU) operating its training nuclear reactor VR1, in cooperation with the North West University of South Africa (NWU), is applying for accreditation of the experimental training course ''Advanced experimental course on designing the new reactor core'' that will guide the students, young nuclear engineering professionals, through designing, calculating, approval, and assembling a new nuclear reactor core. Students, young professionals from the South African nuclear industry, face the situation when a new nuclear reactor core is to be build from scratch. Several reactor core design options are pre-calculated. The selected design is re-calculated by the students, the result is then scrutinized by the regulator and, once all the analysis is approved, physical dismantling of the current core and assembling of the new core is done by the students, under a close supervision of the CTU staff. Finally the reactor is made critical with the new core. The presentation focuses on practical issues of such a course, desired reactor features and namely pedagogical and safety aspects. (orig.)

  20. 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.)

  1. Core arrangement in BWR type reactors

    International Nuclear Information System (INIS)

    Asano, Masayuki.

    1981-01-01

    Purpose: To decrease the number of fuel assemblies whose locations are to be changed upon fuel exchange, as well as unify the power distribution in the core by arranging, in a chess board configuration, a plurality pattern of unit reactor lattices each containing fuel assemblies of different burnup degrees in orthogonal positions to each other. Constitution: A first pattern of unit reactor lattice is formed by disposing fuel assemblies of burnup degree 1 and fuel assemblies of burnup degree 3 at orthogonal positions to each other. A second pattern of unit reactor lattice is formed by disposing fuel assemblies of burnup degree 2 and fuel assemblies of burnup degree 1 at orthogonal positions to each other. The unit lattices each in such a dispositions are arranged in a chess board arrangement. Since, the fuel assemblies of the burnup degree 1 in the first pattern unit lattices proceed to the burnup degree 2 and the fuel assemblies of the burnup degree 2 in the second pattern unit lattices proceed to the burnup degree 3 up to the fuel exchange stage, fuel exchange and movement have only to be made, not for those fuel assemblies, but for another half of the fuel assemblies. (Kawakami, Y.)

  2. Snubber assembly for a control rod drive

    International Nuclear Information System (INIS)

    Matthews, J.C.

    1978-01-01

    A snubber cartridge assembly is mounted to the nozzle of a control rod drive mechanism to insure that the snubber assembly will be located within the liquid filled section of a nuclear reactor vessel whenever the control rod drive is assembled thereto. The snubber assembly includes a piston mounted proximate to the control rod connecting end of the control rod drive leadscrew to allow the piston to travel within the liquid filled snubber cartridge and controllably exhaust liquid therefrom during a ''scram'' condition. The snubber cartridge provides three separate areas of increasing resistance to piston travel to insure a speedy but safe ''scram'' of the control rod into the reactor

  3. Effects of low heterogeneity in fast critical assemblies

    International Nuclear Information System (INIS)

    Belov, S.P.; Dulin, V.A.; Zhukov, A.V.; Kuzin, E.N.; Mozhaev, V.K.; Sitnikov, V.I.; Tsibulya, A.M.; Shapar', A.V.; Zayfert, E.; Kuntsman, B.; Khayntsel'man, B.

    1989-01-01

    The problem of the low heterogeneity of fast critical assemblies, which are used to simulate fast reactors that are under design, has begun to assume increasing importance as the errors in nuclear data and group constants decrease and the capabilities of design codes improve. The design of the fuel channels of the fast critical assemblies of a BFS differs from that of the fuel subassemblies of a power reactor. The principal difference is that critical assemblies have a more heterogeneous structure than a reactor core does. As a result, the effects of this heterogeneity turn out to be appreciable for a number of functionals. Of particular interest was the measurement of the main neutronic characteristics of a fast reactor in its actual design and in the mockup produced by using BFS facilities. The authors measured and calculated the most important functionals (the ratios of the average cross sections of the main absorbing and fissioning elements, etc.) for both a homogeneous medium (fuel assemblies) and a heterogeneous medium (slugs, tubes) of practically identical composition. The objective of this work was to compare the discrepancy between experiment and calculations for the central functionals in the homogeneous and heterogeneous cases after corrections. This is a check of the accuracy of the simulation of homogeneous cores in fast power reactors by using the tools of the BFS fast critical assembly

  4. Computer simulation of radiation processes in reactor facilities

    International Nuclear Information System (INIS)

    Gann, V.V.; Abdulaev, A.M.; Zhukov, A.I.; Marekhin, S.V.; Soldatov, S.A.

    2009-01-01

    The paper describes experience of the code system ALPHA-H/PHOENIX-H/ANC-H (APA) and the code MCNP usage for fuel assembly neutronic calculations and modeling of VVER-1000 reactor core. Using Monte Carlo code MCNP, calculations of neutron field and pin-by-pin energy deposition distributions are provided for different type of assemblies in reactor core. An MCNP model for unit 3 Zaporozhye NPP reactor core was designed. Calculations for pin-by-pin energy deposition in the reactor core were performed using the code system APA and the code MCNP. Comparison of these calculations shows rather high precision of APA calculation for energy deposition in the fuel rods and assemblies operated in the reactor core

  5. 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

  6. File list: Oth.Bld.05.AllAg.Peripheral_blood_stem_cells [Chip-atlas[Archive

    Lifescience Database Archive (English)

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    Lifescience Database Archive (English)

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  8. File list: DNS.Bld.10.AllAg.Peripheral_blood_stem_cells [Chip-atlas[Archive

    Lifescience Database Archive (English)

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  9. File list: Oth.Bld.10.AllAg.Peripheral_blood_stem_cells [Chip-atlas[Archive

    Lifescience Database Archive (English)

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  10. File list: Oth.Bld.50.AllAg.Peripheral_blood_stem_cells [Chip-atlas[Archive

    Lifescience Database Archive (English)

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    Lifescience Database Archive (English)

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    Lifescience Database Archive (English)

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    Lifescience Database Archive (English)

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    Lifescience Database Archive (English)

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  17. File list: NoD.Bld.20.AllAg.Peripheral_blood [Chip-atlas[Archive

    Lifescience Database Archive (English)

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  18. Modelling of flow stabilization by the swirl of a peripheral flow as applied to plasma reactors

    International Nuclear Information System (INIS)

    Volchkov, E.P.; Lebedev, V.P.; Terekhov, V.I.; Shishkin, N.E.

    2000-01-01

    The gas-swirl stabilization of plasma jets is one of effective methods of its retention in the near-axial area of channels in generators of low-temperature plasma. Except the effect of gas-dynamic compression, the peripheral swirl allows to solve another urgent problem - to protect the reactor walls from the heat influence of the plasma jet. Swirl flows are also used for the flow structure formation and control of the heat and gas-dynamic characteristics of different power devices and apparatuses, using high-temperature working media: in swirl furnaces and burners, in aviation engines, etc. Investigations show that during swirl stabilization the gas-dynamic structure of the flow influences significantly the spatial stability of the plasma column and its characteristics

  19. Results of trial operation of new generation assemblies with improved vibration stability for WWER-440 reactors of V-230 and V-213 plants

    International Nuclear Information System (INIS)

    Vasilchenko, I.; Lushin, V.; Shumeev, A.

    2006-01-01

    The main task in developing a fuel design of new generation for WWER-440 power units (particularly for power units of V-213 type) is to increase the effectiveness of fuel utilization. For this purpose the following design solutions were realized: 1) increase of fuel loading (increase of fuel stack in a fuel rod); 2) optimization of the water-uranium relation (at the cost of some decrease of outside diameter of a fuel rod from 9,1 mm to 9,07 mm and increase of fuel rod pitch in a bundle from 12,2 mm up to 12,3 mm); 3) decrease of harmful absorption of neutrons (due to decrease of hafnium content in zirconium materials from 0,05 % to 0,01 % and decrease of zirconium in transition to thickness of FA casing of ERC assembly - 1,5 mm). The parallel task is to increase vibration stability of fuel assemblies based on the verification and justification of the following technical solutions: 1)improvement of vibration stability due to SG rearrangement along the fuel rod bundle height and increasement of height of the first three grids along the course of coolant flow; 2) decreasement of backlashes in the assembly 'fuel rod - support grid' due to introduction of a new elastic tip of fuel rods; 3) introduction of a stiffening rib under the support grid and attachment of the central tube in the support grid by welding; 4) introduction of special sleeve in the protective grid, which permits to fasten a bundle against radial displacements in its upper part; 5) introduction of slots, located at different height marks of the central tube, permits to exclude distortion and destruction of spacing grids as a result of increase of fuel rod temperature; 6) installation of filters at the inlet. According to design considerations with due regard for statistics of fuel assemblies damage, the filters will be mounted at the working assembly (WA) inlet. In new generation fuel assemblies, the problem of decrease of local flash-up of neutrons in the WA peripheral fuel rods, surrounding the CPS

  20. 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.)

  1. Fabrication and quality assurance of some important components and sub-assemblies for Prototype Fast Breeder Reactor (PFBR) project

    International Nuclear Information System (INIS)

    Dutta, N.G.; More, S.S.

    2010-01-01

    Bharatiya Nabhikiya Vidyut Nigam (BHAVINI) is engaged in construction of 500 MWe prototype fast breeder reactor (PFBR) at Kalpakkam, Chennai. In this very important and prestigious national programmed M/s Kay Bouvet Engg. Pvt. Ltd. (M/s KBEPL) Satara is contributing in a major way by supplying many important sub-assemblies. M/s KBEPL is engaged in manufacturing, quality assurance and supply of many subassemblies of PFBR like under water trolley (UWT), shielding door, container and container storage rack (CSR), vessel in fuel transfer cell (FTC), personnel air lock (PAL), emergency air lock (EAL) and material air lock (MAL), absorber rod drive mechanism (ARDM) flask assembly and carriage in MAL etc. Two partition doors and four nos. of embedded parts (SS 304L) have already been supplied to Bhavini. The paper deals with manufacturing and Q.A. activities being carried out for supply of these important assemblies to PFBR projects. (author)

  2. 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.

  3. 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)

  4. 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

  5. Nuclear reactor refueling system

    International Nuclear Information System (INIS)

    Wade, E.E.

    1978-01-01

    A system for transferring fuel assemblies between a nuclear reactor core and a fuel storage area while the fuel assembies remain completely submerged in a continuous body of coolant is described. The system comprises an in-vessel fuel transfer machine located inside the reactor vessel and an ex-vessel fuel transfer machine located in a fuel storage tank. The in-vessel fuel transfer machine comprises two independently rotatable frames with a pivotable fuel transfer apparatus disposed on the lower rotatable frame. The ex-vessel fuel transfer machine comprises one frame with a pivotable fuel transfer apparatus disposed thereon. The pivotable apparatuses are capable of being aligned with each other to transfer a fuel assembly between the reactor vessel and fuel storage tank while the fuel assembly remains completely submerged in a continuous body of coolant. 9 claims, 7 figures

  6. 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)

  7. Performance of 9 x 9 demonstration assemblies in Dresden-2

    International Nuclear Information System (INIS)

    Bain, G.M.

    1992-06-01

    The Electric Power Research Institute, Empire State Electric energy Research Corporation and Siemens Nuclear Power corporation jointly sponsored a program to monitor the in-reactor performance of 9x9 BWR fuel. The program was conducted in Dresden-2, with four 9x9 lead assemblies and one 8x8 reference assembly. These assemblies were loaded at the beginning of reactor Cycle 9 and completed four cycles of operation. All five assemblies were discharged after reactor Cycle 12 (EOC12) in September 1990. the 9x9 assemblies reached an average exposure of 35.7 GWd/MTU and the 8x8 reference assembly reached a burnup of 34.2 GWd/MTU. This final program report evaluates the performance of the 9x9 and 8x8 fuel assemblies, based on results from all four poolside examinations, analysis of the operating histories, and ramp tests conducted on rod segments under another program. Overall, both 9x9 and 8x assemblies performed well during the four cycles of irradiation

  8. File list: NoD.Bld.05.AllAg.Peripheral_blood_stem_cells [Chip-atlas[Archive

    Lifescience Database Archive (English)

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  9. File list: InP.Bld.10.AllAg.Peripheral_blood_stem_cells [Chip-atlas[Archive

    Lifescience Database Archive (English)

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  10. File list: NoD.Bld.20.AllAg.Peripheral_blood_stem_cells [Chip-atlas[Archive

    Lifescience Database Archive (English)

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  11. File list: NoD.Bld.10.AllAg.Peripheral_blood_stem_cells [Chip-atlas[Archive

    Lifescience Database Archive (English)

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  12. File list: NoD.Bld.50.AllAg.Peripheral_blood_stem_cells [Chip-atlas[Archive

    Lifescience Database Archive (English)

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  13. File list: InP.Bld.05.AllAg.Peripheral_blood_stem_cells [Chip-atlas[Archive

    Lifescience Database Archive (English)

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  14. File list: InP.Bld.50.AllAg.Peripheral_blood_stem_cells [Chip-atlas[Archive

    Lifescience Database Archive (English)

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  15. Fundamentals of boiling water reactor systems

    International Nuclear Information System (INIS)

    Mattern, J.

    1976-01-01

    The reactor assembly consists of the reactor vessel, its internal components of the core, shroud, steam separator, dryer assemblies, feedwater spargers, internal recirculation pumps and control rod drive housings. Connected to the steam lines are the pressure relief valves which protect the pressure boundary from damage due to overpressure. (orig./TK) [de

  16. 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

  17. Nuclear reactors built, being built, or planned, 1994

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-07-01

    This document contains unclassified information about facilities built, being built, or planned in the United States for domestic use or export as of December 31, 1994. The Office of Scientific and Technical Information, US Department of Energy, gathers this information annually from Washington headquarters and field offices of DOE; from the US Nuclear Regulatory Commission (NRC); from the US reactor manufacturers who are the principal nuclear contractors for foreign reactor locations; from US and foreign embassies; and from foreign governmental nuclear departments. The book consists of three divisions, as follows: a commercial reactor locator map and tables of the characteristic and statistical data that follow; a table of abbreviations; tables of data for reactors operating, being built, or planned; and tables of data for reactors that have been shut down permanently or dismantled. The reactors are subdivided into the following parts: Civilian, Production, Military, Export, and Critical Assembly. Export reactor refers to a reactor for which the principal nuclear contractor is a US company -- working either independently or in cooperation with a foreign company (Part 4). Critical assembly refers to an assembly of fuel and moderator that requires an external source of neutrons to initiate and maintain fission. A critical assembly is used for experimental measurements (Part 5).

  18. Nuclear reactors built, being built, or planned: 1995

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-08-01

    This report contains unclassified information about facilities built, being built, or planned in the US for domestic use or export as of December 31, 1995. The Office of Scientific and Technical Information, US Department of Energy, gathers this information annually from Washington headquarters and field offices of DOE; from the US Nuclear Regulatory Commission (NRC); from the US reactor manufacturers who are the principal nuclear contractors for foreign reactor locations; from US and foreign embassies; and from foreign governmental nuclear departments. The book consists of three divisions, as follows: (1) a commercial reactor locator map and tables of the characteristic and statistical data that follow; a table of abbreviations; (2) tables of data for reactors operating, being built, or planned; and (3) tables of data for reactors that have been shut down permanently or dismantled. The reactors are subdivided into the following parts: Civilian, Production, Military, Export, and Critical Assembly. Export reactor refers to a reactor for which the principal nuclear contractor is a US company--working either independently or in cooperation with a foreign company (Part 4). Critical assembly refers to an assembly of fuel and moderator that requires an external source of neutrons to initiate and maintain fission. A critical assembly is used for experimental measurements (Part 5).

  19. Nuclear reactors built, being built, or planned, 1994

    International Nuclear Information System (INIS)

    1995-07-01

    This document contains unclassified information about facilities built, being built, or planned in the United States for domestic use or export as of December 31, 1994. The Office of Scientific and Technical Information, US Department of Energy, gathers this information annually from Washington headquarters and field offices of DOE; from the US Nuclear Regulatory Commission (NRC); from the US reactor manufacturers who are the principal nuclear contractors for foreign reactor locations; from US and foreign embassies; and from foreign governmental nuclear departments. The book consists of three divisions, as follows: a commercial reactor locator map and tables of the characteristic and statistical data that follow; a table of abbreviations; tables of data for reactors operating, being built, or planned; and tables of data for reactors that have been shut down permanently or dismantled. The reactors are subdivided into the following parts: Civilian, Production, Military, Export, and Critical Assembly. Export reactor refers to a reactor for which the principal nuclear contractor is a US company -- working either independently or in cooperation with a foreign company (Part 4). Critical assembly refers to an assembly of fuel and moderator that requires an external source of neutrons to initiate and maintain fission. A critical assembly is used for experimental measurements (Part 5)

  20. Nuclear reactors built, being built, or planned: 1995

    International Nuclear Information System (INIS)

    1996-08-01

    This report contains unclassified information about facilities built, being built, or planned in the US for domestic use or export as of December 31, 1995. The Office of Scientific and Technical Information, US Department of Energy, gathers this information annually from Washington headquarters and field offices of DOE; from the US Nuclear Regulatory Commission (NRC); from the US reactor manufacturers who are the principal nuclear contractors for foreign reactor locations; from US and foreign embassies; and from foreign governmental nuclear departments. The book consists of three divisions, as follows: (1) a commercial reactor locator map and tables of the characteristic and statistical data that follow; a table of abbreviations; (2) tables of data for reactors operating, being built, or planned; and (3) tables of data for reactors that have been shut down permanently or dismantled. The reactors are subdivided into the following parts: Civilian, Production, Military, Export, and Critical Assembly. Export reactor refers to a reactor for which the principal nuclear contractor is a US company--working either independently or in cooperation with a foreign company (Part 4). Critical assembly refers to an assembly of fuel and moderator that requires an external source of neutrons to initiate and maintain fission. A critical assembly is used for experimental measurements (Part 5)

  1. File list: His.Bld.05.AllAg.Peripheral_blood_stem_cells [Chip-atlas[Archive

    Lifescience Database Archive (English)

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  2. File list: His.Bld.20.AllAg.Peripheral_blood_stem_cells [Chip-atlas[Archive

    Lifescience Database Archive (English)

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  3. File list: His.Bld.50.AllAg.Peripheral_blood_stem_cells [Chip-atlas[Archive

    Lifescience Database Archive (English)

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  4. File list: His.Bld.10.AllAg.Peripheral_blood_stem_cells [Chip-atlas[Archive

    Lifescience Database Archive (English)

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  5. Design and fabrication of self-powered in-core neutron flux monitor assembly

    International Nuclear Information System (INIS)

    Chung, M.K.; Cho, S.W.; Kang, H.D.; Cho, K.K.; Cho, B.S.; Kang, S.S.

    1980-01-01

    This is the final report on the prototypical fabrication of an in-core neutron flux monitor detector assembly for a specific power reactor conducted by KAERI from July 1, 1978 to December 31, 1979. It is well known that power reactors require a large number of in-core neutron flux detector for reactor regulation and the structures of detector assemblies are different from reactor to reactor. Therefore, from the nature of this project, it should be noted here that the target model of the prototypical farbrication of an in-core neutron flux monitor detector assembly is a VFD-2 System for Wolsung CANDU. It is concluded that fabrication of in-core neutron flux monitor detector assembly for CANDU reactor is technically feasible and will bring economical benefit as much as 50 % of the unit price if they are fabricated in Korea by using partially materials which are available from local market. (author)

  6. 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)

  7. Reactor monitor

    International Nuclear Information System (INIS)

    Takada, Tamotsu.

    1992-01-01

    The device of the present invention monitors a reactor so that each of the operations for the relocation of fuel assemblies and the withdrawal and the insertion of control rods upon exchange of fuel assemblies and control rods in the reactor. That is, when an operator conducts relocating operation by way of a fuel assembly operation section, the device of the present invention judges whether the operation indication is adequate or not, based on the information of control rod arrangement in a control rod memory section. When the operation indication is wrong, a stop signal is sent to a fuel assembly relocating device. Further, when the operator conducts control rod operation by way of a control rod operation section, the device of the present invention judges in the control rod withdrawal judging section, as to whether the operation indication given by the operator is adequate or not by comparing it with fuel assembly arrangement information. When the operation indication is wrong, a stop signal is sent to control rod drives. With such procedures, increase of nuclear heating upon occurrence of erroneous operation can be prevented. (I.S.)

  8. Completion of the Heysham 2 peripheral manipulator

    International Nuclear Information System (INIS)

    Shipp, R.; Ewen, R.O.

    1996-01-01

    The in-service inspection strategy for the AGR power station at Heysham 2 envisaged a suite of five manipulators to be used for inserting TRIUMPH television cameras into the reactor vessel. Prior to power raising, four of the five had been successfully commissioned and have been in regular use during the subsequent statutory outages. The final device, the Peripheral Manipulator (PM), was eventually completed prior to the 1994 outage and has been successfully deployed on reactor for both the 1994 and 1995 outages. The paper describes the design of the manipulator, its operation and scope of use in the Heysham 2 reactors. (Author)

  9. 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.)

  10. Reactor PIK construction

    International Nuclear Information System (INIS)

    Konoplev, Kir

    2003-01-01

    The construction work at the 100 MW researches reactor PIK in year 2002 was in progress. The main activity was concentrated on mechanical, ventilation and electrical equipment. Some systems and subsystems are under adjustment. Hydraulic driving gear for beam shutters are finished in installation, rinsing, and adjusting. Regulating rods test assembling was done. On the critical assembly the first reactor fueling was tested to evaluate the starting neutron source intensity and a sufficiency of existing control and instrument board. Mainline of the PIK facility design and neutron parameters are presented. (author)

  11. Preparation and planning for the replacement of the Oregon State University TRIGA reactor rotary specimen rack assembly

    International Nuclear Information System (INIS)

    Anderson, T.V.; Dodd, B.; Johnson, A.G.; Carpenter, W.T.

    1984-01-01

    Recently there have been a number of indications that the rotating rack may be approaching the end of its useful life. In order to benefit from the experience of other reactors who have removed and replaced their rotating racks, General Atomic (GA) was contacted and previous TRIGA Conference proceedings were scanned. It was determined that a number of facilities, had experienced difficulties with their lazy susans and eventually had to replace them. However, most of the written descriptions of this project were not sufficiently detailed to be of great use. The purpose of this paper is to identify some of the more important questions related to the replacement of our rotating rack assembly and OSU's currently proposed solutions, with a view towards soliciting ideas from other members of the TRIGA reactor community

  12. Nuclear reactor

    International Nuclear Information System (INIS)

    Anthony, A.J.; Gruber, E.A.

    1979-01-01

    A nuclear reactor with control rods in channels between fuel assemblies wherein the fuel assemblies incorporate guide rods which protrude outwardly into the control rod channels to prevent the control rods from engaging the fuel elements. The guide rods also extend back into the fuel assembly such that they are relatively rigid members. The guide rods are tied to the fuel assembly end or support plates and serve as structural members which are supported independently of the fuel element. Fuel element spacing and support means may be attached to the guide rods. 9 claims

  13. 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

  14. Simulation of the core flowering End-of-life test realized on Phenix reactor

    International Nuclear Information System (INIS)

    Prulhiere, G.; Fontaine, B.; Frosio, T.

    2013-01-01

    After the definitive shutdown of the Phenix sodium cooled fast reactor and before its decommissioning, a final set of tests were performed covering core physics, fuel behavior and thermal hydraulics areas. In addition, the program included two tests related to the comprehension of the four negative reactivity transients experienced during the reactor operation in 1989 and 1990. One of these tests, called 'core flowering test' focused on the relation between sub-assemblies mechanical displacements and reactivity variations. This test was carried out by introducing a mechanical device pushing on the six fuel assemblies neighbors. This device was located at two different core positions: at the center and at a peripheral one. The reactivity effect induced by core flowering was measured at different temperatures in the range of 180 to 350 Celsius degrees. The simulation of such a test requires the use of a neutronic computing code which is not compelled to the definition of regular geometrical lattices. Moreover, a system permitting an easy and change-allowing way to define geometries and deformations is needed. That is why the use of a Monte Carlo code like TRIPOLI coupled to ROOT system was chosen to simulate this test. The displacement of each sub-assembly was estimated upstream of this study using the static mechanics code HARMONIE. To perform this calculations with a satisfying precision, several hundreds millions of neutrons particles were needed for the modelling. (author)

  15. 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

  16. Rodded shutdown system for a nuclear reactor

    International Nuclear Information System (INIS)

    Golden, M.P.; Govi, A.R.

    1978-01-01

    A top mounted nuclear reactor diverse rodded shutdown system utilizing gas fed into a pressure bearing bellows region sealed at the upper extremity to an armature is described. The armature is attached to a neutron absorber assembly by a series of shafts and connecting means. The armature is held in an uppermost position by an electromagnet assembly or by pressurized gas in a second embodiment. Deenergizing the electromagnet assembly, or venting the pressurized gas, causes the armature to fall by the force of gravity, thereby lowering the attached absorber assembly into the reactor core

  17. Effect of reactor size on the breeding economics of LMFBR blankets

    International Nuclear Information System (INIS)

    Tagishi, A.; Driscoll, M.J.

    1975-02-01

    The effect of reactor size on the neutronic and economic performance of LMFBR blankets driven by radially-power-flattened cores has been investigated using both simple models and state-of-the-art computer methods. Reactor power ratings in the range 250 to 3000 MW(e) were considered. Correlations for economic breakeven and optimum irradiation times and blanket thicknesses have been developed for batch-irradiated blankets. It is shown that a given distance from the core-blanket interface the fissile buildup rate per unit volume remains very nearly constant in the radial blanket as (radially-power-flattened, constant-height) core size increases. As a consequence, annual revenue per blanket assembly, and breakeven and optimum irradiation times and optimum blanket dimensions, are the same for all reactor sizes. It is also shown that the peripheral core fissile enrichment, hence neutron leakage spectra, of the (radially-power-flattened, constant-height) cores remains essentially constant as core size increases. Coupled with the preceding observations, this insures that radial blanket breeding performance in demonstration-size LMFBR units will be a good measure of that in much larger commercial LMFBR's

  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. LMFBR type reactor

    International Nuclear Information System (INIS)

    Masumi, Ryoji; Kawashima, Katsuyuki; Kurihara, Kunitoshi.

    1988-01-01

    Purpose: To flatten the power distribution while maintaining the flattening in the axial power distribution in LMFBR type reactors. Constitution: Main system control rods are divided into control rods used for the operation and starting rods used for the starting of the reactor, and the starting rods are disposed in the radial periphery of the reactor core, while the control rods are disposed to the inside of the starting rods. With such a constitution, adjusting rods can be disposed in the region where the radial power peaking is generated to facilitate the flattening of the power distribution even in such a design that the ratio of the number of control rods to that of fuel assemblies is relatively large. That is, in this reactor, the radial power peaking is reduced by about 10% as compared with the conventional reactor core. As a result, the maximum linear power density during operation is reduced by about 10% to increase the thermal margin of the reactor core. If the maximum linear power density is set identical, the number of the fuel assemblies can be decreased by about 10%, to thereby reduce the fuel production cost. (K.M.)

  1. 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

  2. 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.

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

    International Nuclear Information System (INIS)

    Koguchi, Kazushige.

    1995-01-01

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

  4. 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.

  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. 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)

  7. BWR type reactors

    International Nuclear Information System (INIS)

    Nakajima, Yoshitaka

    1983-01-01

    Purpose: To decrease the control rod exchanging frequency by increasing the working life of control rods for ordinary operation with large neutron irradiation dose, to thereby decrease the exposure dose for operators performing exchanging work, as well as decrease the amount of radioactive wastes resulted upon exchange of the control rods. Constitution: Hafnium solid metal is employed as the neutron absorber of control rods for usual operation inserted into and withdrawn from fuel assemblies for the reactor power control over the entire cycle of the ordinary reactor operation and boron carbide powder is employed as the neutron absorber for emergency control rods to be inserted between the fuel assemblies only upon reactor scram or shutdown, whereby the working life of the control rods for ordinary reactor operation with greater neutron irradiation dose can be improved. Accordingly, the control rod exchanging frequency can be reduced to decrease the exposure dose to the operator for conducting the exchanging work. (Yoshihara, H.)

  8. PV module mounting method and mounting assembly

    Science.gov (United States)

    Lenox, Carl J.S.; Johnson, Kurt M.

    2013-04-23

    A method for mounting PV modules to a deck includes selecting PV module layout pattern so that adjacent PV module edges are spaced apart. PV mounting and support assemblies are secured to the deck according to the layout pattern using fasteners extending into the deck. The PV modules are placed on the PV mounting and support assemblies. Retaining elements are located over and secured against the upper peripheral edge surfaces of the PV modules so to secure them to the deck with the peripheral edges of the PV modules spaced apart from the deck. In some examples a PV module mounting assembly, for use on a shingled deck, comprises flashing, a base mountable on the flashing, a deck-penetrating fastener engageable with the base and securable to the deck so to secure the flashing and the base to the shingled deck, and PV module mounting hardware securable to the base.

  9. 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.)

  10. 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)

  11. Snubber assembly for a control rod drive

    International Nuclear Information System (INIS)

    1976-01-01

    A snubber cartridge assembly is described which is mounted to the nozzle of a control rod drive mechanism to insure that it will be located within the liquid filled section of a nuclear reactor vessel whenever the control rod drive is assembled thereto. The snubber assembly includes a piston-mounted proximate to the control rod connecting end of the control rod drive leadscrew to allow the piston to travel within the liquid filled snubber cartridge and controllable exhaust the liquid during a 'scram' condition. The snubber cartridge provides three separate areas of increasing resistance to piston travel to insure a speedy but safe 'scram' of the control rod into the reactor

  12. 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

  13. Reactor

    International Nuclear Information System (INIS)

    Toyama, Masahiro; Kasai, Shigeo.

    1978-01-01

    Purpose: To provide a lmfbr type reactor wherein effusion of coolants through a loop contact portion is reduced even when fuel assemblies float up, and misloading of reactor core constituting elements is prevented thereby improving the reactor safety. Constitution: The reactor core constituents are secured in the reactor by utilizing the differential pressure between the high-pressure cooling chamber and low-pressure cooling chamber. A resistance port is formed at the upper part of a connecting pipe, and which is connect the low-pressure cooling chamber and the lower surface of the reactor core constituent. This resistance part is formed such that the internal sectional area of the connecting pipe is made larger stepwise toward the upper part, and the cylinder is formed larger so that it profiles the inner surface of the connecting pipe. (Aizawa, K.)

  14. 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

  15. Analysis of strategies for improving uranium utilization in pressurized water reactors. Annual technical progress report for FY 1980

    International Nuclear Information System (INIS)

    Sefcik, J.A.; Driscoll, M.J.; Lanning, D.D.

    1981-01-01

    Systematic procedures have been devised and applied to evaluate core design and fuel management strategies for improving uranium utilization in Pressurized Water Reactors operated on a once-through fuel cycle. A principal objective has been the evaluation of suggested improvements on a self-consistent basis, allowing for concurrent changes in dependent variables such as core leakage and batch power histories, which might otherwise obscure the sometimes subtle effects of interest. Two levels of evaluation have been devised: a simple but accurate analytic model based on the observed linear variations in assembly reactivity as a function of burnup; and a numerical approach, embodied in a computer program, which relaxes this assumption and combines it with empirical prescriptions for assembly (or batch) power as a function of reactivity, and core leakage as a function of peripheral assembly power. State-of-the-art physics methods, such as PDQ-7, were used to verify and supplement these techniques

  16. The review of the reactor physics experiments carried out on the LR-0 research reactor NRI Rez plc for reactors of the VVER type

    International Nuclear Information System (INIS)

    Hudec, Frantisek; Jansky, Bohumil; Juricek, Vlastimil; Mikus, Jan; Novak, Evzen; Osmera, Bohumil; Posta, Severin; Rypar, Vojtech; Svadlenkova, Marie

    2010-01-01

    LR-0 is an experimental zero power reactor mainly used for the determination of the neutron-physical characteristics of WWER and PWR type reactor lattices and shielding with UO2 or MOX fuel. Its major assets include capability to design and operate multizone cores, i.e. substituted cores, with an inner inserted part in hexagonal or square geometry (driven by LR-0 standard assemblies); Standard and special supporting plates for mock-up experiments; special supporting plates, which enables the triangular symmetrical assembly arrangement with an arbitrary pitch; Modeling neutron field parameters of power reactors; Wide range benchmarking possibilities, with high reproducibility of the benchmark design parameters; Wide range of measurement techniques including equipment and experienced personal; Flexible rearrangements of the core. The main experiments included: Pin wise flux distribution measurements; VVER-440 and VVER-1000 mock-ups; compact spent fuel storage; space kinetics experiment; core parameters experimental determination; experiment with new design fuel assembly; WWER-440 control assembly influence; and burnable absorber study. International research projects are also described. (P.A.)

  17. Automatically closing swing gate closure assembly

    Science.gov (United States)

    Chang, Shih-Chih; Schuck, William J.; Gilmore, Richard F.

    1988-01-01

    A swing gate closure assembly for nuclear reactor tipoff assembly wherein the swing gate is cammed open by a fuel element or spacer but is reliably closed at a desired closing rate primarily by hydraulic forces in the absence of a fuel charge.

  18. Hold-down device for nuclear reactor

    International Nuclear Information System (INIS)

    Leclercq, J.; Bonnamour, M.

    1984-01-01

    The present device can be used in nuclear reactors and more particularly in pressurized water reactors consisting of coupled fuel assemblies, certain of which are equipped with non-displaceable elements carried by an unsertable member. The device comprises the unsertable member provided with at least two sets of springs which transmit the load of an upper structure common to the fuel assemblies ajacent that which supports the unsertable member. The device is used to hold-down fuel assemblies which are subjected to the forces of circulating coolant [fr

  19. 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

  20. Method of operating FBR type reactors

    International Nuclear Information System (INIS)

    Arie, Kazuo.

    1984-01-01

    Purpose: To secure the controlling performance and the safety of FBR type reactors by decreasing the amount of deformation due to the difference in the heat expansion of a control rod guide tube. Method: The reactor is operated while disposing reactor core fuel assemblies of a same power at point-to-point symmetrical positions relative to the axial center for the control rod assembly. This can eliminate the temperature difference between opposing surfaces of the control rod guide tube and eliminate the difference in the thermal expansion. (Yoshino, Y.)

  1. 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)

  2. 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)

  3. The assembly and use of continuous flow systems for chemical synthesis.

    Science.gov (United States)

    Britton, Joshua; Jamison, Timothy F

    2017-11-01

    The adoption of and opportunities in continuous flow synthesis ('flow chemistry') have increased significantly over the past several years. Continuous flow systems provide improved reaction safety and accelerated reaction kinetics, and have synthesised several active pharmaceutical ingredients in automated reconfigurable systems. Although continuous flow platforms are commercially available, systems constructed 'in-lab' provide researchers with a flexible, versatile, and cost-effective alternative. Herein, we describe the assembly and use of a modular continuous flow apparatus from readily available and affordable parts in as little as 30 min. Once assembled, the synthesis of a sulfonamide by reacting 4-chlorobenzenesulfonyl chloride with dibenzylamine in a single reactor coil with an in-line quench is presented. This example reaction offers the opportunity to learn several important skills including reactor construction, charging of a back-pressure regulator, assembly of stainless-steel syringes, assembly of a continuous flow system with multiple junctions, and yield determination. From our extensive experience of single-step and multistep continuous flow synthesis, we also describe solutions to commonly encountered technical problems such as precipitation of solids ('clogging') and reactor failure. Following this protocol, a nonspecialist can assemble a continuous flow system from reactor coils, syringes, pumps, in-line liquid-liquid separators, drying columns, back-pressure regulators, static mixers, and packed-bed reactors.

  4. Subcritical assemblies, use and their feasibility assessment

    International Nuclear Information System (INIS)

    Haroon, M.R.

    1982-03-01

    In developing countries, subcritical assemblies can be a useful tool for training and research in the field of nuclear technology with minimum cost. The historical development of subcritical assemblies and the reactor physics experiments which can be carried out using this facility are outlined. The different types of subcritical assemblies have been described and material requirements for each assembly have been pointed out. (author)

  5. Preliminary shielding design evaluation for reactor assembly of SMART

    International Nuclear Information System (INIS)

    Kim, Kyo Youn; Kang, Chang M.; Kim, Ha Yong; Zee, Sung Quun; Chang, Moon Hee

    1999-03-01

    This report describes a preliminary evaluations of SMART shielding design near the reactor core by using the DORT two-dimensional discrete ordinates transport code. The results indicate that maximum neutron fluence at the bottom of reactor vessel is 1.64x10 17 n/cm 2 and that on the radial surface of reactor vessel is 6.71x10 16 n/cm 2 . These results meet the requirement, 1.0x10 20 n/cm 2 , in 10 CFR 50.61 and the integrity of SMART reactor vessel is confirmed during the lifetime of reactor. (Author). 20 refs., 11 tabs., 8 figs

  6. Experimental Studies on Assemblies 1 and 2 of the Fast Reactor FR-0. Part 2

    Energy Technology Data Exchange (ETDEWEB)

    Hellstrand, E; Andersson, T L; Brunfelter, B; Kockum, J; Londen, S O; Tiren, L I

    1965-12-15

    In a first part of this report, published as AE-195, an account was given of critical mass determinations and measurements of flux distribution and reaction ratios in the first assemblies of the fast zero power reactor FR0. This second part of the report deals with various investigations involving the measurement of reactivity. Control rod calibrations have been made using the positive period, the inverse multiplication, the rod drop and the pulsed source techniques, and show satisfactory agreement between the various methods. The reactivity worths of samples of different materials and different sizes have been measured at the core centre. Comparisons with perturbation calculations show that the regular and adjoint fluxes are well predicted in the central region of the core. The variation in the prompt neutron life-time with reactivity has been studied by means of the pulsed source and the Rossi-{alpha} techniques. Comparison with one region calculations reveals large discrepancies, indicating that this simple model is inadequate. Some investigations of streaming effects in an empty channel in the reactor and of interaction effects between channels have been made and are compared with theoretical estimates. Measurements of the reactivity worth of an air gap between the reactor halves and of the temperature coefficient are also described in the report. The work has been performed as a joint effort by AB Atomenergi and the Research Institute of National Defence.

  7. A swivelling transfer device for nuclear reactors

    International Nuclear Information System (INIS)

    Allain, Albert; Mulot, Pierre; Filloleau, Etienne

    1974-01-01

    The invention relates to a swivelling transfer device for fuel-assemblies. According to the invention, the device comprises, within a protective enclosure, a swivelling system comprising two sets of rails rotatable about an axis and so arranged that the lower and thereof penetrates into the extensions of the extremities of ramps dipped into the reactor and into a storage enclosure. This can apply to the transfer of nuclear reactor fuel assemblies, in particular for reactors of the molten sodium fast neutron type [fr

  8. Nuclear reactors

    International Nuclear Information System (INIS)

    Matheson, J.E.

    1983-01-01

    A nuclear reactor has an upper and a lower grid plate. Protrusions project from the upper grid plate. Fuel assemblies having end fittings fit between the grid plates. An arrangement is provided for accepting axial forces generated during the operation of the nuclear reactor by the flow of the cooling medium and thermal expansion and irradiation-induced growth of the fuel assembly, which comprises rods. Each fuel assembly rests on the lower grid plate and its upper end is elastically supported against the upper grid plate by the above-mentioned arrangement. The arrangement comprises four (for example) torsion springs each having a torsion tube and a torsion bar nested within the torsion tube and connected at one end thereto. The other end of the torsion bar is connected to an associated one of four lever arms. The torsion tube is rigidly connected to the other end fitting and the springs are disposed such that the lever arms are biassed against the protrusions. (author)

  9. 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.)

  10. Physics of pressurized water reactors

    International Nuclear Information System (INIS)

    Gruen, A.

    1980-01-01

    The objective of this lecture is to demonstrate typical problems and solutions encountered in the design and operation of PWR power plants. The examples selected for illustration refer to PWR's of KWU design and to results of KWU design methods. In order to understand the physics of a power reactor it is necessary to have some knowledge of the structure and design of the power plant system of which the reactor is a part. It is therefore assumed that the reader is familiar with the design of the more important components and systems of a PWR, such as fuel assemblies, control assemblies, core lay-out, reactor coolant system, instrumentation. (author)

  11. WWER safety investigations on LR-0 reactor

    International Nuclear Information System (INIS)

    Mikus, J.

    2001-01-01

    A set of the measurement needed for the WWER-440 and WWER-1000 reactor lifetime assessment, verification of the methods, codes and input cross section libraries for the WWER reactor pressure vessel exposure evaluation has been performed on the LR-0 experimental reactor. The WWER Mock-ups (engineering benchmarks) has been carried out on the reactor, with the aim to investigate differential neutron spectra for reactor dosimetry purposes. Critical experiments have also been performed to determine the perturbation of the fission density distribution caused by the WWER-440 control assembly. Such assembly, partially inserted in the core, has significant influence on the space power distribution. A wide research program for sub-criticality investigations of the spent nuclear fuel storage has been realized on the LR-0 reactor. A benchmark experiment is realized on the reactor in corresponding geometry for CASTOR 440/84 container for storage and transportation of spent fuel. Critical experiments with new fuel assemblies including various burnable absorbers and different enrichments are performed. A set of critical experiments is performed using the fuel assemblies with 3,6% and 4,4% enrichment, arranged in the WWER-440 type cores with various lattice pitch. The critical high of the moderator level and the moderator level coefficient of reactivity are measured and the effect of the fuel assembly, placed in a hexagonal tube of stainless steel containing boron absorber (ATABOR - STANDARD) is investigated. The obtained results are used for the validation of the codes (MCNP, KENO and SCALE) in the frame of the contract 'Burn-up credit implementation for the storage and transport containers of the spent fuel'. Combined neutron-gamma spectra measurements in the WWER-1000 Mock-up are carried out during 2001

  12. 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)

  13. 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.)

  14. Core construction in a pressure tube type heavy water reactor

    International Nuclear Information System (INIS)

    Ueda, Makoto; Aoki, Katsutada.

    1975-01-01

    Object: To replace a centrally positioned fuel assembly of a fuel assembly unit with a reactor controlling machinery to decrease a distance between the fuel assemblies thereby saving use of heavy water and enhancing economy. Structure: A centrally positioned fuel assembly of a fuel assembly unit, which is composed of a plurality of fuel assemblies orderly arranged in lattice fashion, is replaced with a reactor controlling members such as control rods, poison tubes and the like to provide an arrangement of lattice-free type fuel assembly, thus reducing the pitch as small as possible. (Kamimura, M.)

  15. Scientific programme and manufacture of types DK-1 and DK-2 diagnostic assemblies

    International Nuclear Information System (INIS)

    Krett, V.; Kott, J.; Vlcek, J.; Mlady, Z.

    1980-01-01

    The programme is described of measurements to be effected on the Rheinsberg WWER-2 reactor using diagnostic assemblies DK-1 and DK-2. The DK-1 assemblies were manufactured in the USSR and tested in the big water loop at SKODA Works. The insertion of the assemblies in the reactor is being prepared. The DK-2 assemblies are developed by SKODA Works in cooperation with the USSR, Hungary and Poland. (Ha)

  16. Three core concepts for producing uranium-233 in commercial pressurized light water reactors for possible use in water-cooled breeder reactors

    International Nuclear Information System (INIS)

    Conley, G.H.; Cowell, G.K.; Detrick, C.A.; Kusenko, J.; Johnson, E.G.; Dunyak, J.; Flanery, B.K.; Shinko, M.S.; Giffen, R.H.; Rampolla, D.S.

    1979-12-01

    Selected prebreeder core concepts are described which could be backfit into a reference light water reactor similar to current commercial reactors, and produce uranium-233 for use in water-cooled breeder reactors. The prebreeder concepts were selected on the basis of minimizing fuel system development and reactor changes required to permit a backfit. The fuel assemblies for the prebreeder core concepts discussed would occupy the same space envelope as those in the reference core but contain a 19 by 19 array of fuel rods instead of the reference 17 by 17 array. An instrument well and 28 guide tubes for control rods have been allocated to each prebreeder fuel assembly in a pattern similar to that for the reference fuel assemblies. Backfit of these prebreeder concepts into the reference reactor would require changes only to the upper core support structure while providing flexibility for alternatives in the type of fuel used

  17. Development of a capsule assembly machine for the re-irradiation tests in HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Y. H.; Choi, M. H.; Sohn, J. M.; Choo, K. N.; Cho, M. S.; Kim, B. G. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2008-11-15

    A capsule assembly machine (CAM) for the long term irradiation tests in the HANARO reactor has been designed, developed and demonstrated at the Korea Atomic Energy Reasearch Institute (KAERI). The CAM will provide a technical base for viable re-irradiation servives. This machine will be installed in the reactor service pool of the HANARO reactor. The new assembly technique by using a mockup of the CAM in air demonstrated its suitability for an assembly operation, and for an application of this technique to a reactor. The technique will be upgraded after a commissioning test under water environments. This would be expected to be recommended for a country where an under water canal for transporting irradiated devices and enough space of a hot cell for assembling capsule components are not available.

  18. Eddy current monitoring of spacers in coolant channel assemblies of nuclear reactor

    International Nuclear Information System (INIS)

    Bhole, V.M.; Rastogi, P.K.; Kulkarni, P.G.; Vijayaraghavan, R.

    1993-01-01

    An eddy current testing method has been standardised for monitoring spacer springs which are used in coolant channel assemblies of pressurised heavy water nuclear reactors (PHWRs). The standard bobbin coil probe used for monitoring the spacer spring detects only the location but does not monitor the tilt orientation and tilt angle of a tilted spacer spring. The knowledge of location along with the tilt orientation of the spacer spring greatly improves the performance of repositioning methods. A modified probe with angular windings has been developed in laboratory tests for monitoring the location as well as the tilt orientation of the spacer springs. Experimental results are presented showing excellent performance of the modified probe in monitoring the exact location as well as tilt orientation of a spacer spring. The modified probe has also been used successfully in the field during repositioning of spacer springs in PHWRs before commissioning. (Author)

  19. 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.)

  20. 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.)

  1. On detection of the possible use of VVERs for unreported production of plutonium. Final report for the period July 1988 - December 1989

    International Nuclear Information System (INIS)

    Simov, R.; Nelov, N.; Stoyanova, I.; Kovachev, N.; Yonchev, P.

    1989-01-01

    The study includes an analysis of the feasibility of unreported production of plutonium-239 in VVER-440 reactors. It is shown that for VVER-440 reactors 36 natural uranium oxide fuel assemblies in the peripheral region of the core need to be loaded to produce 8 kg of extra plutonium in one cycle. Substituting the peripheral fuel assemblies with natural uranium oxide fuel assemblies, the changes in the power peaking are negligible and do not affect reactor safety. Unreported production outside the core is not practical due to physical and mechanical constraints, low flux level, etc. The feasibility of unreported removal of irradiated material in spent fuel cask has been also assessed. After about a month cooling time, still within the refueling period, the irradiated natural uranium fuel assemblies could be removed off-site without significant health hazard to the workers. To improve the effectiveness of the safeguards objectives, additional inspection activities are suggested. 10 figs

  2. Fluid moderator control system reactor internals distribution system

    International Nuclear Information System (INIS)

    Fensterer, H.F.; Klassen, W.E.; Veronesi, L.; Boyle, D.E.; Salton, R.B.

    1987-01-01

    This patent describes a spectral shift pressurized water nuclear reactor employing a low neutron moderating fluid for the spectral shift including a reactor pressure vessel, a core comprising a plurality of fuel assemblies, a core support plate, apparatus comprising means for penetrating the reactor vessel for introducing the moderating fluid into the reactor vessel. Means associated with the core support plate for directly distributing the moderating fluid to and from the fuel assemblies comprises at least one inlet flow channel in the core plate; branch inlet feed lines connect to the inlet flow channel in the core plate; vertical inlet flow lines flow connected to the branch inlet feed lines; each vertical flow line communicates with a fuel assembly; the distribution means further comprise lines serving as return flow lines, each of which is connected to one of the fuel assemblies; branch exit flow lines in the core plate flow connected to the return flow lines of the fuel assembly; and at least one outlet flow channel flow connected to the branch exit flow lines; and a flow port interposed between the penetration means and the distribution means for flow connecting the penetration means with the distribution means

  3. Modelling of Control Bars in Calculations of Boiling Water Reactors

    International Nuclear Information System (INIS)

    Khlaifi, A.; Buiron, L.

    2004-01-01

    The core of a nuclear reactor is generally composed of a neat assemblies of fissile material from where neutrons were descended. In general, the energy of fission is extracted by a fluid serving to cool clusters. A reflector is arranged around the assemblies to reduce escaping of neutrons. This is made outside the reactor core. Different mechanisms of reactivity are generally necessary to control the chain reaction. Manoeuvring of Boiling Water Reactor takes place by controlling insertion of absorbent rods to various places of the core. If no blocked assembly calculations are known and mastered, blocked assembly neutronic calculation are delicate and often treated by case to case in present studies [1]. Answering the question how to model crossbar for the control of a boiling water reactor ? requires the choice of a representation level for every chain of variables, the physical model, and its representing equations, etc. The aim of this study is to select the best applicable parameter serving to calculate blocked assembly of a Boiling Water Reactor. This will be made through a range of representative configurations of these reactors and used absorbing environment, in order to illustrate strategies of modelling in the case of an industrial calculation. (authors)

  4. Structural characteristics of a graphite moderated critical assembly for a Zero Power reactor at IEA (Brazil)

    International Nuclear Information System (INIS)

    Almeida Ferreira, A.C. de; Hukai, R.Y.

    1975-01-01

    The structural characteristics of a graphite moderated core of a critical assembly to be installed in the Zero Power Reactor of IEA have been defined. These characteristics are the graphite block dimensions, the number and dimensions of the holes in the graphite, the pitch, the dimensions of the sticks of fuel and graphite to be inserted in the holes, and the mechanical reproducibility of the system. The composition of the fuel and moderator sticks were also defined. The main boundary conditions were the range of the relation C/U and C/TH used in commercial HTGR and the neutronics homogeneity

  5. Development of conceptual nuclear design of 10MWt research reactor core

    International Nuclear Information System (INIS)

    Kim, M. H.; Lim, J. Y.; Win, Naing; Park, J. M.

    2008-03-01

    KAERI has been devoted to develop export-oriented research reactors for a growing world-wide demand of new research reactor construction. Their ambition is that design of Korean research reactor must be competitive in commercial and technological based on the experience of the HANARO core design concept with thermal power of 30MW. They are developing a new research reactor named Advanced HANARO research Reactor (AHR) with thermal power of 20 MW. KAERI has export records of nuclear technology. In 1954-1967 two series of pool type research reactors based on the Russian design, VVR type and IRT type, have been constructed and commissioned in some countries as well as Russia. Nowadays Russian design is introducing again for export to developing countries such as Union of Myanmar. Therefore the objective of this research is that to build and innovative 10 MW research reactor core design based on the concept of HANARO core design to be competitive with Russian research reactor core design. system tool of HELIOS was used at the first stage in both cases which are research reactor using tubular type fuel assemblies and that reactor using pin type fuel assemblies. The reference core design of first kind of research reactor includes one in-core irradiation site at the core center. The neutron flux evaluations for core as well as reflector region were done through logical consistency of neutron flux distributions for individual assemblies. In order to find the optimum design, the parametric studies were carried out for assembly pitch, active fuel length, number of fuel ring in each assembly and so on. Design result shows the feasibility to have high neutron flux at in-core irradiation site. The second kind of research reactor is used the same kind of assemblies as HANARO and hence there is no optimization about basic design parameters. That core has only difference composition of assemblies and smaller specific power than HANARO. Since it is a reference core at first stage

  6. Nuclear reactor

    International Nuclear Information System (INIS)

    Hattori, Sadao; Sato, Morihiko.

    1994-01-01

    Liquid metals such as liquid metal sodium are filled in a reactor container as primary coolants. A plurality of reactor core containers are disposed in a row in the circumferential direction along with the inner circumferential wall of the reactor container. One or a plurality of intermediate coolers are disposed at the inside of an annular row of the reactor core containers. A reactor core constituted with fuel rods and control rods (module reactor core) is contained at the inside of each of the reactor core containers. Each of the intermediate coolers comprises a cylindrical intermediate cooling vessels. The intermediate cooling vessel comprises an intermediate heat exchanger for heat exchange of primary coolants and secondary coolants and recycling pumps for compulsorily recycling primary coolants at the inside thereof. Since a plurality of reactor core containers are thus assembled, a great reactor power can be attained. Further, the module reactor core contained in one reactor core vessel may be small sized, to facilitate the control for the reactor core operation. (I.N.)

  7. Method of operating water cooled reactor with blanket

    International Nuclear Information System (INIS)

    Suzuki, Katsuo.

    1988-01-01

    Purpose: To increase the production amount of fissionable plutonium by increasing the burnup degree of blanket fuels in a water cooled reactor with blanket. Method: Incore insertion assemblies comprising water elimination rods, fertile material rods or burnable poison rods are inserted to those fuel assemblies at the central portion of the reactor core that are situated at the positions not inserted with control rods in the earlier half of the operation cycle, while the incore reactor insertion assemblies are withdrawn at the latter half of the operation cycle of a nuclear reactor. As a result, it is possible to increase the power share of the blanket fuels and increase the fuel burnup degree to thereby increase the production amount of fissionable plutonium. Furthermore, at the initial stage of the cycle, the excess reactivity of the reactor can be suppressed to decrease the reactivity control share on the control rod. At the final stage of the cycle, the excess reactivity of the reactor core can be increased to improve the cycle life. (Kamimura, M.)

  8. 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

  9. 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

  10. Subcritical nuclear assembly

    Energy Technology Data Exchange (ETDEWEB)

    Vega C, H. R., E-mail: fermineutron@yahoo.com [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas (Mexico)

    2014-08-15

    A Subcritical Nuclear Assembly is a device where the nuclear-fission chain reaction is initiated and maintained using an external neutron source. It is a valuable educational and research tool where in a safe way many reactor parameters can be measured. Here, we have used the Wigner-Seitz method in the six-factor formula to calculate the effective multiplication factor of a subcritical nuclear reactor Nuclear Chicago model 9000. This reactor has approximately 2500 kg of natural uranium heterogeneously distributed in slugs. The reactor uses a {sup 239}PuBe neutron source that is located in the center of an hexagonal array. Using Monte Carlo methods, with the MCNP5 code, a three-dimensional model of the subcritical reactor was designed to estimate the effective multiplication factor, the neutron spectra, the total and thermal neutron fluences along the radial and axial axis. With the neutron spectra in two locations outside the reactor the ambient dose equivalent were estimated. (Author)

  11. Subcritical nuclear assembly

    International Nuclear Information System (INIS)

    Vega C, H. R.

    2014-08-01

    A Subcritical Nuclear Assembly is a device where the nuclear-fission chain reaction is initiated and maintained using an external neutron source. It is a valuable educational and research tool where in a safe way many reactor parameters can be measured. Here, we have used the Wigner-Seitz method in the six-factor formula to calculate the effective multiplication factor of a subcritical nuclear reactor Nuclear Chicago model 9000. This reactor has approximately 2500 kg of natural uranium heterogeneously distributed in slugs. The reactor uses a 239 PuBe neutron source that is located in the center of an hexagonal array. Using Monte Carlo methods, with the MCNP5 code, a three-dimensional model of the subcritical reactor was designed to estimate the effective multiplication factor, the neutron spectra, the total and thermal neutron fluences along the radial and axial axis. With the neutron spectra in two locations outside the reactor the ambient dose equivalent were estimated. (Author)

  12. Dismantling method for reactor shielding wall and device therefor

    International Nuclear Information System (INIS)

    Akagawa, Katsuhiko.

    1995-01-01

    A ring member having an outer diameter slightly smaller than an inner diameter of a reactor shielding wall to be dismantled is lowered in the inside of the reactor shielding wall while keeping a horizontal posture. A cutting device is disposed at the lower peripheral edge of the ring member. The cutting device can move along the peripheral edge of the circular shape of the ring member. The ring member is urged against the inner surface of the reactor shielding wall by using an urging member to immobilize the ring member. Then, the cutting device is operated to cut the reactor shielding wall into a plurality of ring-like blocks at a plurality of inner horizontal ribs or block connection ribs. Then, the blocks of the cut reactor shielding wall are supported by the ring member, and transported out of the reactor container by a lift. The cut blocks transported to the outside are finely dismantled for every block in a closed chamber. (I.N.)

  13. Simulation numerique de l'effet du reflecteur radial sur les cellules rep en utilisant les codes DRAGON et DONJON

    Science.gov (United States)

    Bejaoui, Najoua

    The pressurized water nuclear reactors (PWRs) is the largest fleet of nuclear reactors in operation around the world. Although these reactors have been studied extensively by designers and operators using efficient numerical methods, there are still some calculation weaknesses, given the geometric complexity of the core, still unresolved such as the analysis of the neutron flux's behavior at the core-reflector interface. The standard calculation scheme is a two steps process. In the first step, a detailed calculation at the assembly level with reflective boundary conditions, provides homogenized cross-sections for the assemblies, condensed to a reduced number of groups; this step is called the lattice calculation. The second step uses homogenized properties in each assemblies to calculate reactor properties at the core level. This step is called the full-core calculation or whole-core calculation. This decoupling of the two calculation steps is the origin of methodological bias particularly at the interface core reflector: the periodicity hypothesis used to calculate cross section librairies becomes less pertinent for assemblies that are adjacent to the reflector generally represented by these two models: thus the introduction of equivalent reflector or albedo matrices. The reflector helps to slowdown neutrons leaving the reactor and returning them to the core. This effect leads to two fission peaks in fuel assemblies localised at the core/reflector interface, the fission rate increasing due to the greater proportion of reentrant neutrons. This change in the neutron spectrum arises deep inside the fuel located on the outskirts of the core. To remedy this we simulated a peripheral assembly reflected with TMI-PWR reflector and developed an advanced calculation scheme that takes into account the environment of the peripheral assemblies and generate equivalent neutronic properties for the reflector. This scheme is tested on a core without control mechanisms and charged

  14. 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)

  15. 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.)

  16. Nuclear reactor instrumentation

    International Nuclear Information System (INIS)

    Duncombe, E.; McGonigal, G.

    1975-01-01

    A liquid metal cooled nuclear reactor is described which has an equal number of fuel sub-assemblies and sensing instruments. Each instrument senses temperature and rate of coolant flow of a coolant derived from a group of three sub-assemblies so that an abnormal value for one sub-assembly will be indicated on three instruments thereby providing for redundancy of up to two of the three instruments. The abnormal value may be a precurser to unstable boiling of coolant

  17. 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.

  18. In-core assembly configuration having a dual-wall pressure boundary for nuclear reactor

    International Nuclear Information System (INIS)

    Todt, W.H. Sr.; Playfoot, K.C.

    1988-01-01

    This patent describes an in-core detector assembly of the type having an in-core part and an out-of-core part and having an elongated outer hollow housing tube with a wall thickness, an inner hollow calibration tube with a wall thickness and disposed concentrically within the outer tube to define an annular space therewith, and a plurality of discrete, circular, rod-like elements extending through the annular space, the improvement comprising: the elements having outer diameters and being of a number to substantially occupy the entire annular space of both the incore and out-of-core parts without significant voids between elements; each of the elements including at least an outer sheath and interior highly compacted mineral insulation for the entire length of the element; a first number of the elements also including center lead means connected to condition responsive element means in the in-core part of the length of the assembly and a second, remaining number of the elements being non-operating elements. The wall thickness of the housing tube and the wall thickness of the calibration tube, taken together with the diameter of the elements, provide a thickness dimension adequate to meet code primary pressure requirements for normal nuclear reactor in-core conditions, while the wall thickness of the calibration tube alone provides a thickness dimension less than adequate to meet such requirements

  19. Nuclear reactor core stabilizing arrangement

    International Nuclear Information System (INIS)

    Jabsen, F.S.

    1976-01-01

    A nuclear reactor core stabilizing arrangement is described wherein a plurality of actuators, disposed in a pattern laterally surrounding a group of elongated fuel assemblies, press against respective contiguous fuel assemblies on the periphery of the group to reduce the clearance between adjacent fuel assemblies thereby forming a more compacted, vibration resistant core structure. 7 claims, 4 drawing figures

  20. High conversion Th-U{sup 233} fuel assembly for current generation of PWRs

    Energy Technology Data Exchange (ETDEWEB)

    Baldova, D.; Fridman, E. [Reactor Safety Div., Helmholtz-Zentrum Dresden-Rossendorf, POB 510119, Dresden, 01314 (Germany)

    2012-07-01

    This paper presents a preliminary design of a high conversion Th-U{sup 233} fuel assembly applicable for current generation of Pressurized Water Reactor (PWRs). The considered fuel assembly has a typical 17 x 17 PWR lattice. However in order to increase the conversion of Th{sup 232} to U{sup 233}, the assembly was subdivided into the two regions called seed and blanket. The central seed region has a higher than blanket U{sup 233} content and acts as a neutron source for the peripheral blanket region. The latest acts as a U{sup 233} breeder. While the seed fuel pins have a standard dimensions the blanket fuel radius was increased in order to reduce the moderation and to facilitate the resonance neutron absorption in blanket Th{sup 232}. The U{sup 233} content in the seed and blanket regions was optimized to achieve maximal initial to discharged fissile inventory ratio (FIR) taking into account the target fuel cycle length of 12 months with 3-batch reloading scheme. In this study the neutronic calculations were performed on the fuel assembly level using Helios deterministic lattice transport code. The fuel cycle length and the core k{sub eff} were estimated by applying the Non Linear Reactivity Model. The applicability of the HELIOS code for the analysis of the Th-based high conversion designs was confirmed with the help of continuous-energy Monte-Carlo code SERPENT. The results of optimization studies show that for the heterogeneous seed and blanket (SB) fuel assembly the FIR of about 0.95 can be achieved. (authors)

  1. Monitoring device for the power distribution within a nuclear reactor core

    International Nuclear Information System (INIS)

    Tanzawa, Tomio; Kumanomido, Hironori; Toyoshi, Isamu.

    1986-01-01

    Purpose: To provide a monitoring device for the power distribution in the reactor core that calculates the power distribution based on the measurement by instruments disposed within the reactor core of BWR type reactors. Constitution: The power distribution monitoring device in a reactor core comprises a signal correcting device, a signal normalizing device and a power distribution calculating device, in which the power distribution calculating device is constituted with an average power calculating device for four fuel assemblies and an average power calculating device for fuel assemblies. Gamma-ray signals corrected by the signal correcting device and signals from neutron detectors are inputted to the signal normalizing device, both of which are calibrated to determine the axial gamma-ray signal distribution in the central water gap region with the four fuel assemblies being as the unit. The average power from the four fuel assemblies are inputted to the fuel assembly average power calculating device to allocate to each of the fuel assembly average power thereby attaining the purpose. Further, thermal restriction values are calculated thereby enabling to secure the fuel integrity. (Kamimura, M.)

  2. Contribution to the WWER-440 Gd-assembly evolution

    International Nuclear Information System (INIS)

    Cudrnak, P.; Darilek, P.; Necas, V.

    2009-01-01

    Development activities, presented in the paper, contribute to the evolution of gadolinium fuel for the WWER-440 reactors. Short overview of existing and already proposed Gd-assemblies is given. Two new assembly types with improved features are described. Multiplication coefficients and pin power peaking factors of all mentioned gadolinium assemblies are compared. Animations of power distribution behaviour in selected Gd-assemblies are shown. (Authors)

  3. Corrosion of assemblies in fuel-storage basins at Savannah River Plant

    International Nuclear Information System (INIS)

    Wollam, C.D.

    1980-09-01

    Pitting of reactor assemblies has been the major corrosion problem in the Savannah River Plant fuel storage basins. From 1972 to 1976 many reactor assemblies experienced severe pitting corrosion with rates up to 9.3 mm/y. Poor cladding, high concentrations of iron and chloride ions in the water, a galvanic couple between the aluminum clad assemblies and the stainless steel hangers, and scratches in the oxide layer on assemblies have been identified as contributors to the problem. This paper describes the examinations and tests that were conducted and discusses a theory that explains the observed phenomena

  4. Self-lubricating fluid bearing assembly

    International Nuclear Information System (INIS)

    Kapich, D.D.

    1981-01-01

    A sealed self-lubricating fluid bearing assembly is described for circulating fluid in the form of a gas coolant in a nuclear reactor, the power for the circulator being provided by a shaft located within the primary containment vessel. In such a system the reactor coolant is isolated from the fluid region at the far end of the drive shaft. (U.K.)

  5. Automatic optimized reload and depletion method for a pressurized water reactor

    International Nuclear Information System (INIS)

    Ahn, D.H.; Levene, S.H.

    1985-01-01

    A new method has been developed to automatically reload and deplete a pressurized water reactor (PWR) so that both the enriched inventory requirements during the reactor cycle and the cost of reloading the core are minimized. This is achieved through four stepwise optimization calculations: (a) determination of the minimum fuel requirement for an equivalent three-region core model, (b) optimal selection and allocation of fuel assemblies for each of the three regions to minimize the reload cost, (c) optimal placement of fuel assemblies to conserve regionwise optimal conditions, and (d) optimal control through poison management to deplete individual fuel assemblies to maximize end-of-cycle k /SUB eff/ . The new method differs from previous methods in that the optimization process automatically performs all tasks required to reload and deplete a PWR. In addition, the previous work that developed optimization methods principally for the initial reactor cycle was modified to handle subsequent cycles with fuel assemblies having burnup at beginning of cycle. Application of the method to the fourth reactor cycle at Three Mile Island Unit 1 has shown that both the enrichment and the number of fresh reload fuel assemblies can be decreased and fully amortized fuel assemblies can be reused to minimize the fuel cost of the reactor

  6. Fast reactors - Dounreay and the future

    International Nuclear Information System (INIS)

    Jordan, G.

    1988-01-01

    In 1960 at Dounreay, the Dounreay Fast Reactor (DFR) supplied the world's first fast reactor grid electricity, and went on to a highly successful career as a test facility, as fuel designs advanced. In the 1960s, the Prototype Fast Reactor (PFR) was designed and built, beginning operation in 1974. The PFR was built to provide a sound technical and experienced base to support the UK's future Fast Reactor development and design. The in-vessel fuel handling facilities have demonstrated the flexibility of the pool design and a considerable body of in-core fuel handling experience is available. A key issue for further Fast Reactor application is the performance of fuel and, because PFR was designed to take full-scale fuel assemblies, the fuel performance experience is directly relevant to commercial designs. The original PFR design irradiation target of 60000 MWd/t U (equivalent to 7.5 % burn-up) has already been exceeded by a factor of more than two and a 15.9 % burn-up sub-assembly has been discharged and reprocessed without difficulty. Soon a 20 % sub-assembly will follow. Also the PFR reprocessing plant has demonstrated the safety and efficiency of this essential adjunct to Fast Reactor operation. The safety and the environmental protection features of both the PFR and its fuel reprocessing plant have been demonstrated over the last 14 years. 2 refs., 3 figs

  7. Hydrodynamic study of the adiabatic two-phase flow in the draught region of the VK-50 reactor

    International Nuclear Information System (INIS)

    Solodkij, V.A.; Bartolomej, G.G.; Fedulin, V.N.; Kharitonov, Yu.V.; Shmelev, V.E.; Abasov, A.V.

    1981-01-01

    Aimed at obtaining the spatial distribution of steam content PHI in the draught region (2.7 m high, equivalent diameter of 2 m) of the VK-50 experimental power boiling water reactor the local PHI values have been measured by electroprobing. The experiments were performed in a wide range of operating parameters (1.2-6.5 MPa and 10-180 MW (th)). Characteristic probe signal oscillograms and the axial and radial PHI distributions are presented. The local PHI values have occurred to oscillate in time, that proves structural inhomogeneity of the steemwater flow related to large scale turbulent pulsations. The effect is most prominent at low reactor pressures and powers. The effect of migration of the steam phase generated by peripheral fuel assemblies to the central part of the draught region has been observed up to the height of 0.8 m from the core outlet [ru

  8. 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)

  9. Development of ORIGEN libraries for mixed oxide (MOX) fuel assembly designs

    International Nuclear Information System (INIS)

    Mertyurek, Ugur; Gauld, Ian C.

    2016-01-01

    Highlights: • ORIGEN MOX library generation process is described. • SCALE burnup calculations are validated against measured MOX fuel samples from the MALIBU program. • ORIGEN MOX libraries are verified using the OECD Phase IV-B benchmark. • There is good agreement for calculated-to-measured isotopic distributions. - Abstract: ORIGEN cross section libraries for reactor-grade mixed oxide (MOX) fuel assembly designs have been developed to provide fast and accurate depletion calculations to predict nuclide inventories, radiation sources and thermal decay heat information needed in safety evaluations and safeguards verification measurements of spent nuclear fuel. These ORIGEN libraries are generated using two-dimensional lattice physics assembly models that include enrichment zoning and cross section data based on ENDF/B-VII.0 evaluations. Using the SCALE depletion sequence, burnup-dependent cross sections are created for selected commercial reactor assembly designs and a representative range of reactor operating conditions, fuel enrichments, and fuel burnup. The burnup dependent cross sections are then interpolated to provide problem-dependent cross sections for ORIGEN, avoiding the need for time-consuming lattice physics calculations. The ORIGEN libraries for MOX assembly designs are validated against destructive radiochemical assay measurements of MOX fuel from the MALIBU international experimental program. This program included measurements of MOX fuel from a 15 × 15 pressurized water reactor assembly and a 9 × 9 boiling water reactor assembly. The ORIGEN MOX libraries are also compared against detailed assembly calculations from the Phase IV-B numerical MOX fuel burnup credit benchmark coordinated by the Nuclear Energy Agency within the Organization for Economic Cooperation and Development. The nuclide compositions calculated by ORIGEN using the MOX libraries are shown to be in good agreement with other physics codes and with experimental data.

  10. 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.)

  11. 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)

  12. Calibration assembly for nuclear reactor vessel inspection apparatus

    International Nuclear Information System (INIS)

    Elsner, H.J.

    1981-01-01

    A removable calibration assembly can be utilized to verify the angular mounting of transducers in an array employed in an ultrasonic inpsection apparatus, to calibrate one axis of movement of the array with reference to a starting point, or to measure and calibrate the speed per unit of distance of the transducer's ultrasonic beam in the operating medium. The calibration assembly includes both a large and small reflecting surface separated by known distances, and several large cones, the tips of which are machined or adjusted to angles at which certain of the transducers are to be mounted. Clamping means for securing the calibration assembly to the inspection apparatus at a predetermined orientation is provided

  13. Control rod guide tube assembly

    International Nuclear Information System (INIS)

    Jabsen, F.S.

    1982-01-01

    An improved fuel assembly is described as consisting of a sleeve that engages one end of a control rod guide tube essentially fixing the guide tube to one of the fuel assembly end structures. The end of the sleeve protrudes above the surface of the end fitting. The outer surface of the sleeve has a peripheral groove that engages the resilient sides of a cellular grid or lattice shaped lock. This lock fixes the sleeve in position between the various elements that comprise the end fitting, thereby eliminating a profusion of costly and potentially troublesome nuts, threaded studs and the like that are frequently employed in the fuel assemblies that are presently in use

  14. Liquid metal cooled nuclear reactor

    International Nuclear Information System (INIS)

    Leigh, K.M.

    1980-01-01

    A liquid metal cooled nuclear reactor is described, wherein coolant is arranged to be flowed upwardly through a fuel assembly and having one or more baffles located above the coolant exit of the fuel assembly, the baffles being arranged so as to convert the upwardly directed motion of liquid metal coolant leaving the fuel assembly into a substantially horizontal motion. (author)

  15. Disposition of weapons-grade plutonium in Westinghouse reactors

    International Nuclear Information System (INIS)

    Alsaed, A.A.; Adams, M.

    1998-03-01

    The authors have studied the feasibility of using weapons-grade plutonium in the form of mixed-oxide (MOX) fuel in existing Westinghouse reactors. They have designed three transition Cycles from an all LEU core to a partial MOX core. They found that four-loop Westinghouse reactors such as the Vogtle power plant are capable of handling up to 45 percent weapons-grade MOX loading without any modifications. The authors have also designed two kinds of weapons-grade MOX assemblies with three enrichments per assembly and four total enrichments. Wet annular burnable absorber (WABA) rods were used in all the MOX feed assemblies, some burned MOX assemblies, and some LEU feed assemblies. Integral fuel burnable absorber (IFBA) was used in the rest of the LEU feed assemblies. The average discharge burnup of MOX assemblies was over 47,000 MWD/MTM, which is more than enough to meet the open-quotes spent fuel standard.close quotes One unit is capable of consuming 0.462 MT of weapons-grade plutonium per year. Preliminary analyses showed that important reactor physics parameters for the three transitions cycles are comparable to those of LEU cores including boron levels, reactivity coefficients, peaking factors, and shutdown margins. Further transient analyses will need to be performed

  16. Reactor Simulations for Safeguards with the MCNP Utility for Reactor Evolution Code

    International Nuclear Information System (INIS)

    Shiba, T.; Fallot, M.

    2015-01-01

    To tackle nuclear material proliferation, we conducted several proliferation scenarios using the MURE (MCNP Utility for Reactor Evolution) code. The MURE code, developed by CNRS laboratories, is a precision, open-source code written in C++ that automates the preparation and computation of successive MCNP (Monte Carlo N-Particle) calculations and solves the Bateman equations in between, for burnup or thermal-hydraulics purposes. In addition, MURE has been completed recently with a module for the CHaracterization of Radioactive Sources, called CHARS, which computes the emitted gamma, beta and alpha rays associated to any fuel composition. Reactor simulations could allow knowing how plutonium or other material generation evolves inside reactors in terms of time and amount. The MURE code is appropriate for this purpose and can also provide knowledge on associated particle emissions. Using MURE, we have both developed a cell simulation of a typical CANDU reactor and a detailed model of light water PWR core, which could be used to analyze the composition of fuel assemblies as a function of time or burnup. MURE is also able to provide, thanks to its extension MURE-CHARTS, the emitted gamma rays from fuel assemblies unloaded from the core at any burnup. Diversion cases of Generation IV reactors have been also developed; a design of Very High Temperature Reactor (a Pebble Bed Reactor (PBR), loaded with UOx, PuOx and ThUOx fuels), and a Na-cooled Fast Breeder Reactor (FBR) (with depleted Uranium or Minor Actinides in the blanket). The loading of Protected Plutonium Production (P3) in the FBR was simulated. The simulations of various reactor designs taking into account reactor physics constraints may bring valuable information to inspectors. At this symposium, we propose to show the results of these reactor simulations as examples of the potentiality of reactor simulations for safeguards. (author)

  17. 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

  18. 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)

  19. Nuclear reactor instrumentation

    International Nuclear Information System (INIS)

    Duncombe, E.; McGonigal, G.

    1976-01-01

    Reference is made to the instrumentation of liquid metal cooled fast reactors. In order to ensure the safe operation of such reactors it is necessary to constantly monitor the coolant flowing through the fuel assemblies for temperature and rate of flow, requiring a large number of sensors. An improved and simplified arrangement is claimed in which the fuel assemblies feed a fraction of coolant to three instrument units arranged to sense the temperature and rate of flow of samples of coolant. Each instrument unit comprises a sleeve housing a sensing unit and has a number of inlet ducts arranged for receiving coolant from a fuel assembly together with a single outlet. The sensing unit has three thermocouple hot junctions connected in series, the hot junctions and inlet ducts being arranged in pairs. Electromagnetic windings around an inductive core are arranged to sense variation in flow of liquid metal by flux distortion. Fission product sensing means may also be provided. Full constructional details are given. (U.K.)

  20. 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)

  1. NABUB a non-saturated model of coolant boiling in a fast reactor sub-assembly

    International Nuclear Information System (INIS)

    Brook, A.J.; Mills, D.S.

    1975-08-01

    A theoretical model is described of sodium boiling in a fast reactor sub-assembly in which the usual assumptions of a saturated vapour are not made. Instead, vapour pressure is calculated in a perfect gas basis, which enables some allowance to be made for the possible presence of non-condensables, which may inhibit the condensation f the vapour. Indications are given of the circumstances under which such inhibition might be expected to show the most marked effects, and some sample results ontained by the code are presented. These show that the coolant voiding pattern is most sensitive to restrictions on the condensing flux in the 100 to 200w/cm 2 range. If unrestricted condensation is assumed, the results of the code are in excellent agreement with more conventional saturation models. (author)

  2. Liquid metal cooled fast breeder nuclear reactor

    International Nuclear Information System (INIS)

    Scott, D.

    1979-01-01

    A liquid metal cooled fast breeder nuclear reactor has a core comprising a plurality of fuel assemblies supported on a diagrid and submerged in a pool of liquid metal coolant within a containment vessel, the diagrid being of triple component construction and formed of a short cylindrical plenum mounted on a conical undershell and loosely embraced by a fuel store carrier. The plenum merely distributes coolant through the fuel assemblies, the load of the assemblies being carried by the undershell by means of struts which penetrate the plenum. The reactor core, fuel store carrier and undershell provide secondary containment for the plenum. (UK)

  3. Buildup of radioxenon isotopes in MOX-assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Gniffke, Thomas; Kirchner, Gerald [Carl Friedrich von Weizsaecker-Centre for Science and Peace Research, Hamburg (Germany)

    2015-07-01

    Radioxenon is the main tracer for detection of nuclear tests conducted underground under the verification regime of the Comprehensive Nuclear Test Ban Treaty (CTBT). Since radioxenon is emitted by civilian sources too, like commercial nuclear reactors, source discrimination is still an important issue. Inventory calculations are necessary to predict which xenon isotopic ratios are built up in a reactor and how they differ from those generated by a nuclear explosion. The screening line actually used by the CTBT Organization for source discrimination is based on calculations for uranium fuel of various enrichments used in pressurized water reactors (PWRs). The usage of different fuel, especially mixed U/Pu oxide (MOX) assemblies with reprocessed plutonium, may alter the radioxenon signature of civilian reactors. In this talk, calculations of the radioxenon buildup in a MOX-assembly used in a commercial PWR are presented. Implications for the CTBT verification regimes are discussed and open questions are addressed.

  4. 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)

  5. 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.

  6. Axial power monitoring uncertainty in the Savannah River Reactors

    International Nuclear Information System (INIS)

    Losey, D.C.; Revolinski, S.M.

    1990-01-01

    The results of this analysis quantified the uncertainty associated with monitoring the Axial Power Shape (APS) in the Savannah River Reactors. Thermocouples at each assembly flow exit map the radial power distribution and are the primary means of monitoring power in these reactors. The remaining uncertainty in power monitoring is associated with the relative axial power distribution. The APS is monitored by seven sensors that respond to power on each of nine vertical Axial Power Monitor (APM) rods. Computation of the APS uncertainty, for the reactor power limits analysis, started with a large database of APM rod measurements spanning several years of reactor operation. A computer algorithm was used to randomly select a sample of APSs which were input to a code. This code modeled the thermal-hydraulic performance of a single fuel assembly during a design basis Loss-of Coolant Accident. The assembly power limit at Onset of Significant Voiding was computed for each APS. The output was a distribution of expected assembly power limits that was adjusted to account for the biases caused by instrumentation error and by measuring 7 points rather than a continuous APS. Statistical analysis of the final assembly power limit distribution showed that reducing reactor power by approximately 3% was sufficient to account for APS variation. This data confirmed expectations that the assembly exit thermocouples provide all information needed for monitoring core power. The computational analysis results also quantified the contribution to power limits of the various uncertainties such as instrumentation error

  7. Device for protecting deformations of reactor cores

    International Nuclear Information System (INIS)

    Kato, Yasuyoshi; Urushihara, Hiroshi.

    1975-01-01

    Object: To provide a fluid pressure cylinder, which is operated according to change in temperature of coolant for a reactor to restrain or release a core, to simply and effectively protect deformation of the core. Structure: A closed fluid pressure cylinder interiorly filled with suitable fluid is disposed in peripherally equally spaced relation in an annular space between a core barrel of a reactor and a reactor vessel. A piston is mounted in fluid-tight fashion in a plurality of piston openings made in the cylinder, the piston being slidably moved according to expansion and contraction of the fluid filled in the cylinder. The piston has a movable frame mounted at the foremost end thereof, the movable frame being moved integral with the piston, and the surface opposite the mount thereof biasing the outermost peripheral surface of the core. (Kamimura, M.)

  8. Seals in nuclear reactors

    International Nuclear Information System (INIS)

    1979-01-01

    The aim of this invention is the provision of improved seals for reactor vessels in which fuel assemblies are located together with inlets and outlets for the circulation of a coolant. The object is to provide a seal arrangement for the rotatable plugs of nuclear reactor closure heads which has good sealing capacities over a wide gap during operation of the reactor but which also permits uninhibited rotation of the plugs for maintenance. (U.K.)

  9. Liquid metal cooled fast breeder nuclear reactors

    International Nuclear Information System (INIS)

    Barnes, S.

    1976-01-01

    Reference is made to liquid metal cooled fast breeder reactors of the 'pool' kind. In this type of reactor the irradiated fuel is lowered into a transfer rotor for removal to storage facilities, this rotor normally having provision for the temporary storage of 20 irradiated fuel assemblies, each within a stainless steel bucket. For insertion or withdrawal of a fuel assembly the rotor is rotated to bring the fuel assembly to a loading or discharging station. The irradiated fuel assembly is withdrawn from the rotor within its bucket and the total weight is approximately 1000 kg, which is lifted about 27 m. In the event of malfunction the combination falls back into the rotor with considerable force. In order to prevent damage to the rotor fracture pins are provided, and to prevent damage to the reactor vessel and other parts of the reactor structure deformable energy absorbing devices are provided. After a malfunction the fractured pins and the energy absorbing devices must be replaced by remote control means operated from outside the reactor vault - a complex operation. The object of the arrangement described is to provide improved energy absorbing means for fuel assemblies falling into a fuel transfer rotor. The fuel assemblies are supported in the rotor by elastic means during transfer to storage and a hydraulic dash pot is provided in at least one position below the rotor for absorbing the energy of a falling fuel assembly. It is preferable to provide dash pots immediately below a receiving station for irradiated fuel assemblies and immediately below a discharge station. Each bucket is carried in a container that is elastically supported in the transfer rotor on a helical coil compression spring, so that, in the event of a malfunction the container and bucket are returned to their normal operating position after the force of the falling load has been absorbed by the dash pot. The transfer rotor may also be provided with recoil springs to absorb the recoil energy

  10. BWR type reactor core

    International Nuclear Information System (INIS)

    Tatemichi, Shin-ichiro.

    1981-01-01

    Purpose: To eliminate the variation in the power distribution of a BWR type reactor core in the axial direction even if the flow rate is increased or decreased by providing a difference in the void coefficient between the upper part and the lower parts of the reactor core, and increasing the void coefficient at the lower part of the reactor core. Constitution: The void coefficient of the lower region from the center to the lower part along the axial direction of a nuclear fuel assembly is increased to decrease the dependence on the flow rate of the axial power distribution of the nuclear fuel assembly. That is, a water/fuel ratio is varied, the water in non-boiled region is increased or the neutron spectrum is varied so as to vary the void coefficient. In order to exemplify it, the rate of the internal pellets of the fuel rod of the nuclear fuel assembly or the shape of the channel box is varied. Accordingly, the power does not considerably vary even if the flow rate is altered since the power is varied in the power operation. (Yoshihara, H.)

  11. 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)

  12. Applications: fission, nuclear reactors. Fission: the various ways for reactors and cycles

    International Nuclear Information System (INIS)

    Bacher, P.

    1997-01-01

    A historical review is presented concerning the various nuclear reactor systems developed in France by the CEA: the UNGG (graphite-gas) system with higher CO 2 pressures, bigger fuel assemblies and powers higher than 500 MW e, allowed by studies on reactor physics, cladding material developments and reactor optimization; the fast neutron reactor system, following the graphite-gas development, led to the Superphenix reactor and important progress in simulation based on experiment and return of experience; and the PWR system, based on the american license, which has been successfully accommodated to the french industry and generates up to 75% of the electric power in France

  13. 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

  14. Flow measurement by Laser Doppler Anemometry in a nuclear fuel assembly

    International Nuclear Information System (INIS)

    Kehoe, A.

    1984-12-01

    Development of a Laser Doppler Anemometer measurement system and its operation are examined in this research. The system is designed for flow measurement in laboratory models of nuclear fuel assemblies. Use of the system is demonstrated by measuring turbulent velocity profiles in the laboratory model at full scale reactor flow rates. The reactors at the Savanah River Plant (SRP) are heavy water moderated and operate at low temperatures and pressures. Reactor power is currently limited by the temperature of the water in the nuclear fuel assembly. These temperature limits are conservatively calculated without allowing for any turbulent mixing. This research incorporates the design, fabriction and operation of a plexiglas model fuel assembly for the purpose of making turbulent velocity measurement via a Laser Doppler Anemometer System

  15. 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)

  16. Fast reactor core monitoring device

    International Nuclear Information System (INIS)

    Sanda, Toshio; Inoue, Kotaro; Azekura, Kazuo.

    1982-01-01

    Purpose: To enable the rapid and accurate on-line identification of the state of a fast reactor core by effectively utilizing the measured data on the temperature and flow rate of the coolant. Constitution: The spacial power distribution and average assembly power are quickly calculated using an approximate calculating method, the measured values and the calculated values of the inlet and outlet temperature difference, flow rate and coolant physical values of an assembly are combined and are individually obtained, the most definite respective values and their errors are obtained by a least square method utilizing a formula of the relation between these values, and the power distribution and the temperature distribution of a reactor core are estimated in this manner. Accordingly, even when the measuring accuracy and the calculating accuracy are equal as in a fast reactor, the power distribution and the temperature distribution can be accurately estimated on-line at a high speed in a nuclear reactor, information required for the operator is provided, and the reactor can thus be safely and efficiently operated. (Yoshihara, H.)

  17. Improvements in or relating to nuclear reactors

    International Nuclear Information System (INIS)

    Savin, N.I.; Khramov, D.A.; Filippov, V.J.; Bugrov, V.V.

    1979-01-01

    A nuclear reactor is described, comprising a core accommodating a plurality of fuel assemblies and a refuelling device for replacing spent fuel assemblies. The design of the fuel assembly and of the refuelling device, and the method of carrying out the refuelling operation, are specified. (U.K.)

  18. Nuclear reactor power supply system

    International Nuclear Information System (INIS)

    Cook, B.M.

    1982-01-01

    The redundant signals from the sensor assemblies measuring the process parameters of a nuclear reactor power supply are transmitted each in its turn to a protection system which operates to actuate the protection apparatus for signals indicating off-process conditions. Each sensor assembly includes a number of like sensors measuring the same parameters. The sets of process signals derived from the sensor assemblies are each in its turn transmitted from the protection system to the control system which impresses control signals on the reactor or its components to counteract the tendency for conditions to drift off-normal status requiring operation of the protection system. A parameter signal selector prevents a parameter signal which differs from the other parameter signals of the set by more than twice the allowable variation from passing to the control system. Test signals are periodically impressed by a test unit on a selected pair of a selection unit and control channels. This arrangement eliminates the possibility that a single component failure which may be spurious will cause an inadvertent trip of the reactor during test. (author)

  19. 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

  20. Emergency cooling system for nuclear reactors

    International Nuclear Information System (INIS)

    Frisch, E.; Andrews, H.N.

    1976-01-01

    Upon the occasion of loss of coolant in a nuclear reactor as when a coolant supply or return line breaks, or both lines break, borated liquid coolant from an emergency source is supplied in an amount to absorb heat being generated in the reactor even after the control rods have been inserted. The liquid coolant flows from pressurized storage vessels outside the reactor to an internal manifold from which it is distributed to unused control rod guide thimbles in the reactor fuel assemblies. Since the guide thimbles are mounted at predetermined positions relative to heat generating fuel elements in the fuel assemblies, holes bored at selected locations in the guide thimble walls, sprays the coolant against the reactor fuel elements which continue to dissipate heat but at a reduced level. The cooling water evaporates upon contacting the fuel rods thereby removing the maximum amount of heat (970 BTU per pound of water) and after heat absorption will leave the reactor in the form of steam through the break which is the cause of the accident to help assure immediate core cooldown