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)

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.)

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.)

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

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)

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.)

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.)

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.)

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

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

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

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

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)

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)

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

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

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.)

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

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.)

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)

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.)

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)

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

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

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

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)

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.)

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

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

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)

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)

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

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.

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

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.)

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

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

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.

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.)

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

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.)

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

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

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)

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)

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)

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)

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.)

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.)

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

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

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.

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

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)

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

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

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.)

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

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.)

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)

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

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

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

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

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

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)

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

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.)

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.)

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.)

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)

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

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

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

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)

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.)

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

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.)

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

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. )

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.)

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.)

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)

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

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

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

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.

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.)

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.)

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.)

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

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.)

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.)

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.)

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)

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)

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.

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

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

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

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.)

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.)

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.)

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)

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

International Nuclear Information System (INIS)

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

1997-01-01

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

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.)

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)

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

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)

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

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.

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

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

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

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.

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)

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)

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)

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.)

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)

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)

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)

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

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.)

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

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

Reactor fueling system

International Nuclear Information System (INIS)

Hattori, Noriaki; Hirano, Haruyoshi.

1983-01-01

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

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.

Fuel assembly

International Nuclear Information System (INIS)

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

1998-01-01

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

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.

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.

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

International Nuclear Information System (INIS)

Memmott, Matthew; Buongiorno, Jacopo; Hejzlar, Pavel

2009-01-01

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

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

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

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

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

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

International Nuclear Information System (INIS)

Krause, F.

1986-01-01

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

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.)

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

Fuel Assembly Damping Summary

Energy Technology Data Exchange (ETDEWEB)

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

2013-10-15

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

Fuel sub-assembly

International Nuclear Information System (INIS)

Jolly, R.

1982-01-01

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

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)

Nuclear fuel assembly seismic amplitude limiter

International Nuclear Information System (INIS)

Anthony, A.J.

1977-01-01

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

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)

Impact Analysis for Fuel Assemblies in Spent Fuel Storage Rack

International Nuclear Information System (INIS)

Oh, Jinho

2013-01-01

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

Impact Analysis for Fuel Assemblies in Spent Fuel Storage Rack

Energy Technology Data Exchange (ETDEWEB)

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

2013-07-01

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

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)

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)

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

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

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

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)

Fuel assembly

International Nuclear Information System (INIS)

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

1995-01-01

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

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.)

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

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

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

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

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.

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)

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)

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)

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)

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)

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

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)

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

Nuclear reactor fuel elements

International Nuclear Information System (INIS)

Hindle, E.D.

1981-01-01

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

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)

Experimental studies of resistance fretting-wear of fuel rods for VVER-1000 and TVS-KVADRAT fuel assemblies

International Nuclear Information System (INIS)

Makarov, V.; Afanasiev, A.; Egorov, Yu.; Matvienko, I.

2015-01-01

The paper covers the results of the studies performed to justify the wear resistance of fuel rods in contact with the spacer grids of TVS VVER-1000 fuel assembly and TVS-KVADRAT square fuel assembly of Russian design for PWR-900 reactor. The presented results of three testing stages comprise: Testing of mockup fuel rods of VVER TVS fuel assembly for fretting wear under the conditions of the water chemistry of VVER reactor; Testing models of different design embodiments of the fuel rods for VVER TVS fuel assembly for fretting wear in still cold water; Testing mockup fuel rods of TVS-KVADRAT square fuel assembly for PWR reactor for frettingwear under the conditions of PWR water chemistry. The effect of structural and operational factors was determined (amplitudes, fuel rod vibration frequencies, values of cladding-to-spacer grid cell gap for the depth of fuel rod cladding wear etc.), an assessment was made of the threshold values of fuel rod vibration parameters, which, if not exceeded, provide the absence of the fuel rod cladding fretting wear in the fuel rod-to spacer grid contact area. Key words: fretting wear, fuel rod, spacer grid, VVER, PWR (author)

The fuel of nuclear reactors

International Nuclear Information System (INIS)

1995-03-01

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

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

Calibration of spent fuel measurement assembly

International Nuclear Information System (INIS)

Koleska, Michal; Viererbl, Ladislav; Marek, Milan

2014-01-01

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

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

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...

Bimetallic spacer means for a nuclear fuel assembly

International Nuclear Information System (INIS)

Anthony, A.J.

1981-01-01

A bimetallic spacer means designed to be cooperatively associated with a nuclear fuel assembly and operative to resist the occurrence of in-reactor bowing of the nuclear fuel assembly. The subject bimetallic spacer means in accord with one embodiment of the invention includes a member formed, at least principally, of Zircaloy to which are attached a plurality of stainless steel strips. The latter stainless steel strips are located on the external surface of the Zircaloy member and with the major axis of each of the plurality of stainless steel strips extending substantially perpendicular to the major axis of the Zircaloy member. In accord with another embodiment of the invention, the subject bimetallic spacer means includes a member formed at least principally of Zircaloy to which a plurality of stainless steel strips are attached so as to be positioned thereon externally thereof and with the major axis of each of the plurality of stainless steel strips extending substantially parallel to the major axis of the Zircaloy member. In accord with a further embodiment of the invention, the stainless steel strips are attached to preselected members, each embodying at least a cladding of Zircaloy, which are located in the rows of fuel rods that define the perimeter of the fuel matrix of the nuclear fuel assembly. In each of the embodiments, the stainless steel strips during power production expand outwardly to a greater extent than do the members to which the stainless steel strips are attached, thereby forming stiff springs which abut against like bimetallic spacer means with which the other nuclear fuel assemblies are provided in a given nuclear reactor core to thus prevent the occurrence of in-reactor bowing of the nuclear fuel assemblies. Namely, the stainless steel strips expand laterally relative to the fuel assembly and thereby occupy the space adjacent to the external surface of the fuel assembly

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

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

Nuclear fuel assembly

International Nuclear Information System (INIS)

1975-01-01

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

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)

Nuclear fuel assembly with improved spectral shift-producing rods

International Nuclear Information System (INIS)

Ferrari, H.M.

1987-01-01

This patent describes a nuclear reactor having fuel assemblies and a moderator-coolant liquid flowing through the fuel assemblies, each fuel assembly including an organized array of nuclear fuel rods wherein the moderator-coolant liquid flows along the fuel rods, at least one improved spectral shift-producing rod disposed among the fuel rods. The spectra shift-producing rod consists of: (a) an elongated hollow hermetically-sealed tubular member; (b) a weakened region formed in a portion of the member, the portion being subject to rupture at a given level of internal pressure; and (c) burnable poison material contained in the member which generates gas in the member as operation of the reactor proceeds normally, the material being soluble in the moderator-coolant liquid when brought into contact therewith; (d) the given level of internal pressure being less than the maximum level of internal pressure normally expected to be generated within the member by the poison material by normal operation of the reactor

Fuel assembly

International Nuclear Information System (INIS)

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

1998-01-01

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

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

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

Overview of neutronic fuel assembly design and in-core fuel management

International Nuclear Information System (INIS)

Porsch, D.; Charlier, A.; Meier, G.; Mougniot, J.C.; Tsuda, K.

2000-01-01

The civil and military utilization of nuclear power results in stockpiles of spent fuel and separated plutonium. Recycling of the recovered plutonium in Light Water Reactors (LWR) is currently practiced in Belgium, France, Germany, and Switzerland, in Japan it is in preparation. Modern MOX fuel, with its optimized irradiation and reprocessing behavior, was introduced in 1981. Since then, about 1700 MOX fuel assemblies of different mechanical and neutronic design were irradiated in commercial LWRs and reached fuel assembly averaged exposures of up to 51.000 MWd/t HM. MOX fuel assemblies reloaded in PWR have an average fissile plutonium content of up to 4.8 w/o. For BWR, the average fissile plutonium content in actual reloads is 3.0 w/o. Targets for the MOX fuel assembly design are the compatibility to uranium fuel assemblies with respect to their mechanical fuel rod and fuel assembly design, they should have no impact on the flexibility of the reactor operation, and its reload should be economically feasible. In either cycle independent safety analyses or individually for each designed core it has to be demonstrated that recycling cores meet the same safety criteria as uranium cores. The safety criteria are determined for normal operation and for operational as well as design basis transients. Experience with realized MOX core loadings confirms the reliability of the applied modern design codes. Studies for reloads of advanced MOX assemblies in LWRs demonstrate the feasibility of a future development of the thermal plutonium recycling. New concepts for the utilization of plutonium are under consideration and reveal an attractive potential for further developments on the plutonium exploitation sector. (author)

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)

Advanced fuel in the Budapest research reactor

International Nuclear Information System (INIS)

Hargitai, T.; Vidovsky, I.

1997-01-01

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

Unification of fuel elements for research reactors

International Nuclear Information System (INIS)

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

1997-01-01

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

Fuel assembly

International Nuclear Information System (INIS)

Bando, Masaru.

1993-01-01

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

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.)

Gaseous fuel reactors for power systems

Science.gov (United States)

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

1977-01-01

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

Appearance detection device for fuel assembly

International Nuclear Information System (INIS)

Matsuoka, Toshihiro

1998-01-01

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

Nuclear fuel assembly

International Nuclear Information System (INIS)

Ito, Arata; Wakamatsu, Mitsuo.

1976-01-01

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

Fuel assembly

International Nuclear Information System (INIS)

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

1998-01-01

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

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

International Nuclear Information System (INIS)

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

2008-06-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2008-06-15

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

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.)

Spent nuclear fuel discharges from U.S. reactors 1994

International Nuclear Information System (INIS)

1996-02-01

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

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)

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)

Benchmark problem suite for reactor physics study of LWR next generation fuels

International Nuclear Information System (INIS)

Yamamoto, Akio; Ikehara, Tadashi; Ito, Takuya; Saji, Etsuro

2002-01-01

This paper proposes a benchmark problem suite for studying the physics of next-generation fuels of light water reactors. The target discharge burnup of the next-generation fuel was set to 70 GWd/t considering the increasing trend in discharge burnup of light water reactor fuels. The UO 2 and MOX fuels are included in the benchmark specifications. The benchmark problem consists of three different geometries: fuel pin cell, PWR fuel assembly and BWR fuel assembly. In the pin cell problem, detailed nuclear characteristics such as burnup dependence of nuclide-wise reactivity were included in the required calculation results to facilitate the study of reactor physics. In the assembly benchmark problems, important parameters for in-core fuel management such as local peaking factors and reactivity coefficients were included in the required results. The benchmark problems provide comprehensive test problems for next-generation light water reactor fuels with extended high burnup. Furthermore, since the pin cell, the PWR assembly and the BWR assembly problems are independent, analyses of the entire benchmark suite is not necessary: e.g., the set of pin cell and PWR fuel assembly problems will be suitable for those in charge of PWR in-core fuel management, and the set of pin cell and BWR fuel assembly problems for those in charge of BWR in-core fuel management. (author)

Fuel Fabrication and Nuclear Reactors

Energy Technology Data Exchange (ETDEWEB)

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

2017-02-02

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

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.

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

International Nuclear Information System (INIS)

Delafosse, Jacques.

1976-01-01

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

Status of research reactor spent fuel world-wide

International Nuclear Information System (INIS)

Ritchie, I.G.

2004-01-01

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

Plant-scale anodic dissolution of unirradiated N-Reactor fuel

International Nuclear Information System (INIS)

Gay, E.C.; Miller, W.E.; Laidler, J.J.

1995-01-01

Anodic dissolution tests were made with unirradiated N-Reactor fuel to determine the fuel segment length, diameter, and shape required for high throughput electrorefiner treatment for ultimate disposal in a geologic repository. Based on these tests, a conceptual design was produced of an electrorefiner for a full-scale plant to process N-Reactor spent fuel. In this design, the diameter of an electrode assembly is about 0.6 m (25 in.). Eight of these assemblies in an electrorefiner would accommodate a 1.333-metric-ton batch of N-Reactor fuel. Electrorefining would proceed at a rate of 40 kg uranium per hour

Design improvement for fretting-wear reduction of HANARO fuel assembly

Energy Technology Data Exchange (ETDEWEB)

Cho, Yeong Garp; Chae, H. T.; Ryu, J. S.; Kim, H. R

2000-06-01

In the course of the visual inspection of the fuel assemblies un-loaded from the reactor core in December 1996, it was observed that many of fuel assemblies had mechanical damages on some components. The major damage was the freting-wear on spacer plates and endplates due to the flow induced vibration of the fuel assembly in the flow tube. Since the reactor is activated and the system modification for complete removal of the driving factors of the vibration of fuel assemblies is practically very difficult, the focus has been on the design change of the fuel assemblies. Consequently, various design changes were proposed to strengthen the wear resistance of the components based on the evaluation of the visual inspection results. The validity of the proposals was verified through the performance tests for the modified components, and the vibration test and endurance test for the fuel assemblies using the single-channel test rig(SCTR) in AECL.The subsequent design changes were additionally proposed based on the visual inspections for the fuel assemblies that had been fabricated according to the first design change and loaded in the core. As the effects of the first design change, the fretting-wear of spacer plates was remarkably reduced and the period until fretting-wear damage was extended by 60% for the first modified 36-rod fuel assembly. It is too early to say the endurance life time for the first modified 18-rod fuel assembly because of insufficient statistical data of only two bundles damaged, but the fretting-wear at the bottom endplate slot was reduced to about 50%. The second modified fuel assemblies, that were not loaded into the core yet, are expected to meet the design requirements for the core residence time due to strengthening the weak parts from the fretting-wear point of view. This report describes design changes and tests for fuel assemblies of HANARO to reduce the fretting-wear, and estimates the effects of design improvement quantitatively compared

Design improvement for fretting-wear reduction of HANARO fuel assembly

International Nuclear Information System (INIS)

Cho, Yeong Garp; Chae, H. T.; Ryu, J. S.; Kim, H. R.

2000-06-01

In the course of the visual inspection of the fuel assemblies un-loaded from the reactor core in December 1996, it was observed that many of fuel assemblies had mechanical damages on some components. The major damage was the freting-wear on spacer plates and endplates due to the flow induced vibration of the fuel assembly in the flow tube. Since the reactor is activated and the system modification for complete removal of the driving factors of the vibration of fuel assemblies is practically very difficult, the focus has been on the design change of the fuel assemblies. Consequently, various design changes were proposed to strengthen the wear resistance of the components based on the evaluation of the visual inspection results. The validity of the proposals was verified through the performance tests for the modified components, and the vibration test and endurance test for the fuel assemblies using the single-channel test rig(SCTR) in AECL.The subsequent design changes were additionally proposed based on the visual inspections for the fuel assemblies that had been fabricated according to the first design change and loaded in the core. As the effects of the first design change, the fretting-wear of spacer plates was remarkably reduced and the period until fretting-wear damage was extended by 60% for the first modified 36-rod fuel assembly. It is too early to say the endurance life time for the first modified 18-rod fuel assembly because of insufficient statistical data of only two bundles damaged, but the fretting-wear at the bottom endplate slot was reduced to about 50%. The second modified fuel assemblies, that were not loaded into the core yet, are expected to meet the design requirements for the core residence time due to strengthening the weak parts from the fretting-wear point of view. This report describes design changes and tests for fuel assemblies of HANARO to reduce the fretting-wear, and estimates the effects of design improvement quantitatively compared

Fuel assembly for a nuclear reactor

International Nuclear Information System (INIS)

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

1975-01-01

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

Fuel and fuel cycles with high burnup for WWER reactors

International Nuclear Information System (INIS)

Chernushev, V.; Sokolov, F.

2002-01-01

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

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

Fuel element shipping shim for nuclear reactor

International Nuclear Information System (INIS)

Gehri, A.

1975-01-01

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

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)

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

Fuel bundle for nuclear reactor

International Nuclear Information System (INIS)

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

1977-01-01

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

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.)

Spent nuclear fuel discharges from US reactors 1993

International Nuclear Information System (INIS)

1995-02-01

The Energy Information Administration (EIA) of the U.S. Department of Energy (DOE) administers the Nuclear Fuel Data Survey, Form RW-859. This form is used to collect data on fuel assemblies irradiated at commercial nuclear reactors operating in the United States, and the current inventories and storage capacities of those reactors. These data are important to the design and operation of the equipment and facilities that DOE will use for the future acceptance, transportation, and disposal of spent fuels. The data collected and presented identifies trends in burnup, enrichment, and spent nuclear fuel discharged form commercial light-water reactor as of December 31, 1993. The document covers not only spent nuclear fuel discharges; but also site capacities and inventories; canisters and nonfuel components; and assembly type characteristics

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

Fuel assembly supporting structure

International Nuclear Information System (INIS)

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

1976-01-01

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

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)

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)

Plant-scale anodic dissolution of unirradiated N-Reactor fuel

International Nuclear Information System (INIS)

Gay, E.C.; Miller, W.E.; Laidler, J.J.

1995-01-01

Anodic dissolution tests were made with unirradiated N-Reactor fuel to determine the fuel segment length, diameter, and shape required for high throughput electro-refiner treatment for ultimate disposal in a geologic repository. Based on these tests, a conceptual design was produced of an electro-refiner for a full-scale plant to process N-Reactor spent fuel. In this design, the diameter of an electrode assembly is about 0.6 m (25 in.). Eight of these assemblies in an electro-refiner would accommodate a 1.333-metric-ton batch of N-Reactor fuel. Electrorefining would proceed at a rate of 40 kg uranium per hour. (author)

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)

Inlet for fuel assembly having finger control rods

International Nuclear Information System (INIS)

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

1975-01-01

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

Fast reactor fuel design and development

International Nuclear Information System (INIS)

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

1977-01-01

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

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.

Enhanced Westinghouse WWER-1000 fuel design for Ukraine reactors

International Nuclear Information System (INIS)

Dye, M.; Shah, H.

2015-01-01

Westinghouse has completed design, development, and region quantity delivery of an enhanced Westinghouse fuel assembly for WWER-1000 reactors to support continued safe reactor operations. The enhanced design builds on the successful performance of an earlier generation design which has operated in the South Ukraine 3 reactor for multiple cycles without any fuel rod failures. Incorporated design enhancements include a thicker spacer grid outer strap, an enhanced spacer grid outer strap profile to limit the risk for, and impact of, mechanical interaction/interference with coresident fuel, an all Alloy 718 grid structure for improved stability and strength, and improvements to the top and bottom nozzles. Capable of meeting increased lateral loads generated from using a higher axial trip limit for the refueling machine crane, the design was verified by extensive mechanical and thermalhydraulic testing, which included a newly developed fuel assembly-to-fuel assembly handling test rig to assess performance during bounding core loading and unloading conditions. Through these extensive design enhancements and comprehensive testing program, the enhanced WWER-1000 design provides additional performance, handling, and reliability margins for safe reactor operation. (authors)

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

Fuel assembly for nuclear reactor

International Nuclear Information System (INIS)

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

1996-01-01

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

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.)

Impact analysis of spent fuel jacket assemblies

International Nuclear Information System (INIS)

Aramayo, G.A.

1994-01-01

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

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)

Research reactor fuel - an update

International Nuclear Information System (INIS)

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

2003-01-01

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

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

Transfer of fuel assemblies

International Nuclear Information System (INIS)

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

1984-01-01

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

Blockages in LMFBR fuel assemblies: a review

International Nuclear Information System (INIS)

Han, J.T.; Fontana, M.H.

1977-01-01

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

Blockages in LMFBR fuel assemblies: a review

Energy Technology Data Exchange (ETDEWEB)

Han, J T; Fontana, M H

1977-01-01

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

Damage of fuel assembly premature changing in a power reactor

International Nuclear Information System (INIS)

Rudik, A.P.

1987-01-01

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

Siemens advance PWR fuel assemblies (HTP) and cladding

International Nuclear Information System (INIS)

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

1997-01-01

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

Storage experience in Hungary with fuel from research reactors

International Nuclear Information System (INIS)

Gado, J.; Hargitai, T.

1996-01-01

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

Development of WWER-440 fuel. Use of fuel assemblies of 2-nd and 3-rd generations with increased enrichment

International Nuclear Information System (INIS)

Vasilchenko, I.; Lushin, V.; Ananev, U.; Baranov, A.; Kukushkin, U.

2009-01-01

The problem of increasing the power of units at NPPs with WWER-440 is of current importance. There are all the necessary prerequisites for the above-stated problem as a result of updating the design of fuel assemblies and codes. The decrease of power peaking factor in the core is achieved by using profiled fuel assemblies, fuel-integrated burning absorber, FAs with modernized docking unit, modern codes, which allows decreasing conservatism of RP safety substantiation. A wide range of experimental studies of fuel behaviour has been performed which has reached burn-up of (50-60) MW·day/kgU in transition and emergency conditions, post-reactor studies of fuel assemblies, fuel rods and fuel pellets with a 5-year operating period have been performed, which prove high reliability of fuel, presence of a large margin in the fuel pillar, which helps reactor operation at increased power. The results of the work performed on introduction of 5-6 fuel cycles show that the ultimate fuel state on operability in WWER-440 reactors is far from being achieved. Neutron-physical and thermal-hydraulic characteristics of the cores of working power units with RP V-213 are such that actual (design and measured) power peaking factors on fuel assemblies and fuel rods, as a rule, are smaller than the maximum design values. This factor is a real reserve for power forcing. There is experience of operating Units 1, 2, 4 of the Kola NPP and Unit 2 of the Rovno NPP at increased power. Units of the Loviisa NPP are operated at 109 % power. During transfer to work at increased power it is reasonable to use fuel assemblies with increased height of the fuel pillar, which allows decreasing medium linear power distribution. Further development of the 2-nd generation fuel assembly design and consequent transition to working fuel assemblies of the 3-rd generation provides significant improvement of fuel consumption under the conditions of WWER-440 reactors operation with more continuous fuel cycles and

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)

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

Removable fuel assembly for nuclear reactor

International Nuclear Information System (INIS)

Dubief, J.M.; Bonnamour, M.

1984-01-01

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

Inspection and repair apparatus for a nuclear reactor fuel assembly

International Nuclear Information System (INIS)

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

1975-01-01

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

Tools for LWR spent fuel characterization: Assembly classes and fuel designs

International Nuclear Information System (INIS)

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

1991-01-01

The Characteristics Data Base (CDB) is sponsored by the DOE's Office of Civilian Radioactive Waste Management (OCRWM). The CDB provides a single, comprehensive source of data pertaining to radioactive wastes that will or may require geologic disposal, including detailed data describing the physical, quantitative, and radiological characteristics of light-water reactor (LWR) spent fuel. In developing the CDB, tools for the classification of fuel assembly types have been developed. The assembly class scheme is particularly useful for size- and handling-based describes these tools and presents results of their applications in the areas of fuel assembly type identification, characterization of projected discharges, cask accommodation analyses, and defective fuel analyses. Suggestions for additional applications are also made. 7 refs., 1 fig., 2 tabs

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)

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

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.

A study on 80 fuel assemblies core for HFETR

International Nuclear Information System (INIS)

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

1996-12-01

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

Results of trial operation of the WWER advanced fuel assemblies

International Nuclear Information System (INIS)

Vasilchenko, I.; Dragunov, Y.; Mikhalchuk, A.

2001-01-01

The paper describes results from experimental operation of advanced WWER-1000 fuel assemblies (AFA) at five units in Balakovo NPP. Advanced fuel is developed according to the concept of standard WWER-1000 fuel assembly (jacket-free). The new features includes: 1) zirconium guiding channels (alloy E-635 and E-110) and spacer grids (alloy E-110); 2) integrated burnable absorber gadolinium; 3) extended service life of fuel assemblies (FA) and absorber rods (possibility of repair of FA); 4) improved adoption to reactor conditions. Some results of AFA pilot operation of a three year operation are presented and analyses of effectiveness of improvements are made concerning application of zirconium channels and grids; application of integrated burnable absorbers; extension of FA and absorbing rods service life and FA repairability. These new features of WWER-1000 fuel design allow: 1) to reduce the average fuel enrichment to the 3.77% instead of 4.31% in U-235; 2) to reduce the FA axial load in reactor hot state by 40%,; 3) increasing of fuel operation in reactor to the 30000 effective days with possibility to have a 5-year residence time in the reactor. The design of new generation FA for WWER-440 reactors involves few key changes. Fuel inventory in new fuel design is increased due to elongation of fuel stack and reducing the diameter of the central hole. Vibration stability is enhanced as a result of: no-play junction of the fuel rod with the lower grid; change of SG arrangements; strengthening of the lower grid unit; secure of the central tube in the gap. Water-uranium ration is increased. Introduction of all these kinds of modernization in a 5-year fuel cycle reduces fuel component in the energy cost to the 7%

Proposed high throughput electrorefining treatment for spent N- Reactor fuel

International Nuclear Information System (INIS)

Gay, E.C.; Miller, W.E.; Laidler, J.J.

1996-01-01

A high-throughput electrorefining process is being adapted to treat spent N-Reactor fuel for ultimate disposal in a geologic repository. Anodic dissolution tests were made with unirradiated N-Reactor fuel to determine the type of fragmentation necessary to provide fuel segments suitable for this process. Based on these tests, a conceptual design was produced of a plant-scale electrorefiner. In this design, the diameter of an electrode assembly is about 1.07 m (42 in.). Three of these assemblies in an electrorefiner would accommodate a 3-metric-ton batch of N-Reactor fuel that would be processed at a rate of 42 kg of uranium per hour

Main results of post-irradiation examinations of new-generation fuel assemblies VVER-1000

International Nuclear Information System (INIS)

Zvir, E.; Markov, D.; Polenok, V.; Zhitelev, V.; Kobylyansky, G.

2009-01-01

To increase the competitiveness of Russian nuclear fuel at the foreign market and to improve its technical and economic performance in order to provide a necessary level of safety, it is necessary to solve certain important tasks: Increase of fuel burn-up; Extension of operational lifetime of fuel assemblies and operational reliability of nuclear fuel; Introduction of cost-beneficial and flexible fuel cycles. Alternative fuel assemblies TVSA VVER-1000 and TVS-2 are used as a basis to optimize the nuclear fuel and develop advanced fuel cycles for nuclear power plants with VVER-1000 reactor types. Four fuel assemblies TVSA operated during 1 and up to 6 reactor cycles, reference fuel assembly TVS-2 operated during three reactor cycles and achieved an average fuel burnup of 48MW·day/kgU as well as failed fuel assembly TVS-2 operated during one cycle were examined at RIAR in recent years. The main objectives of these examinations were to obtain experimental data in support of operational integrity of products or to find out reasons of their failure. The performed post-irradiation examinations confirmed the operational integrity of alternative fuel assemblies TVSA including their geometrical stability up to the average fuel burnup of 55 MW·day/kgU over the fuel assembly (FA) (up to the maximal fuel burnup of ∼73 MW·day/kgU in fuel rods) and of TVS-2 up to the average fuel burnup of 48 MW·day/kgU over the fuel assembly. The changes introduced in the design of VVER-1000 fuel assembly during the development of alternative fuel assembly TVSA and TVS-2 did not make any negative effect on fuel rods. It was proved that causes of fuel rod failure were not related to design features of fuel assemblies. The design features and operating conditions of fuel assemblies under examinations are briefly described. Post-irradiation examinations proved the geometrical stability of fuel assemblies TVSA and TVS-2 under operation up to the fuel burnup of ∼50 MW day/kgU, as for the

History of research reactor fuel fabrication at Babcock and Wilcox

International Nuclear Information System (INIS)

Freim, James B.

1983-01-01

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

The effect of the fuel rod friction force to the fuel assembly lateral mechanical characteristics

International Nuclear Information System (INIS)

Ha, Dong Geun; Jeon, Sang Youn; Suh, Jung Min

2012-01-01

The Fuel Assembly (FA) for light water reactor consists of hundreds of fuel rods, guide tubes, spacer grids, top/bottom nozzles. The guide tubes transmit vertical loads between the top and bottom nozzles, position the fuel rod support grids vertically, react the loads from the fuel rods that are applied to the grids, and provide some of the lateral load capability for the overall fuel assembly. The guide tubes are the structural members of the skeleton assembly. And the spacer grids maintain the fuel rod array by providing positive lateral restraint to the fuel rod but only frictional restraint in the axial direction. Figure 1 shows the outline of skeleton, FA and the location of guide tubes in the view of cross section. 17x17 FA has 24 guide tubes and one instrumentation tube. When the FA is in reactor, the lateral stiffness is one of very important factors from the view point of in reactor integrity of fuel assembly such as guarantee of the cool able geometry, the control rod insertion etc. The lateral stiffness of FA is mainly determined by skeleton lateral stiffness. And the fuel rods loaded in the spacer grids reinforce the FA lateral stiffness. Generally, fuel rods and spacer grids create the nonlinear friction force between fuel rod tube and grid spring/dimple against external lateral force of FA. Thus, it is necessary to study the contribution of the fuel rods friction force to the FA lateral stiffness. So, this paper is to show how much amount of the fuel rod grid interaction contributes to the FA lateral stiffness based on the test results

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)

U.S. Commercial Spent Nuclear Fuel Assembly Characteristics - 1968-2013

Energy Technology Data Exchange (ETDEWEB)

Hu, Jianwei [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Peterson, Joshua L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gauld, Ian C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bowman, Stephen M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2016-09-01

Activities related to management of spent nuclear fuel (SNF) are increasing in the US and many other countries. Over 240,000 SNF assemblies have been discharged from US commercial reactors since the late 1960s. The enrichment and burnup of SNF have changed significantly over the past 40 years, and fuel assembly designs have also evolved. Understanding the general characteristics of SNF helps regulators and other stakeholders form overall strategies towards the final disposal of US SNF. This report documents a survey of all US commercial SNF assemblies in the GC-859 database and provides reference SNF source terms (e.g., nuclide inventories, decay heat, and neutron/photon emission) at various cooling times up to 200 years after fuel discharge. This study reviews the distribution and evolution of fuel parameters of all SNF assemblies discharged over the past 40 years. Assemblies were categorized into three groups based on discharge year, and the median burnups and enrichments of each group were used to establish representative cases. An extended burnup case was created for boiling water reactor (BWR) fuels, and another was created for the pressurized water reactor (PWR) fuels. Two additional cases were developed to represent the eight mixed oxide (MOX) fuel assemblies in the database. Burnup calculations were performed for each representative case. Realistic parameters for fuel design and operations were used to model the SNF and to provide reference fuel characteristics representative of the current inventory. Burnup calculations were performed using the ORIGEN code, which is part of the SCALE nuclear modeling and simulation code system. Results include total activity, decay heat, photon emission, neutron flux, gamma heat, and plutonium content, as well as concentrations for 115 significant nuclides. These quantities are important in the design, regulation, and operations of SNF storage, transportation, and disposal systems.

Caramel, uranium oxide fuel plates for water cooled reactors

International Nuclear Information System (INIS)

Bussy, Pierre; Delafosse, Jacques; Lestiboudois, Guy; Cerles, J.-M.; Schwartz, J.-P.

1979-01-01

The fuel is composed of thin plates assembled parallel to each other to form bundles or assemblies. Each plate is composed of a pavement of uranium oxide pellets, insulated from each other by a zircaloy cladding. The 235 U enrichment does not exceed 8%. The range of uses for this fuel extends from electric power generating reactors to irradiation reactors for research work. A parametric study in test loops has made it possible to determine the operating limits of this thick fuel, without bursting. The resulting diagram gives the permissible power densities, with and without cycling for specific burn-ups beyond 50,000 MWd/t. The thinnest plates were also irradiated in total in the form of advance assemblies irradiated in the core of the OSIRIS pile prior to its transformation. This transformation and the operation of this reactor with a core of 'Caramel' elements is the main trial experiment of this fuel [fr

Application of ultra-sons to on-site spent fuel assemblies metrology

International Nuclear Information System (INIS)

Gondard, C.; Saglio, R.; Vouillot, M.; Delaroche, P.; Vaubert, Y.; Van Craeynest, J.C.

1983-12-01

Fuel assemblies inspection on the site of a power reactor, between two irradiation campaigns, allows to estimate the behaviour of prototype fuel assemblies and to permit their refueling for the continuation of the irradiation; the utilization of non-destructive, reliable and high-performance techniques, is of a great interest in the application. For, this reason, the C.E.A. has been led to carry out new techniques allowing the visual examination and the dimensional inspection of spent fuel assemblies of 900 MWe French pressurized water reactors, with a transportable Fuel Examination Module (MEC) on every reactor site. This module includes a television camera, and uses for the first time as ''position sensor'' the properties offered by a set of ultrasonic transducers. The main principle of the design, of the operation way of the module, of the measuring methods, and, of the data acquisition and processing, are presented [fr

Recent BWR fuel management reactor physics advances

International Nuclear Information System (INIS)

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

1982-01-01

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

Methods for estimating the reliability of the RBMK fuel assemblies and elements

International Nuclear Information System (INIS)

Klemin, A.I.; Sitkarev, A.G.

1985-01-01

Applied non-parametric methods for calculation of point and interval estimations for the basic nomenclature of reliability factors for the RBMK fuel assemblies and elements are described. As the fuel assembly and element reliability factors, the average lifetime is considered at a preset operating time up to unloading due to fuel burnout as well as the average lifetime at the reactor transient operation and at the steady-state fuel reloading mode of reactor operation. The formulae obtained are included into the special standardized engineering documentation

Tomographic imaging of severely disrupted fuel assemblies tested in TREAT

International Nuclear Information System (INIS)

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

1990-01-01

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

Refueling the RPI reactor facility with low-enrichment fuel

International Nuclear Information System (INIS)

Harris, D.R.; Rodriguez-Vera, F.; Wicks, F.E.

1985-01-01

The RPI Critical Facility has operated since 1963 with a core of thin, highly enriched fuel plates in twenty-five fuel assembly boxes. A program is underway to refuel the reactor with 4.81 w/o enriched SPERT (F-1) fuel rods. Use of these fuel rods will upgrade the capabilities of the reactor and will eliminate a security risk. Adequate quantities of SPERT (F-1) fuel rods are available, and their use will result in a great cost saving relative to manufacturing new low-enrichment fuel plates. The SPERT fuel rods are 19 inches longer than are the present fuel plates, so a modified core support structure is required. It is planned to support and position the SPERT fuel pins by upper and lower lattice plates, thus avoiding the considerable cost of new fuel assembly boxes. The lattice plates will be secured to the existing top and bottom plates. The design permits the fabrication and use of other lattice plates for critical experiment research programs in support of long-lived full development for power reactors. (author)

Nuclear fuel bundle disassembly and assembly tool

International Nuclear Information System (INIS)

Yates, J.; Long, J.W.

1975-01-01

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

In-core fuel management for nuclear reactor

International Nuclear Information System (INIS)

Ross, M.F.; Visner, S.

1986-01-01

This patent describes in-core fuel management for nuclear reactor in which the first cycle of a pressurized water nuclear power reactor has a multiplicity of elongated, square fuel assemblies supported side-by-side to form a generally cylindrical, stationary core consisting entirely of fresh fuel assemblies. Each assembly of the first type has a substantially similar low average fissile enrichment of at least about 1.8 weight percent U-235, each assembly of the second type having a substantially similar intermediate average fissile enrichment at least about 0.4 weight percent greater than that of the first type, and each assembly of the third type having a substantially similar high average fissile enrichment at least about 0.4 weight percent greater than that of the intermediate type, the arrangement of the low, intermediate, and high enrichment assembly types which consists of: a generally cylindrical inner core region consisting of approximately two-thirds the total assemblies in the core and forming a figurative checkerboard array having a first checkerboard component at least two-thirds of which consists of high enrichment and intermediate enrichment assemblies, at least some of the high enrichment assemblies containing fixed burnable poison shims, and a second checkerboard component consisting of assemblies other than the high enrichment type; and a generally annular outer region consisting of the remaining assemblies and including at least some but less than two-thirds of the high enrichment type assemblies

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.

About fuel assemblies optimization in research reactor

International Nuclear Information System (INIS)

Malers, Yu.P.

1992-01-01

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

Fuel arrangement for high temperature gas cooled reactor

International Nuclear Information System (INIS)

Tobin, J.M.

1978-01-01

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

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

Unified fuel elements development for research reactors

International Nuclear Information System (INIS)

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

1998-01-01

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

Fuel assemblies

International Nuclear Information System (INIS)

Yoshioka, Ritsuo.

1983-01-01

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

Top Nozzle Holddown Spring Optimization of KSNP Fuel Assembly

International Nuclear Information System (INIS)

Lee, Seong Ki; Park, Nam Kyu; Kim, Hyeong Koo; Lee, Joon Ro; Kim, Jae Won

2002-01-01

Nuclear fuel assembly for Korea Standard Nuclear Power (KSNP) Plant has 4 helical compression springs at the upper end of it. The springs, in conjunction with the fuel assembly weight, apply a holddown force against excess of buoyancy forces and the upward hydraulic forces due to the reactor coolant flow. Thus the holddown spring is to be designed such that the positive net downward force will be maintained for all normal and anticipated transient flow and temperature conditions in the nuclear reactor. With satisfying these in-reactor requirements of the fuel assembly holddown spring. Under the assumption that spring density is constant, the volume nozzle holddown spring. Under the assumption that spring density is constant, the volume minimization is executed by using the design variables, viz., wire diameter, mean coil diameter, minimization is executed by using the design variables, viz., wire diameter, mean coil diameter are within the compatible range of the fuel assembly structural components. Based on these conditions, the optimum design of the holddown spring is obtained considering the reactor operating condition and by using ANSYS code. The optimized spring has the properties that are a decreased volume and increased stiffness, compared with the existing one even if the absolute values are very similar each other. The holddown spring design features and the algorithm developed in this study could be directly applicable to the current commercial production. Therefore, it could be used to enhance the design efficiency and the functional performance of the spring, and to reduce a material cost a little

Thermohydraulic analysis of assemblies containing up to 2/7 fuel rods

International Nuclear Information System (INIS)

Ferreira, W.J.; Luz, M.

1985-01-01

The COBRA IV-I computer code was tested using data from the Fast Flux Test Facility. Then this code was applied to the analysis of fuel assemblies from the Binary Breeder Reactor. Previously this analysis was carried out using the COBRA III-C code which allows only for the calculations of fuel assemblies having seven fuel pins. The COBRA IV-I permits the calculation of fuel assemblies containing up to 217 fuel pins and the inclusion of blanket and shielding effects. (F.E.) [pt

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

Status of research reactor spent fuel world-wide: Database summary

International Nuclear Information System (INIS)

Ritchie, I.G.

1996-01-01

Results complied in the research reactor spent fuel database are used to assess the status of research reactor spent fuel world-wide. Fuel assemblies, their types, enrichment, origin of enrichment and geological distribution among the industrialized and developed countries of the world are discussed. Fuel management practices in wet and dry storage facilities and the concerns of reactor operators about long-term storage of their spent fuel are presented and some of the activities carried out by the International Atomic Energy Agency to address the issues associated with research reactor spent fuel are outlined. (author). 4 refs, 17 figs, 4 tabs

Assessment of Startup Fuel Options for the GNEP Advanced Burner Reactor (ABR)

Energy Technology Data Exchange (ETDEWEB)

Jon Carmack (062056); Kemal O. Pasamehmetoglu (103171); David Alberstein

2008-02-01

The Global Nuclear Energy Program (GNEP) includes a program element for the development and construction of an advanced sodium cooled fast reactor to demonstrate the burning (transmutation) of significant quantities of minor actinides obtained from a separations process and fabricated into a transuranic bearing fuel assembly. To demonstrate and qualify transuranic (TRU) fuel in a fast reactor, an Advanced Burner Reactor (ABR) prototype is needed. The ABR would necessarily be started up using conventional metal alloy or oxide (U or U, Pu) fuel. Startup fuel is needed for the ABR for the first 2 to 4 core loads of fuel in the ABR. Following start up, a series of advanced TRU bearing fuel assemblies will be irradiated in qualification lead test assemblies in the ABR. There are multiple options for this startup fuel. This report provides a description of the possible startup fuel options as well as possible fabrication alternatives available to the program in the current domestic and international facilities and infrastructure.

Fuel assembly for light-water cooled nuclear reactors

International Nuclear Information System (INIS)

Leroux, J.C.; Burfin, P.

1995-01-01

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

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

Nuclear fuel assembly

International Nuclear Information System (INIS)

Domoto, Noboru; Masuda, Hiroyuki

1989-01-01

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

Grapples for manipulating end fittings for nuclear reactor fuel assemblies

International Nuclear Information System (INIS)

Jabsen, F.S.

1982-01-01

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

Subcritical Measurements Research Program for Fresh and Spent Materials Test Reactor Fuels

International Nuclear Information System (INIS)

Blanchard, A.

1999-01-01

'A series of subcritical noise measurements were performed on fresh and spent University of Missouri Research Reactor fuel assemblies. These experimental measurements were performed for the purposes of providing benchmark quality data for validating transport theory computer codes and nuclear cross-section data used to perform criticality safety analyses for highly enriched, uranium-aluminum Material Test Reactor fuel assemblies. A mechanical test rig was designed and built to hold up to four fuel assemblies and neutron detectors in a subcritical array. The rig provided researchers with the ability to evaluate the reactivity effects of variable fuel/detector spacing, fuel rotation, and insertion of metal reflector plates into the lattice.'

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

Fuel assembly and reactor core

International Nuclear Information System (INIS)

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

1994-01-01

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

Establishment of China Nuclear Fuel Assembly Database

International Nuclear Information System (INIS)

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

2009-01-01

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

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

Concept of innovative water reactor for flexible fuel cycle (FLWR)

International Nuclear Information System (INIS)

Iwamura, T.; Uchikawa, S.; Okubo, T.; Kugo, T.; Akie, H.; Nakatsuka, T.

2005-01-01

In order to ensure sustainable energy supply in the future based on the matured Light Water Reactor (LWR) and coming LWR-Mixed Oxide (MOX) technologies, a concept of Innovative Water Reactor for Flexible Fuel Cycle (FLWR) has been investigated in Japan Atomic Energy Research Institute (JAERI). The concept consists of two parts in the chronological sequence. The first part realizes a high conversion type core concept, which is basically intended to keep the smooth technical continuity from current LWR and coming LWR-MOX technologies without significant gaps in technical point of view. The second part represents the Reduced-Moderation Water Reactor (RMWR) core concept, which realizes a high conversion ratio over 1.0 being useful for the long-term sustainable energy supply through plutonium multiple recycling based on the well-experienced LWR technologies. The key point is that the two core concepts utilize the compatible and the same size fuel assemblies, and hence, the former concept can proceed to the latter in the same reactor system, based flexibly on the fuel cycle circumstances during the reactor operation period around 60 years. At present, since the fuel cycle for the plutonium multiple recycling with MOX fuel reprocessing has not been realized yet, reprocessed plutonium from the LWR spent fuel is to be utilized in LWR-MOX. After this stage, the first part of FLWR, i.e. the high conversion type, can be introduced as a replacement of LWR or LWR-MOX. Since the plutonium inventory of FLWR is much larger, the number of the reactor with MOX fuel will be significantly reduced compared to the LWR-MOX utilization. The size of the fuel assembly for the first part is the same as in the RMWR concept, i.e. the hexagonal fuel assembly with the inner face-to-face distance of about 200 mm. Fuel rods are arranged in the triangular lattice with a relatively wide gap size around 3 mm between rods, and the effective MOX length is less than 1.5 m without using the blanket. When

Fuel assembly for BWR type reactor

International Nuclear Information System (INIS)

Ueda, Makoto

1990-01-01

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

Beginning-of-Life Data Report for the Instrumented Fuel Assembly (IFA)-527

Energy Technology Data Exchange (ETDEWEB)

Lanning, D. D.

1981-09-01

This report presents beginning-of-life (BOL) data from the first four months of operation of the six-rod instrumented fuel assembly (IFA)-527 in the Halden Boiling Water Reactor (HBWR), Halden, Norway. This assembly is the last in a series of U.S. Nuclear Regulatory Commission (NRC)-sponsored tests to verify steady-state fuel performance computer codes. IFA-527 contains five identical rods with high-density stable fuel pellets and 0.23-mm diametral gaps and one rod with similar fuel pellets but with a 0.06-mm diametral gap. All six rods were xenon-filled to provide simulation of the effects of fission gas and to enhance the observable effects of fuel cracking and relocation on fuel temperatures. The assembly operated successfully from July 1, 1980, to August 15, 1980; and then the reactor was shut down until September 10, 1980. Sometime during the shutdown, four of the six rods suffered pressure boundary failure. The decision was made to restart the reactor to collect operating data with failed rods. This report presents both pre- and postfailure data for IFA-527.

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

International Nuclear Information System (INIS)

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

1995-01-01

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

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

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)

Preparations for the shipment of RA-3 reactor irradiated fuel

International Nuclear Information System (INIS)

Goldschmidt, Adrian; Novara, Oscar; Lafuente, Jose

2002-01-01

During the last quarter of 2000, in the Radioactive Waste Management Area of the Argentine National Commission of Atomic Energy (CNEA), located at Ezeiza Atomic Center (CAE), activities associated to the shipment of 207 MTR spent fuels containing high enrichment uranium were carried out within the Foreign Research Reactor/Domestic Research Reactor Receipt Program launched by the US Department of Energy (DOE). The MTR spent fuel shipped to Savannah River Site (SRS) was fabricated in Argentina with 90% enriched uranium of US origin and it was utilized in the operation of the research and radioisotope production reactor RA-3 from 1968 until 1987. After a cooling period at the reactor, the spent fuel was transferred to the Central Storage Facility (CSF) located in the waste management area of CAE for interim storage. The spent fuel (SF) inventory consisted of 166 standard assemblies (SA) and 41 control assemblies (CA). Basically, the activities performed were the fuel conditioning operations inside the storage facility (remote transference of the assemblies to the operation pool, fuel cropping, fuel re-identification, loading in transport baskets, etc.) conducted by CNEA. The loading of the filled baskets in the transport casks (NAC-LWT) by means of intermediate transfer systems and loaded casks final preparations were conducted by NAC personnel (DOE's contractor) with the support of CNEA personnel. (author)

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

International Nuclear Information System (INIS)

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

1976-01-01

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

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

International Nuclear Information System (INIS)

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

2011-01-01

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

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

Nuclear reactor internals construction and failed fuel rod detection system

International Nuclear Information System (INIS)

Frisch, E.; Andrews, H.N.

1976-01-01

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

The Model of Temperature Dynamics of Pulsed Fuel Assembly

CERN Document Server

Bondarchenko, E A; Popov, A K

2002-01-01

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

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

Science.gov (United States)

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

2016-06-01

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

Fuel assembly leak tightness control on WWER-1000 reactor

International Nuclear Information System (INIS)

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

2001-01-01

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

PWR and BWR spent fuel assembly gamma spectra measurements

Energy Technology Data Exchange (ETDEWEB)

Vaccaro, S. [European Commission, DG Energy, Directorate EURATOM Safeguards Luxembourg (Luxembourg); Tobin, S.J.; Favalli, A. [Los Alamos National Laboratory, Los Alamos, NM (United States); Grogan, B. [Oak Ridge National Laboratory, Oak Ridge (United States); Jansson, P. [Uppsala University, Uppsala (Sweden); Liljenfeldt, H. [Oak Ridge National Laboratory, Oak Ridge (United States); Mozin, V. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Hu, J. [Oak Ridge National Laboratory, Oak Ridge (United States); Schwalbach, P. [European Commission, DG Energy, Directorate EURATOM Safeguards Luxembourg (Luxembourg); Sjöland, A. [Swedish Nuclear Fuel and Waste Management Company (SKB) (Sweden); Trellue, H.; Vo, D. [Los Alamos National Laboratory, Los Alamos, NM (United States)

2016-10-11

A project to research the application of nondestructive assay (NDA) to spent fuel assemblies is underway. The research team comprises the European Atomic Energy Community (EURATOM), embodied by the European Commission, DG Energy, Directorate EURATOM Safeguards; the Swedish Nuclear Fuel and Waste Management Company (SKB); two universities; and several United States national laboratories. The Next Generation of Safeguards Initiative–Spent Fuel project team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. This study focuses on spectrally resolved gamma-ray measurements performed on a diverse set of 50 assemblies [25 pressurized water reactor (PWR) assemblies and 25 boiling water reactor (BWR) assemblies]; these same 50 assemblies will be measured with neutron-based NDA instruments and a full-length calorimeter. Given that encapsulation/repository and dry storage safeguards are the primarily intended applications, the analysis focused on the dominant gamma-ray lines of {sup 137}Cs, {sup 154}Eu, and {sup 134}Cs because these isotopes will be the primary gamma-ray emitters during the time frames of interest to these applications. This study addresses the impact on the measured passive gamma-ray signals due to the following factors: burnup, initial enrichment, cooling time, assembly type (eight different PWR and six different BWR fuel designs), presence of gadolinium rods, and anomalies in operating history. To compare the measured results with theory, a limited number of ORIGEN-ARP simulations were performed.

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

International Nuclear Information System (INIS)

Yoshisaki, Magno B.; Leopando, Leonardo S.

1999-03-01

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

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.

Power generation costs for alternate reactor fuel cycles

International Nuclear Information System (INIS)

Smolen, G.R.; Delene, J.G.

1980-09-01

The total electric generating costs at the power plant busbar are estimated for various nuclear reactor fuel cycles which may be considered for power generation in the future. The reactor systems include pressurized water reactors (PWR), heavy-water reactors (HWR), high-temperature gas cooled reactors (HTGR), liquid-metal fast breeder reactors (LMFBR), light-water pre-breeder and breeder reactors (LWPR, LWBR), and a fast mixed spectrum reactor (FMSR). Fuel cycles include once-through, uranium-only recycle, and full recycle of the uranium and plutonium in the spent fuel assemblies. The U 3 O 8 price for economic transition from once-through LWR fuel cycles to both PWR recycle and LMFBR systems is estimated. Electric power generation costs were determined both for a reference set of unit cost parameters and for a range of uncertainty in these parameters. In addition, cost sensitivity parameters are provided so that independent estimations can be made for alternate cost assumptions

MOX fuel fabrication, in reactor performance and improvement