WorldWideScience

Sample records for nuclear fuel rods

  1. Nuclear fuel rod loading apparatus

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  2. Nuclear fuel rod loading apparatus

    International Nuclear Information System (INIS)

    King, H.B.

    1981-01-01

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

  3. Nuclear fuel rods

    International Nuclear Information System (INIS)

    Wada, Toyoji.

    1979-01-01

    Purpose: To remove failures caused from combination of fuel-cladding interactions, hydrogen absorptions, stress corrosions or the likes by setting the quantity ratio of uranium or uranium and plutonium relative to oxygen to a specific range in fuel pellets and forming a specific size of a through hole at the center of the pellets. Constitution: In a fuel rods of a structure wherein fuel pellets prepared by compacting and sintering uranium dioxide, or oxide mixture consisting of oxides of plutonium and uranium are sealed with a zirconium metal can, the ratio of uranium or uranium and plutonium to oxygen is specified as 1 : 2.01 - 1 : 2.05 in the can and a passing hole of a size in the range of 15 - 30% of the outer diameter of the fuel pellet is formed at the center of the pellet. This increases the oxygen partial pressure in the fuel rod, oxidizes and forms a protection layer on the inner surface of the can to control the hydrogen absorption and stress corrosion. Locallized stress due to fuel cladding interaction (PCMI) can also be moderated. (Horiuchi, T.)

  4. Nuclear reactor fuel rod

    International Nuclear Information System (INIS)

    Busch, H.; Mindnich, F.R.

    1973-01-01

    The fuel rod consists of a can with at least one end cap and a plenum spring between this cap and the fuel. To prevent the hazard that a eutectic mixture is formed during welding of the end cap, a thermal insulation is added between the end cap and plenum spring. It consists of a comical extension of the end cap with a terminal disc against which the spring is supported. The end cap, the extension, and the disc may be formed by one or several pieces. If the disc is separated from the other parts it may be manufactured from chrome steel or VA steel. (DG) [de

  5. Expandable device for a nuclear fuel rod

    International Nuclear Information System (INIS)

    Gesinski, L.T.

    1978-01-01

    A nuclear fuel rod and a device for use within the rod cladding to maintain the axial position of the fuel pellets stacked one atop another within the cladding are described. The device is initially of a smaller external cross-section than the fuel rod cladding internal cross-section so as to accommodate loading into the rod at preselected locations. During power operation the device responds to a rise in temperature, so as to permanently maintain its position and restrain any axial motion of the fuel pellets

  6. Nuclear reactor fuel rod attachment system

    International Nuclear Information System (INIS)

    Christiansen, D.W.

    1983-01-01

    The invention involves a technique to quickly, inexpensively and rigidly attach a nuclear reactor fuel rod to a support member. The invention also allows for the repeated non-destructive removal and replacement of the fuel rod. The proposed fuel rod and support member attachment and removal system consists of a locking cap fastened to the fuel rod and a locking strip fastened to the support member or vice versa. The locking cap has two or more opposing fingers shaped to form a socket. The fingers spring back when moved apart and released. The locking strip has an extension shaped to rigidly attach to the socket's body portion

  7. Nuclear fuel rod end plug weld inspection

    International Nuclear Information System (INIS)

    Parker, M. A.; Patrick, S. S.; Rice, G. F.

    1985-01-01

    Apparatus and method for testing TIG (tungsten inert gas) welds of end plugs on a sealed nuclear reactor fuel rod. An X-ray fluorescent spectrograph testing unit detects tungsten inclusion weld defects in the top end plug's seal weld. Separate ultrasonic weld inspection system testing units test the top end plug's seal and girth welds and test the bottom end plug's girth weld for penetration, porosity and wall thinning defects. The nuclear fuel rod is automatically moved into and out from each testing unit and is automatically transported between the testing units by rod handling devices. A controller supervises the operation of the testing units and the rod handling devices

  8. Individual nuclear fuel rod weighing system

    International Nuclear Information System (INIS)

    Fogg, J. L.; Howell, C. A.; Smith, J. H.; Vining, G. E.

    1985-01-01

    An individual nuclear fuel rod weighing system for rods carried on a tray which moves along a materials handling conveyor. At a first tray position on the conveyor, a lifting device raises the rods off the tray and places them on an overhead ramp. A loading mechanism conveys the rods singly from the overhead ramp onto an overhead scale for individual weighing. When the tray is at a second position on the conveyor, a transfer apparatus transports each weighed rod from the scale back onto the tray

  9. Individual nuclear fuel rod weighing system

    International Nuclear Information System (INIS)

    Fogg, J.L.; Smith, J.H.; Vining, G.E.; Howell, C.A.

    1985-01-01

    An individual nuclear fuel rod weighing system for rods carried on a tray which moves along a materials handling conveyor is discussed. At a first tray position on the conveyor, a lifting device raises the rods off the tray and places them on an overhead ramp. A loading mechanism conveys the rods singly from the overhead ramp onto an overhead scale for individual weighing. When the tray is at a second position on the conveyor, a transfer apparatus transports each weighed rod from the scale back onto the tray

  10. Model of cooling nuclear fuel rod in the nuclear reactor

    International Nuclear Information System (INIS)

    Lavicka, David; Polansky, Jiri

    2010-01-01

    The following topics are described: Some basic requirements for nuclear fuel rods; The VVER 1000 fuel rod; Classification of the two-phase flow in the vertical tube; Type of heat transfer crisis in the vertical tube; Experimental apparatus; Model of the nuclear fuel rod and spacers; Potential of the experimental apparatus (velocity profile measurement via PIV; thermal flow field measurement by the PLIF method; cooling graph in dependence on the fuel rod temperature; comparison of the hydrodynamic properties with respect to the design features of the spacers). (P.A.)

  11. Method for compacting spent nuclear reactor fuel rods

    International Nuclear Information System (INIS)

    Wachter, W.J.

    1988-01-01

    In a nuclear reactor system which requires periodic physical manipulation of spent fuel rods, the method of compacting fuel rods from a fuel rod assembly is described. The method consists of: (1) removing the top end from the fuel rod assembly; (2) passing each of multiple fuel rod pulling elements in sequence through a fuel rod container and thence through respective consolidating passages in a fuel rod directing chamber; (3) engaging one of the pulling elements to the top end of each of the fuel rods; (4) drawing each of the pulling elements axially to draw the respective engaged fuel rods in one axial direction through the respective the passages in the chamber to thereby consolidate the fuel rods into a compacted configuration of a cross-sectional area smaller than the cross-sectional area occupied thereby within the fuel rod assembly; and (5) drawing all of the engaged fuel rods concurrently and substantially parallel to one another in the one axial direction into the fuel rod container while maintaining the compacted configuration whereby the fuel rods are aligned within the container in a fuel rod density of the the fuel rod assembly

  12. Grid for nuclear fuel rod assembly

    International Nuclear Information System (INIS)

    Brayman, K.W.; George, D.K.; Rawlings, J.C.; Dix, G.E.

    1976-01-01

    A grid is described for placing a least four corner fuel rods in a tubular flow channel of a nuclear reactor. It includes a bearer component composed of four side strips joined by four corner strips so as to form a rigid unit structure, each side strip having an L-shaped piece adjacent at each of its ends to a lug of each L-shaped piece extending to the adjacent end of its associated side strip [fr

  13. Automated nuclear fuel rod pattern loading system

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  14. Automated nuclear fuel rod pattern loading system

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  15. Method and apparatus for compacting spent nuclear reactor fuel rods

    International Nuclear Information System (INIS)

    Wachter, W.J.

    1988-01-01

    In a nuclear reactor system requiring periodic physical manipulation of spent fuel rods, the method of compacting fuel rods from a fuel rod assembly is described comprising the steps of: (1) removing the top end from pulling members having electrodes of weld elements in leading ends thereof in sequence through a fuel rod container and thence through respective consolidating passages in a fuel-rod directing chamber; (3) welding the weld elements of the pulling members to the top end of respective fuel rods corresponding to the respective pulling members; (4) drawing each of the pulling members axially to draw the respective engaged fuel rods in one axial direction through the respective passages in the chamber to thereby consolidate the fuel rods into a compacted configuration of a cross-sectional area smaller than the cross-sectional area occupied thereby within the fuel rod assembly; and (5) drawing all of the engaged fuel rods concurrently and substantially parallel to one another to the one axial direction into the fuel rod container while maintaining the compacting configuration in a fuel rod density which is greater than that of the fuel rod density of the fuel rod assembly

  16. Thermal phenomenae in nuclear fuel rods

    International Nuclear Information System (INIS)

    Baigorria, Carlos.

    1983-12-01

    Thermal phenomenae occurring in a nuclear fuel rod under irradiation are studied. The most important parameters of either steady or transient thermal states are determined. The validity of applying the Fourier's approximation equations to these problems is also studied. A computer program TRANS is developed in order to study the transient cases. This program solves a system of coupled, non-linear partial differential equations, of parabolic type, in cylindrical coordinates with various boundary conditions. The benchmarking of the TRANS program is done by comparing its predictions with the analytical solution of some simplified transient cases. Complex transient cases such as those corresponding to characteristic reactor accidents are studied, in particular for typical pressurized heavy water reactor (PHWR) fuel rods, such as those of Atucha I. The Stefan problem emerging in the case of melting of the fuel element is solved. Qualitative differences between the classical Stefan problem, without inner sources, and that one, which includes sources are discussed. The MSA program, for solving the Stefan problem with inner sources is presented; and furthermore, it serves to predict thermal evolution, when the fuel element melts. Finally a model for fuel phase change under irradiation is developed. The model is based on the dimensional invariants of the percolation theory when applied to the connectivity of liquid spires nucleated around each fission fragment track. Suggestions for future research into the subject are also presented. (autor) [es

  17. Welding nuclear reactor fuel rod end plugs

    International Nuclear Information System (INIS)

    Yeo, D.

    1984-01-01

    Apparatus for applying a vacuum to a nuclear fuel rod cladding tube's interior through its open end while girth welding an inserted end plug to its other end. An airtight housing has an orifice with a seal which can hermetically engage the tube's open end. A vacuum hose has one end connected to the housing and the other end connected to a vacuum pump. A mechanized device which moves the housing to engage or disengage its seal with the tube's open end includes at least one arm having one end attached to the housing and the other end pivotally attached to a movable table; an arm rotating device to coaxially align the housing's orifice with the welding-positioned tube; and a table moving device to engage the seal of the coaxially aligned orifice with the tube's open end. (author)

  18. Failure position detection device for nuclear fuel rod

    International Nuclear Information System (INIS)

    Ishida, Takeshi; Higuchi, Shin-ichi; Ito, Masaru; Matsuda, Yasuhiko

    1987-01-01

    Purpose: To easily detect failure position of a nuclear fuel rod by relatively moving an air-tightly shielded detection portion to a fuel rod. Constitution: For detecting the failure position of a leaked fuel assembly, the fuel assembly is dismantled and a portion of withdrawn fuel rod is air-tightly sealed with an inspection portion. The inside of the inspection portion is maintained at a pressure-reduced state. Then, in a case if failed openings are formed at a portion sealed by the inspection portion in the fuel rod, FP gases in the fuel rod are released based on the reduced pressure and the FP gases are detected in the detection portion. Accordingly, by relatively moving the detection portion to the fuel rod, the failure position can be detected. (Yoshino, Y.)

  19. Failure position detection device for nuclear fuel rod

    Energy Technology Data Exchange (ETDEWEB)

    Ishida, Takeshi; Higuchi, Shin-ichi; Ito, Masaru; Matsuda, Yasuhiko

    1987-03-24

    Purpose: To easily detect failure position of a nuclear fuel rod by relatively moving an air-tightly shielded detection portion to a fuel rod. Constitution: For detecting the failure position of a leaked fuel assembly, the fuel assembly is dismantled and a portion of withdrawn fuel rod is air-tightly sealed with an inspection portion. The inside of the inspection portion is maintained at a pressure-reduced state. Then, in a case if failed openings are formed at a portion sealed by the inspection portion in the fuel rod, FP gases in the fuel rod are released based on the reduced pressure and the FP gases are detected in the detection portion. Accordingly, by relatively moving the detection portion to the fuel rod, the failure position can be detected. (Yoshino, Y.).

  20. Method and apparatus for inspection of nuclear fuel rods

    International Nuclear Information System (INIS)

    Wachter, W.J.

    1977-01-01

    A method and apparatus are provided for the inspection of nuclear fuel rods to detect defects or failures in such rods. Assemblies of fuel rods are immersed in water and means are provided for causing a change in the relative pressures in the water and within the fuel rod such that fluid is expelled from the rod through any defects that may exist. Means are also provided for thereafter vibrating the rods to cause additional internal fluid or other material that may be trapped in the rod to be expelled. Sensors are provided for detecting the emission of bubbles of fluid or other material from the rod and for locating the position of the defective rod in the assembly. 5 figures

  1. 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

  2. Method for wrapping a wire round a nuclear fuel rod

    International Nuclear Information System (INIS)

    Nakayasu, Fumio.

    1974-01-01

    Object: To provide a method for winding a wire round a nuclear fuel rod with accurate pitches without imparting any local strain or torsion to the wire. Structure: A wire is fixed on one end of the fuel rod, and the other end of the wire is secured to a universal joint leaving a winding allowance to the fuel rod. The wire is linearly stretched by a predetermined tension through the universal joint so as to provide an angle of development theta corresponding to the desired winding pitch, and then, the fuel rod may be rotated so that the end of the wire on the side of the universal joint is moved towards the fuel rod so as to render the angle of development theta constant in proportion to said rotation of the fuel rod. (Kamimura, M.)

  3. Method of manufacturing nuclear fuel rods

    International Nuclear Information System (INIS)

    Sato, Masao; Oyama, Masatoshi; Yamamoto, Takanobu.

    1976-01-01

    Object: To discriminate the properties of light white deposits on a clad tube during the process of manufacturing nuclear fuel rods and then remove this to reproduce a good clad tube, thereby enhancing a yield of the clad tube. Structure: When a light white deposits is found to be appeared on outer or inner surface of coating during the process of appearance inspection, this is then permitted to subject to treatment of hot water immersion and discrimination. Requirements for removal of adhered matter in the process of treatment of hot water immersion are that deioned water of specific resistance 5 x 10 5 ohms or more is used with water temperature maintained at 60 to 100 0 C for immersion treatment for 10 to 30 minutes. In this case, however, if the water temperature is more than 80 0 C, the immersion time can be set less than 10 minutes. With the addition of such process described above, about 2.5% of total receiving number can be reproduced. (Yoshihara, H.)

  4. Thermal behavior simulation of a nuclear fuel rod through an eletrically heated rod

    International Nuclear Information System (INIS)

    Lima, R. de C.F. de.

    1984-01-01

    In thermalhydraulic loops the nuclear industry often uses electrically heated rods to simulate power transients, which occur in nuclear fuel rods. The development and design of a electrically heated rod, by supplying the dimensions and materials which should be used in order to yeld the same temperature and heat flux at the surfaces of the nuclear rod and the electrically heated rod are presented. To a given nuclear transient this equality was obtained by fitting the linear power through the lumped parameters technique. (Author) [pt

  5. Device for detecting defective nuclear reactor fuel rods

    International Nuclear Information System (INIS)

    Steven, J.

    1976-01-01

    A moisture sensor is provided for a nuclear fuel rod for water-cooled nuclear reactors wherein moisture can be present. The fuel rod has an end cap and a charge of nuclear fuel. The moisture sensor is disposed between the end cap and the charge and serves to detect a leak in the fuel rod. The moisture sensor includes a capsule-like housing having an inner space and having openings through which moisture can pass into the inner space in the event of a leak in the fuel rod. Ferromagnetic material is disposed in the inner space of the housing together with a moisture detector responsive to moisture for altering the diposition of the ferromagnetic material in the inner space. 5 claims, 6 drawing figures

  6. Nuclear Fuel Test Rod Fabrication for Data Acquisition Test

    International Nuclear Information System (INIS)

    Joung, Chang-Young; Hong, Jin-Tae; Kim, Ka-Hye; Huh, Sung-Ho

    2014-01-01

    A nuclear fuel test rod must be fabricated with precise welding and assembly technologies, and confirmed for their soundness. Recently, we have developed various kinds of processing systems such as an orbital TIG welding system, a fiber laser welding system, an automated drilling system and a helium leak analyzer, which are able to fabricate the nuclear fuel test rods and rigs, and keep inspection systems to confirm the soundness of the nuclear fuel test rods and rids. The orbital TIG welding system can be used with two kinds of welding methods. One can perform the round welding for end-caps of a nuclear fuel test rod by an orbital head mounted in a low-pressure chamber. The other can do spot welding for a pin-hole of a nuclear fuel test rod in a high-pressure chamber to fill up helium gas of high pressure. The fiber laser welding system can weld cylindrical and 3 axis samples such as parts of a nuclear fuel test rod and instrumentation sensors which is moved by an index chuck and a 3 axis (X, Y, Z) servo stage controlled by the CNC program. To measure the real-time temperature change at the center of the nuclear fuel during the irradiation test, a thermocouple should be instrumented at that position. Therefore, a hole needs to be made at the center of fuel pellet to instrument the thermocouple. An automated drilling system can drill a fine hole into a fuel pellet without changing tools or breaking the work-piece. The helium leak analyzer (ASM-380 model of DEIXEN Co.) can check the leak of the nuclear fuel test rod filled with helium gas. This paper describes not only the assembly and fabrication methods used by the process systems, but also the results of the data acquisition test for the nuclear fuel test rod. A nuclear fuel test rod for the data acquisition test was fabricated using the welding and assembling echnologies acquired from previous tests

  7. Nuclear Fuel Test Rod Fabrication for Data Acquisition Test

    Energy Technology Data Exchange (ETDEWEB)

    Joung, Chang-Young; Hong, Jin-Tae; Kim, Ka-Hye; Huh, Sung-Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    A nuclear fuel test rod must be fabricated with precise welding and assembly technologies, and confirmed for their soundness. Recently, we have developed various kinds of processing systems such as an orbital TIG welding system, a fiber laser welding system, an automated drilling system and a helium leak analyzer, which are able to fabricate the nuclear fuel test rods and rigs, and keep inspection systems to confirm the soundness of the nuclear fuel test rods and rids. The orbital TIG welding system can be used with two kinds of welding methods. One can perform the round welding for end-caps of a nuclear fuel test rod by an orbital head mounted in a low-pressure chamber. The other can do spot welding for a pin-hole of a nuclear fuel test rod in a high-pressure chamber to fill up helium gas of high pressure. The fiber laser welding system can weld cylindrical and 3 axis samples such as parts of a nuclear fuel test rod and instrumentation sensors which is moved by an index chuck and a 3 axis (X, Y, Z) servo stage controlled by the CNC program. To measure the real-time temperature change at the center of the nuclear fuel during the irradiation test, a thermocouple should be instrumented at that position. Therefore, a hole needs to be made at the center of fuel pellet to instrument the thermocouple. An automated drilling system can drill a fine hole into a fuel pellet without changing tools or breaking the work-piece. The helium leak analyzer (ASM-380 model of DEIXEN Co.) can check the leak of the nuclear fuel test rod filled with helium gas. This paper describes not only the assembly and fabrication methods used by the process systems, but also the results of the data acquisition test for the nuclear fuel test rod. A nuclear fuel test rod for the data acquisition test was fabricated using the welding and assembling echnologies acquired from previous tests.

  8. Nuclear reactor fuel rod behavior modelling and current trends

    International Nuclear Information System (INIS)

    Colak, Ue.

    2001-01-01

    Safety assessment of nuclear reactors is carried out by simulating the events to taking place in nuclear reactors by realistic computer codes. Such codes are developed in a way that each event is represented by differential equations derived based on physical laws. Nuclear fuel is an important barrier against radioactive fission gas release. The release of radioactivity to environment is the main concern and this can be avoided by preserving the integrity of fuel rod. Therefore, safety analyses should cover an assessment of fuel rod behavior with certain extent. In this study, common approaches for fuel behavior modeling are discussed. Methods utilized by widely accepted computer codes are reviewed. Shortcomings of these methods are explained. Current research topics to improve code reliability and problems encountered in fuel rod behavior modeling are presented

  9. Apparatus for injection casting metallic nuclear energy fuel rods

    Science.gov (United States)

    Seidel, Bobby R.; Tracy, Donald B.; Griffiths, Vernon

    1991-01-01

    Molds for making metallic nuclear fuel rods are provided which present reduced risks to the environment by reducing radioactive waste. In one embodiment, the mold is consumable with the fuel rod, and in another embodiment, part of the mold can be re-used. Several molds can be arranged together in a cascaded manner, if desired, or several long cavities can be integrated in a monolithic multiple cavity re-usable mold.

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

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  11. Process for automatic filling of nuclear fuel rod cans

    International Nuclear Information System (INIS)

    Bezold, H.

    1977-01-01

    A drying section is inserted in the production line for the automation of the filling process for fuel rods with nuclear fuel pellets. The pellets are taken in a drum magazine to a drying furnace and then pushed out one after the other into the can to be filled. (TK) [de

  12. Control rod cluster with removable rods for nuclear fuel assembly

    International Nuclear Information System (INIS)

    Denizou, J.P.

    1989-01-01

    For each removable control rod, the open end section of the sleeve has a certain length of reduced diameter with openings in its wall. The top end of the rod is joined to an extension tube that surrounds the shaft over part of its lenght. This extension tube fits over the reduced part of the sleeve when the shaft is screwed into the bore of the sleeve. Rotation of the rod in the sleeve is prevented by deforming the extension tube locally in the openings of the end part of the sleeve. The rod is dismantled by exerting a torque on it using a gripping area near the end of the rod [fr

  13. 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

  14. Radiography inspection of weld for nuclear fuel rod

    International Nuclear Information System (INIS)

    Zhang Kai; Zhang Xichang

    1995-05-01

    The survey of radiography inspection, advantages, disadvantages and applications of main kinds of radiography inspection methods are presented. Emphasis is put upon the structure and functions of X-ray flaw detecting device for nuclear fuel rod welds, the actuating program of the device, as well as the structure of some key mechanism and the functions of them. The analysis is made upon the actuating principles. Finally, the test of long-term operation proves the device to be stable in operation, reliable in action, to possess high level of automation and high sensitivity and it can simultaneously perform on-line X-ray inspection of 25 nuclear fuel rods with a diameter less than 10 mm, and meet the requirements of large-scale production of nuclear fuel rods (5 figs.)

  15. On Cherenkov light production by irradiated nuclear fuel rods

    International Nuclear Information System (INIS)

    Branger, E.; Grape, S.; Svärd, S. Jacobsson; Jansson, P.; Sundén, E. Andersson

    2017-01-01

    Safeguards verification of irradiated nuclear fuel assemblies in wet storage is frequently done by measuring the Cherenkov light in the surrounding water produced due to radioactive decays of fission products in the fuel. This paper accounts for the physical processes behind the Cherenkov light production caused by a single fuel rod in wet storage, and simulations are presented that investigate to what extent various properties of the rod affect the Cherenkov light production. The results show that the fuel properties have a noticeable effect on the Cherenkov light production, and thus that the prediction models for Cherenkov light production which are used in the safeguards verifications could potentially be improved by considering these properties. It is concluded that the dominating source of the Cherenkov light is gamma-ray interactions with electrons in the surrounding water. Electrons created from beta decay may also exit the fuel and produce Cherenkov light, and e.g. Y-90 was identified as a possible contributor to significant levels of the measurable Cherenkov light in long-cooled fuel. The results also show that the cylindrical, elongated fuel rod geometry results in a non-isotropic Cherenkov light production, and the light component parallel to the rod's axis exhibits a dependence on gamma-ray energy that differs from the total intensity, which is of importance since the typical safeguards measurement situation observes the vertical light component. It is also concluded that the radial distributions of the radiation sources in a fuel rod will affect the Cherenkov light production.

  16. 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

  17. Fuel rods

    International Nuclear Information System (INIS)

    Adachi, Hajime; Ueda, Makoto

    1985-01-01

    Purpose: To provide a structure capable of measuring, in a non-destructive manner, the releasing amount of nuclear gaseous fission products from spent fuels easily and at a high accuracy. Constitution: In order to confirm the integrity and the design feasibility of a nuclear fuel rod, it is important to accurately determine the amount of gaseous nuclear fission products released from nuclear pellets. In a structure where a plurality of fuel pellets are charged in a fuel cladding tube and retained by an inconel spring, a hollow and no-sealed type spacer tube made of zirconium or the alloy thereof, for example, not containing iron, cobalt, nickel or manganese is formed between the spring and the upper end plug. In the fuel rod of such a structure, by disposing a gamma ray collimator and a gamma ray detector on the extension of the spacer pipe, the gamma rays from the gaseous nuclear fission products accumulated in the spacer pipe can be detected while avoiding the interference with the induction radioactivity from inconel. (Kamimura, M.)

  18. Calculation of fission gases internal pressure in nuclear fuel rods

    International Nuclear Information System (INIS)

    Vasconcelos Santana, M. de.

    1981-12-01

    Models concerning the principal phenomena, particularly thermal expansion, fuel swelling, densification, reestructuring, relocation, mechanical strain, fission gas production and release, direct or indirectly important to calculate the internal pressure in nuclear fuel rods were analysed and selected. Through these analyses a computer code was developed to calculate fuel pin internal pressure evolution. Three different models were utilized to calculate the internal pressure in order to select the best and the most conservative estimate. (Author) [pt

  19. Express diagnostics of WWER fuel rods at nuclear power plants

    International Nuclear Information System (INIS)

    Pavlov, S.; Amosov, S.; Sagalov, S.; Kostyuchenko, A.

    2009-01-01

    Higher safety and economical efficiency of nuclear power plants (NPP) call for a continuous design modification and technological development of fuel assemblies and fuel rods as well as optimization of their operating conditions. In doing so the efficiency of new fuel introduction depends on the completeness of irradiated fuel data in many respects as well as on the rapidity and cost of such data obtaining. Standard examination techniques of fuel assemblies (FA) and fuel rods (FR) intended for their use in hot cell conditions do not satisfy these requirements in full extent because fuel assemblies require preliminary cooling at NPP to provide their shipment to the research center. Expenditures for FA transportation, capacity of hot cells and expenditures for the examined fuel handling do not make it possible to obtain important information about the condition of fuel assemblies and fuel rods after their operation. In order to increase the comprehensiveness of primary data on fuel assemblies and fuel rods immediately after their removal from the reactor, inspection test facilities are widely used for these purposes. The inspection test facilities make it possible to perform nondestructive inspection of fuel in the NPP cooling pools. Moreover these test facilities can be used to repair failed fuel assemblies. The ultrasonic testing of failed fuel rods inside the fuel assembly was developed for stands of inspection and repair of TVSA WWER-1000 for the Kalinin NPP and Temelin NPP. This method was tested for eight leaking fuel assemblies WWER-440 and WWER-1000 with a burnup of ∼14 up to 38 MW·day/kgU. The ultrasonic testing proved its high degree of reliability and efficiency. The defectoscopy by means of the pulsed eddy-current method was adapted for the stand of inspection and repair of TVSA WWER-1000 for the Kalinin NPP. This method has been used at RIAR as an express testing method of FR claddings during the post-irradiation examinations of fuel assemblies WWER

  20. Nuclear fuel rod helium leak inspection apparatus and method

    International Nuclear Information System (INIS)

    Ahmed, H.J.

    1991-01-01

    This patent describes an inspection apparatus for testing nuclear fuel rods for helium leaks. It comprises a test chamber being openable and closable for receiving at least one nuclear fuel rod; means separate from the fuel rod for supplying helium and constantly leaking helium at a predetermined known positive value into the test chamber to constantly provide an atmosphere of helium at the predetermined known positive value in the test chamber; and means for sampling the atmosphere within the chamber and measuring the helium in the atmosphere such that a measured helium value below a preset minimum helium value substantially equal to the predetermined known positive value of the atmosphere of helium being constantly provided in the test chamber indicates a malfunction in the inspection apparatus, above a preset maximum helium value greater than the predetermined known positive in the test chamber indicates the existence of a helium leak from the fuel rod, or between the preset minimum and maximum helium values indicates the absence of a helium leak from the fuel rod

  1. Apparatus for inspecting the quality of nuclear fuel rod ends

    International Nuclear Information System (INIS)

    Brashier, R.W.; Pfau, E.D.

    1990-01-01

    This patent describes an apparatus for inspecting the quality of both ends of nuclear fuel rods. It comprises: a housing including a pair of longitudinally separated slots for receiving X-ray downwardly therethrough from an external source and so as to define first and second longitudinally spaced apart operating positions, means for serially guiding nuclear fuel rods longitudinally through the housing and to a first rod position wherein the forward ends of the rods are aligned below the first operating position and to a second rod position wherein the rear ends of the rods are aligned below the second operating position, belt conveyor assembly means for serially advancing X-ray film cartridges longitudinally through the housing and below the rods, and so that each cartridge may be selectively aligned below the first and second operating positions; and table means supported by the conveyor frame for selectively lifting the film cartridges supported by the belts and so that the conveyor belts may be advanced while the film cartridges are held stationary

  2. Heat Transfer Coefficient Variations in Nuclear Fuel Rod Bundles

    International Nuclear Information System (INIS)

    Conner, Michael E.; Holloway, Mary V.

    2007-01-01

    The single-phase heat transfer performance of a PWR nuclear fuel rod bundle is enhanced by the use of mixing vanes attached to the downstream edges of the support grid straps. This improved single-phase performance will delay the onset of nucleate boiling, thereby reducing corrosion and delaying crud-related issues. This paper presents the variation in measured single-phase heat transfer coefficients (HTC) for several grid designs. Then, this variation is compared with observations of actual in-core crud patterns. While crud deposition is a function of a number of parameters including rod heat flux, the HTC is assumed to be a primary factor in explaining why crud deposition is a local phenomenon on nuclear fuel rods. The data from this study will be used to examine this assumption by providing a comparison between HTC variations and crud deposition patterns. (authors)

  3. Storage device for fuel rods of nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    Kempf, B.

    1983-01-01

    The storage device, which can be flexibly matched to the number of fuel rods to be stored and is not tied to a space, has a vertical support post situated on the floor and a stiff upright also situated vertically on the floor, which is used to accommodate at least one fuel rod. The stiff upright is connected directly to the support post by connections which can be undone, or form locking via another vertical stiff upright situation on the floor. (orig./HP) [de

  4. Preliminary nuclear design for test MOX Fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Joo, Hyung Kook; Kim, Taek Kyum; Jeong, Hyung Guk; Noh, Jae Man; Cho, Jin Young; Kim, Young Il; Kim, Young Jin; Sohn, Dong Seong

    1997-10-01

    As a part of activity for future fuel development project, test MOX fuel rods are going to be loaded and irradiated in Halden reactor core as a KAERI`s joint international program with Paul Scherrer Institute (PSI). PSI will fabricate test MOX rods with attrition mill device which was developed by KAERI. The test fuel assembly rig contains three MOX rods and three inert matrix rods. One of three MOX rods will be fabricated by BNFL, the other two MOX fuel rods will be manufacturing jointly by KAERI and PSI. Three inert matrix fuel rods will be fabricated with Zr-Y-Er-Pu oxide. Neutronic evaluation was preliminarily performed for test fuel assembly suggested by PSI. The power distribution of test fuel rod in test fuel assembly was analyzed for various fuel rods position in assembly and the depletion characteristic curve for test fuel was also determined. The fuel rods position in test fuel assembly does not effect the rod power distribution, and the proposal for test fuel rods suggested by PSI is proved to be feasible. (author). 2 refs., 13 tabs., 16 figs.

  5. Fuel rod bundles proposed for advanced pressure tube nuclear reactors

    International Nuclear Information System (INIS)

    Prodea, Iosif; Catana, Alexandru

    2010-01-01

    The paper aims to be a general presentation for fuel bundles to be used in Advanced Pressure Tube Nuclear Reactors (APTNR). The characteristics of such a nuclear reactor resemble those of known advanced pressure tube nuclear reactors like: Advanced CANDU Reactor (ACR TM -1000, pertaining to AECL) and Indian Advanced Heavy Water Reactor (AHWR). We have also developed a fuel bundle proposal which will be referred as ASEU-43 (Advanced Slightly Enriched Uranium with 43 rods). The ASEU-43 main design along with a few neutronic and thermalhydraulic characteristics are presented in the paper versus similar ones from INR Pitesti SEU-43 and CANDU-37 standard fuel bundles. General remarks regarding the advantages of each fuel bundle and their suitability to be burned in an APTNR reactor are also revealed. (authors)

  6. Fuel rods

    International Nuclear Information System (INIS)

    Fukushima, Kimichika.

    1984-01-01

    Purpose: To reduce the size of the reactor core upper mechanisms and the reactor container, as well as decrease the nuclear power plant construction costs in reactors using liquid metals as the coolants. Constitution: Isotope capturing devices comprising a plurality of pipes are disposed to the gas plenum portion of a nuclear fuel rod main body at the most downstream end in the flowing direction of the coolants. Each of the capturing devices is made of nickel, nickel alloys, stainless steel applied with nickel plating on the surface, nickel alloys applied with nickel plating on the surface or the like. Thus, radioactive nuclides incorporated in the coolants are surely captured by the capturing devices disposed at the most downstream end of the nuclear fuel main body as the coolants flow along the nuclear fuel main body. Accordingly, since discharging of radioactive nuclides to the intermediate fuel exchange system can be prevented, the maintenance or reparing work for the system can be facilitated. (Moriyama, K.)

  7. Development for analysis system of rods enrichment of nuclear fuels

    International Nuclear Information System (INIS)

    Rojas C, E.L.

    1998-01-01

    Nuclear industry is strongly regulated all over the world and quality assurance is important in every nuclear installation or process related with it. Nuclear fuel manufacture is not the exception. ININ was committed to manufacture four nuclear fuel bundles for the CFE nucleo electric station at Laguna Verde, Veracruz, under General Electric specifications and fulfilling all the requirements of this industry. One of the quality control requisites in nuclear fuel manufacture deals with the enrichment of the pellets inside the fuel bundle rods. To achieve the quality demanded in this aspect, the system described in this work was developed. With this system, developed at ININ it is possible to detect enrichment spikes since 0.4 % in a column of pellets with a 95 % confidence interval and to identify enrichment differences greater than 0.2 % e between homogeneous segments, also with a 95 % confidence interval. ININ delivered the four nuclear fuel bundles to CFE and these were introduced in the core of the nuclear reactor of Unit 1 in the fifth cycle. Nowadays they are producing energy and have shown a correct mechanical performance and neutronic behavior. (Author)

  8. Visualization test facility of nuclear fuel rod emergency cooling system

    International Nuclear Information System (INIS)

    Candido, Marcos Antonio; Mesquita, Amir Zacarias; Rezende, Hugo Cesar; Santos, Andre Augusto Campagnole

    2013-01-01

    The nuclear reactors safety is determined according to their protection against the consequences that may result from postulated accidents. The Loss of Coolant Accident (LOCA) is one the most important design basis accidents (DBA). The failure may be due to rupture of the primary loop piping. Another accident postulated is due to lack of power in the pump motors in the primary circuit. In both cases the reactor shut down automatically due to the decrease of reactivity to maintain the fissions, and to the drop of control rods. In the event of an accident it is necessary to maintain the coolant flow to remove the fuel elements residual heat, which remains after shut down. This heat is a significant amount of the maximum thermal power generated in normal operation (about 7%). Recently this event has been quite prominent in the press due to the reactor accident in Fukushima nuclear power station. This paper presents the experimental facility under rebuilding at the Thermal Hydraulic Laboratory of the Nuclear Technology Development Center (CDTN) that has the objective of monitoring and visualization of the process of emergency cooling of a nuclear fuel rod simulator, heated by Joule effect. The system will help the comprehension of the heat transfer process during reflooding after a loss of coolant accident in the fuel of light water reactor core. (author)

  9. Method and apparatus for sizing nuclear fuel rod cladding tubes

    International Nuclear Information System (INIS)

    Koehler, L.

    1976-01-01

    Nuclear fuel rod cladding tubes are sized internally to diameters precisely fitting nuclear fuel pellets with which the tubes are charged by externally applying hydraulic pressure to short lengths of each tube. The pressure is applied while the tube is stationary. The tube is then moved to bring a new length within the hydraulic pressure zone. The volume of the hydraulic liquid used and the pressure applied to this liquid is such that the liquid is compressed slightly so that the length being sized yields, the expansion of the liquid then completing the sizing. The lengths being sized step-by-step are internally supported by either the fuel pellets or a mandrel having the same diameter as the pellets

  10. Nuclear fuel rod grip with modified diagonal spring structures

    International Nuclear Information System (INIS)

    DeMario, E.E.

    1990-01-01

    This patent describes a spring structure in a nuclear fuel rod grid including a plurality of inner and outer straps being interleaved with one another to form a matrix of hollow cells. Each of the cells is for receiving one fuel rod and being defined by pairs of opposing wall sections of the straps which wall sections are shared with adjacent cells. Each of the cells has a central longitudinal axis, a fuel rod engaging spring structure of resiliently yieldable material being integrally formed on each wall section of the inner straps. The spring structure comprising: a pair of spaced apart opposite outer portions being integrally attached at their outer ends to the respective wall section. The portions extending in alignment with one another and in generally diagonal relation to the direction of the central longitudinal axis of the one cell; and a middle portion disposed between and integrally connected at its outer ends with respective inner ends of the outer portions. The middle portion extending in generally transverse relation to the direction of the central longitudinal axis of the one cell

  11. Interim transfer canister for consolidating nuclear fuel rods

    International Nuclear Information System (INIS)

    Formanek, F.J.

    1987-01-01

    This patent describes a canister for receiving and consolidating a group of uniformly spaced apart nuclear fuel rods, comprising: a rectangular, vertically oriented straight back panel; a pair of oppositely disposed side panels connected perpendicularly to the back panel, having a vertical straight upper portion and an inwardly tapered lower portion; a front panel opposite the back panel and connected to the side panels, having a straight vertical upper portion and inwardly tapered lower portion; whereby the back, side and front panels define a rectangular upper opening at the upper end of the canister and a generally rectangular lower opening at the other end, the lower opening having a cross-sectional area less than one-half that of the upper opening; parallel plate members spanning the canister from the front panel to the back panel, each plate spaced from the other the same uniform distance, the plates extending downwardly into the tapered portion of the canister while remaining spaced above the tapered sidewalls; first base means at the lower end of the canister, removably mounted and having an oblique orientation generally downward from the front panel to the back panel, for guiding the fuel rods to be inserted preferentially toward the lower portion of the back panel; and second base means removably mounted within the canister below first base means and oriented transversely to the longitudinal extent of the canister, for supporting the fuel rods when the first base means is removed from the canister

  12. Fuel rod for liquid metal-cooled nuclear reactors

    International Nuclear Information System (INIS)

    Vinz, P.

    1976-01-01

    In fuel rods for nuclear reactors with liquid-metal cooling (sodium), with stainless steel tubes with a nitrated surface as canning, superheating or boiling delay should be avoided. The inner wall of the can is provided along its total length with a helical fin of stainless steel wire (diameter 0.05 to 0.5 mm) to be wetted by hot sodium. This fin is mounted under prestressing and has a distance in winding of 1/10 of the wire diameter. (UWI) [de

  13. Apparatus for inspecting a irradiated nuclear fuel rod

    International Nuclear Information System (INIS)

    Saura, Hideaki; Yonemura, Eizo.

    1975-01-01

    Object: To increase safety and inspection efficiency by operating irradiated fuel rods, which are accommodated in a water-filled pool after being taken out from the reactor. Structure: When making inspection of irradiated fuel rods, particularly the cladding tube thereof, a fuel box which stores irradiated fuel rods in a water pool is secured to a securement mechanism with slime removal apparatus and inspection apparatus on either side capable of being vertically moved, and it is then stopped at a water depth of about 2 meters. When the lid of the box is opened, irradiated fuel rods are taken out with gripping means and then secured together with the gripping means to an operation base provided on the outside of the pool. Thereafter, the box is lowered by operating pedals on the operation base to completely pull out the irradiated fuel rods from the box, and the irradiated fuel rods are then horizontally moved and then held in a suspended state. Next a slime removal apparatus in raised by operating pedals and an inspection element assembly are progressively raised for inspection of the state of the cladding tube of each fuel rod after removal of slime therefrom. (Nakamura, S.)

  14. SCADOP: Phenomenological modeling of dryout in nuclear fuel rod bundles

    Energy Technology Data Exchange (ETDEWEB)

    Dasgupta, Arnab, E-mail: arnie@barc.gov.in; Chandraker, D.K., E-mail: dineshkc@barc.gov.in; Vijayan, P.K., E-mail: vijayanp@barc.gov.in

    2015-11-15

    Highlights: • Phenomenological model for annular flow dryout is presented. • The model evaluates initial entrained fraction using a new methodology. • The history effect in annular flow is predicted and validated. • Rod bundle dryout is predicted using subchannel methodology. • Model is validated against experimental dryout data in tubes and rod bundles. - Abstract: Analysis and prediction of dryout is of important consequence to safety of nuclear fuel clusters of boiling water type of reactors. Traditionally, experimental correlations are used for dryout predictions. Since these correlations are based on operating parameters and do not aim to model the underlying phenomena, there has been a proliferation of the correlations, each catering to some specific bundle geometry under a specific set of operating conditions. Moreover, such experiments are extremely costly. In general, changes in tested bundle geometry for improvement in thermal-hydraulic performance would require re-experimentation. Understanding and modeling the basic processes leading to dryout in flow boiling thus has great incentive. Such a model has the ability to predict dryout in any rod bundle geometry, unlike the operating parameter based correlation approach. Thus more informed experiments can be carried out. A good model can, reduce the number of experiments required during the iterations in bundle design. In this paper, a phenomenological model as indicated above is presented. The model incorporates a new methodology to estimate the Initial Entrained Fraction (IEF), i.e., entrained fraction at the onset of annular flow. The incorporation of this new methodology is important since IEF is often assumed ad-hoc and sometimes also used as a parameter to tune the model predictions to experimental data. It is highlighted that IEF may be low under certain conditions against the general perception of a high IEF due to influence of churn flow. It is shown that the same phenomenological model is

  15. Fuel rods

    International Nuclear Information System (INIS)

    Hattori, Shinji; Kajiwara, Koichi.

    1980-01-01

    Purpose: To ensure the safety for the fuel rod failures by adapting plenum springs to function when small forces such as during transportation of fuel rods is exerted and not to function the resilient force when a relatively great force is exerted. Constitution: Between an upper end plug and a plenum spring in a fuel rod, is disposed an insertion member to the lower portion of which is mounted a pin. This pin is kept upright and causes the plenum spring to function resiliently to the pellets against the loads due to accelerations and mechanical vibrations exerted during transportation of the fuel rods. While on the other hand, if a compression force of a relatively high level is exerted to the plenum spring during reactor operation, the pin of the insertion member is buckled and the insertion member is inserted to the inside of the plenum spring, whereby the pellets are allowed to expand freely and the failures in the fuel elements can be prevented. (Moriyama, K.)

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

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  17. Out-pile Test of Double Cladding Fuel Rod Mockups for a Nuclear Fuel Irradiation Test

    Energy Technology Data Exchange (ETDEWEB)

    Sohn, Jaemin; Park, Sungjae; Kang, Younghwan; Kim, Harkrho; Kim, Bonggoo; Kim, Youngki [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2008-05-15

    An instrumented capsule for a nuclear fuel irradiation test has been developed to measure fuel characteristics, such as a fuel temperature, internal pressure of a fuel rod, a fuel pellet elongation and a neutron flux during an irradiation test at HANARO. In the future, nuclear fuel irradiation tests under a high temperature condition are expected from users. To prepare for this request, we have continued developing the technology for a high temperature nuclear fuel irradiation test at HANARO. The purpose of this paper is to verify the possibility that the temperature of a nuclear fuel can be controlled at a high temperature during an irradiation test. Therefore we designed and fabricated double cladding fuel rod mockups. And we performed out-pile tests using these mockups. The purposes of a out-pile test is to analyze an effect of a gap size, which is between an outer cladding and an inner cladding, on the temperature and the effect of a mixture ratio of helium gas and neon gas on the temperature. This paper presents the design and fabrication of double cladding fuel rod mockups and the results of the out-pile test.

  18. Simulation of nuclear fuel rods by using process computer-controlled power for indirect electrically heated rods

    International Nuclear Information System (INIS)

    Malang, S.

    1975-11-01

    An investigation was carried out to determine how the simulation of nuclear fuel rods with indirect electrically heated rods could be improved by use of a computer to control the electrical power during a loss-of-coolant accident (LOCA). To aid in the experiment, a new version of the HETRAP code was developed which simulates a LOCA with heater rod power controlled by a computer that adjusts rod power during a blowdown to minimize the difference in heat flux of the fuel and heater rods. Results show that without computer control of heater rod power, only the part of a blowdown up to the time when the heat transfer mode changes from nucleate boiling to transition or film boiling can be simulated well and then only for short times. With computer control, the surface heat flux and temperature of an electrically heated rod can be made nearly identical to that of a reactor fuel rod with the same cooling conditions during much of the LOCA. A small process control computer can be used to achieve close simulation of a nuclear fuel rod with an indirect electrically heated rod

  19. Automatic system of welding for nuclear fuel rods

    International Nuclear Information System (INIS)

    Romero G, M; Romero C, J.

    1998-01-01

    The welding process of nuclear fuel must be realized in an inert gas environment (He) and constant flow of this. In order to reach these conditions it is necessary to do vacuum at the chamber and after it is pressurized with the noble gas (purge) twice in the welding chamber. The purge eliminates impurities that can provoke oxidation in the weld. Once the conditions for initiating the welding are gotten, it is necessary to draw a graph of the flow parameters, pressure, voltage and arc current and to analyse those conditions in which have been carried out the weld. The rod weld must be free of possible pores or cracks which could provoke rod leaks, so reducing the probability of these failures should intervene mechanical and metallurgical factors. Automatizing the process it allows to do reliable welding assuring that conditions have been performed, reaching a high quality welding. Visually it can be observed the welding process by means of a mimic which represents the welding system. There are the parameters acquired such as voltage, current, pressure and flow during the welding arc to be analysed later. (Author)

  20. Drilling Experiments of Dummy Fuel Rods Using a Mock-up Drilling Device and Detail Design of Device for Drilling of Irradiated Nuclear Fuel Rods

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jae Yong; Lee, H. K.; Chun, Y. B.; Park, S. J.; Kim, B. G

    2007-07-15

    KAERI are developing the safety evaluation method and the analysis technology for high burn-up nuclear fuel rod that is the project, re-irradiation for re-instrumented fuel rod. That project includes insertion of a thermocouple in the center hole of PWR nuclear fuel rod with standard burn-up, 3,500{approx}4,000MWD/tU and then inspection of the nuclear fuel rod's heat performance during re-irradiation. To re-fabricate fuel rod, two devices are needed such as a drilling machine and a welding machine. The drilling machine performs grinding a center hole, 2.5 mm in diameter and 50 mm in depth, for inserting a thermocouple. And the welding machine is used to fasten a end plug on a fuel rod. Because these two equipment handle irradiated fuel rods, they are operated in hot cell blocked radioactive rays. Before inserting any device into hot cell, many tests with that machine have to be conducted. This report shows preliminary experiments for drilling a center hole on dummy of fuel rods and optimized drilling parameters to lessen operation time and damage of diamond dills. And the design method of a drilling machine for irradiated nuclear fuel rods and detail design drawings are attached.

  1. Lateral Flow Field Behavior Downstream of Mixing Vanes In a Simulated Nuclear Fuel Rod Bundle

    International Nuclear Information System (INIS)

    Conner, Michael E.; Smith, L. David III; Holloway, Mary V.; Beasley, Donald E.

    2004-01-01

    To assess the fuel assembly performance of PWR nuclear fuel assemblies, average subchannel flow values are used in design analyses. However, for this highly complex flow, it is known that local conditions around fuel rods vary dependent upon the location of the fuel rod in the fuel assembly and upon the support grid design that maintains the fuel rod pitch. To investigate the local flow in a simulated nuclear fuel rod bundle, a testing technique has been employed to measure the lateral flow field in a 5 x 5 rod bundle. Particle Image Velocimetry was used to measure the lateral flow field downstream of a support grid with mixing vanes for four unique subchannels in the 5 x 5 bundle. The dominant lateral flow structures for each subchannel are compared in this paper including the decay of these flow structures. (authors)

  2. Ventilating system for reprocessing of nuclear fuel rods

    International Nuclear Information System (INIS)

    Szulinski, M.J.

    1981-01-01

    In a nuclear facility such as a reprocessing plant for nuclear fuel rods, the central air cleaner discharging ventilating gas to the atmosphere must meet preselected standards not only as to the momentary concentration of radioactive components, but also as to total quantity per year. In order to comply more satisfactorily with such standards, reprocessing steps are conducted by remote control in a plurality of separate compartments. The air flow for each compartment is regulated so that the air inventory for each compartment has a slow turnover rate of more than a day but less than a year, which slow rate is conveniently designated as quasihermetic sealing. The air inventory in each such compartment is recirculated through a specialized processing unit adapted to cool and/or filter and/or otherwise process the gas. Stale air is withdrawn from such recirculating inventory and fresh air is injected (eg., By the less than perfect sealing of a compartment) into such recirculating inventory so that the air turnover rate is more than a day but less than a year. The amount of air directed through the manifold and duct system from the reprocessing units to the central air cleaner is less than in reprocessing plants of conventional design

  3. Neural signal processing for identifying failed fuel rods in nuclear reactors

    International Nuclear Information System (INIS)

    Seixas, Jose M. de; Soares Filho, William; Pereira, Wagner C.A.; Teles, Claudio C.B.

    2002-01-01

    Ultrasonic pulses were used for automatic detection of failed nuclear fuel rods. For experimental tests of the proposed method, an assembly prototype of 16 x 16 rods was built by using genuine rods but without fuel inside (just air). Some rods were partially filled with water to simulate cracked rods. Using neural signal processing on the received echoes of the emitted ultrasonic pulses, a detection efficiency of 97% was obtained. Neural detection is shown to outperform other classical discriminating methods and can also reveal important features of the signal structure of the received echoes. (author)

  4. Considerations about the utilization of electrically heated rods used for simulation of nuclear fuel pins

    International Nuclear Information System (INIS)

    Lima, R. de C.F. de; Carajilescov, P.

    1987-01-01

    The dinamic behavior of electrically heated rods used for simulation of nuclear fuel pins in nuclear power transients, is analysed by the application of the lumped parameter and the finite difference methods. Deviations of the rods surface conditions, for extreme accidental transient conditions are presented and discussed. (author) [pt

  5. Segmented fuel and moderator rod

    International Nuclear Information System (INIS)

    Doshi, P.K.

    1987-01-01

    This patent describes a continuous segmented fuel and moderator rod for use with a water cooled and moderated nuclear fuel assembly. The rod comprises: a lower fuel region containing a column of nuclear fuel; a moderator region, disposed axially above the fuel region. The moderator region has means for admitting and passing the water moderator therethrough for moderating an upper portion of the nuclear fuel assembly. The moderator region is separated from the fuel region by a water tight separator

  6. Positioning device for fuel rods of nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    1987-01-01

    The positioning device consists of individual containers, similar to cases, for the fuel elements. These cases are arranged vertically next to one another and are held by means of vertical support posts and horizontal arms. The openings of the cases can be individually approached by the trolleys. (DG) [de

  7. Underwater Nuclear Fuel Disassembly and Rod Storage Process and Equipment Description. Volume II

    International Nuclear Information System (INIS)

    Viebrock, J.M.

    1981-09-01

    The process, equipment, and the demonstration of the Underwater Nuclear Fuel Disassembly and Rod Storage System are presented. The process was shown to be a viable means of increasing spent fuel pool storage density by taking apart fuel assemblies and storing the fuel rods in a denser fashion than in the original storage racks. The assembly's nonfuel-bearing waste is compacted and containerized. The report documents design criteria and analysis, fabrication, demonstration program results, and proposed enhancements to the system

  8. Gas sealing welding method and device for nuclear fuel rod

    International Nuclear Information System (INIS)

    Seki, Masayuki; Nishiyama, Motokuni; Kamimura, Katsuichiro; Yagi, Eiji; Nakase, Tsuyoshi; Kobogata, Sadao; Taniguchi, Jun-ichi; Uesugi, Yoshisaku.

    1995-01-01

    An end plug and a cladding tube are held by clamping, respectively, by opposing movable electrode and static electrode. The movable electrode is forwarded toward the static electrode. The end plug and the cladding tube are abutted and held at a slight gap between their end faces. A region to be welded is surrounded by a pressurizing chamber and the side of the chamber is evacuated and He gas is filled in the cladding tube. Then, one of the electrodes is forwarded, to seal the abutted end faces of the end plug and the cladding tube. Then, pressure and welding current required for welding are applied to the abutted ends, and He gas is sealed in the vessel. The displacement of pressurization caused by slipping when the required pressure is applied to the abutted ends is detected by a sensor, and the operation of the welding control device for starting current supply is terminated by the detection signals. Abutment accuracy between the abutment of the cladding tube and the end plug as a nuclear fuel rod can be ensured, to further improve and stabilize the welding quality. (N.H.)

  9. The nuclear fuel rod character recognition system based on neural network technique

    International Nuclear Information System (INIS)

    Kim, Woong-Ki; Park, Soon-Yong; Lee, Yong-Bum; Kim, Seung-Ho; Lee, Jong-Min; Chien, Sung-Il.

    1994-01-01

    The nuclear fuel rods should be discriminated and managed systematically by numeric characters which are printed at the end part of each rod in the process of producing fuel assembly. The characters are used to examine manufacturing process of the fuel rods in the inspection process of irradiated fuel rod. Therefore automatic character recognition is one of the most important technologies to establish automatic manufacturing process of fuel assembly. In the developed character recognition system, mesh feature set extracted from each character written in the fuel rod is employed to train a neural network based on back-propagation algorithm as a classifier for character recognition system. Performance evaluation has been achieved on a test set which is not included in a training character set. (author)

  10. Development of Welding and Instrumentation Technology for Nuclear Fuel Test Rod

    International Nuclear Information System (INIS)

    Joung, Chang Young; Ahn, Sung Ho; Heo, Sung Ho; Hong, Jin Tae; Kim, Ka Hye

    2013-01-01

    It is necessary to develop various types of welding, instrumentation and helium gas filling techniques that can conduct TIG spot welding exactly at a pin-hole of the end-cap on the nuclear fuel rod to fill up helium gas. The welding process is one of the most important among the instrumentation processes of the nuclear fuel test rod. To manufacture the nuclear fuel test rod, a precision welding system needs to be fabricated to develop various welding technologies of the fuel test rod jointing the various sensors and end-caps on a fuel cladding tube, which is charged with fuel pellets and component parts. We therefore designed and fabricated an orbital TIG welding system and a laser welding system. This paper describes not only some experiment results from weld tests for the parts of a nuclear fuel test rod, but also the contents for the instrumentation process of the dummy fuel test rod installed with the C-type T. C. A dummy nuclear fuel test rod was successfully fabricated with the welding and instrumentation technologies acquired with various tests. In the test results, the round welding has shown a good weldability at both the orbital TIG welding system and the fiber laser welding system. The spot welding to fill up helium gas has shown a good welding performance at a welding current of 30A, welding time of 0.4 sec and gap of 1 mm in a helium gas atmosphere. The soundness of the nuclear fuel test rod sealed by a mechanical sealing method was confirmed by helium leak tests and microstructural analyses

  11. Development of Welding and Instrumentation Technology for Nuclear Fuel Test Rod

    Energy Technology Data Exchange (ETDEWEB)

    Joung, Chang Young; Ahn, Sung Ho; Heo, Sung Ho; Hong, Jin Tae; Kim, Ka Hye [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    It is necessary to develop various types of welding, instrumentation and helium gas filling techniques that can conduct TIG spot welding exactly at a pin-hole of the end-cap on the nuclear fuel rod to fill up helium gas. The welding process is one of the most important among the instrumentation processes of the nuclear fuel test rod. To manufacture the nuclear fuel test rod, a precision welding system needs to be fabricated to develop various welding technologies of the fuel test rod jointing the various sensors and end-caps on a fuel cladding tube, which is charged with fuel pellets and component parts. We therefore designed and fabricated an orbital TIG welding system and a laser welding system. This paper describes not only some experiment results from weld tests for the parts of a nuclear fuel test rod, but also the contents for the instrumentation process of the dummy fuel test rod installed with the C-type T. C. A dummy nuclear fuel test rod was successfully fabricated with the welding and instrumentation technologies acquired with various tests. In the test results, the round welding has shown a good weldability at both the orbital TIG welding system and the fiber laser welding system. The spot welding to fill up helium gas has shown a good welding performance at a welding current of 30A, welding time of 0.4 sec and gap of 1 mm in a helium gas atmosphere. The soundness of the nuclear fuel test rod sealed by a mechanical sealing method was confirmed by helium leak tests and microstructural analyses.

  12. Welding of stainless steel clad fuel rods for nuclear reactors

    International Nuclear Information System (INIS)

    Neves, Mauricio David Martins das

    1986-01-01

    This work describes the obtainment of austenitic stainless steel clad fuel rods for nuclear reactors. Two aspects have been emphasized: (a) obtainment and qualification of AISI 304 and 304 L stainless steel tubes; b) the circumferential welding of pipe ends to end plugs of the same alloy followed by qualification of the welds. Tubes with special and characteristic dimensions were obtained by set mandrel drawing. Both, seamed and seamless tubes of 304 and 304 L were obtained.The dimensional accuracy, surface roughness, mechanical properties and microstructural characteristics of the tubes were found to be adequate. The differences in the properties of the tubes with and without seams were found to be insignificant. The TIG process of welding was used. The influence of various welding parameters were studied: shielding gas (argon and helium), welding current, tube rotation speed, arc length, electrode position and gas flow. An inert gas welding chamber was developed and constructed with the aim of reducing surface oxidation and the heat affected zone. The welds were evaluated with the aid of destructive tests (burst-test, microhardness profile determination and metallographic analysis) and non destructive tests (visual inspection, dimensional examination, radiography and helium leak detection). As a function of the results obtained, two different welding cycles have been suggested; one for argon and another for helium. The changes in the microstructure caused by welding have been studied in greater detail. The utilization of work hardened tubes, permitted the identification by optical microscopy and microhardness measurements, of the different zones: weld zone; heat affected zone (region of grain growth, region of total and partial recrystallization) and finally, the zone not affected by heat. Some correlations between the welding parameters and metallurgical phenomena such as: solidification, recovery, recrystallization, grain growth and precipitation that occurred

  13. Fuel rod pressure in nuclear power reactors: Statistical evaluation of the fuel rod internal pressure in LWRs with application to lift-off probability

    Energy Technology Data Exchange (ETDEWEB)

    Jelinek, Tomas

    2001-02-01

    In this thesis, a methodology for quantifying the risk of exceeding the Lift-off limit in nuclear light water power reactors is outlined. Due to fission gas release, the pressure in the gap between the fuel pellets and the cladding increases with burnup of the fuel. An increase in the fuel-clad gap due to clad creep would be expected to result in positive feedback, in the form of higher fuel temperatures, leading to more fission gas release, higher rod pressure, etc, until the cladding breaks. An increase in the fuel-clad gap that leads to this positive feedback is a phenomenon called Lift-off and is a limitation that must be considered in the fuel core management. Lift-off is a consequence of very high internal fuel rod pressure. The internal fuel rod pressure is therefore used as a Lift-off indicator. The internal fuel rod pressure is closely connected to the fission gas release into the fuel rod plenum and is thus used to increase the database. It is concluded that the dominating error source in the prediction of the pressure in Boiling Water Reactors (BWR), is the power history. There is a bias in the fuel pressure prediction that is dependent on the fuel rod position in the fuel assembly for BWRs. A methodology to quantify the risk of the fuel rod internal pressure exceeding a certain limit is developed; the risk is dependent of the pressure prediction and the fuel rod position. The methodology is based on statistical treatment of the discrepancies between predicted and measured fuel rod internal pressures. Finally, a methodology to estimate the Lift-off probability of the whole core is outlined.

  14. 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.)

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

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  16. Nuclear fuel rod grid spring and dimple structures having chamfered edges for reduced pressure drop

    International Nuclear Information System (INIS)

    De Mario, E.E.

    1990-01-01

    This patent describes a nuclear fuel rod grid including inner and outer straps being interleaved with one another to form a matrix of hollow cells, each cell for receiving one fuel rod and being defined by pairs of opposing wall sections of the straps which wall sections are shared with adjacent cells, each cell having a central longitudinal axis defining a coolant flow direction through the cell, at least fuel rod engaging dimple structure of resiliently yieldable material being integrally formed on each wall section of the inner straps

  17. Limits on the experimental simulation of nuclear fuel rod response

    International Nuclear Information System (INIS)

    Hagar, R.C.

    1980-01-01

    The steady-state and transient effects of intrinxic geometric and material property differences between typical nuclear fuel pins and electric fuel pin simulators (FPSs) are identified. The effectiveness of varying the transient power supplied to the FPS in reducing the differences between the transient responses of nuclear fuel pins and FPSs is investigated. This effectiveness is shown to be limited by the heat capacity of the FPS, the allowed range of the power program, and different FPS power requirements at different positions on a full-length FPS

  18. Apparatus and method for preventing the rotation of rods used in nuclear fuel assemblies

    International Nuclear Information System (INIS)

    Pilgrim, L.G. Jr.; Jackson, L.F.

    1985-01-01

    Apparatus and method for preventing the rotation of one or more elongated rods used in nuclear fuel assemblies include an end plug secured to one longitudinal end of such an elongated rod and having an out-of-cavity, non-round structure affixed thereto and configured to mate with a complementary shaped structure in a lower tie plate of a nuclear fuel assembly in such a manner as to prevent the rotation of the rod about its longitudinal axis. In one embodiment, the end plug includes a pair of flats formed on a portion of the end plug and configured to abut against a pair of flats formed on the outer surface of a cylindrical boss or sleeve of the lower tie plate, thereby to prevent the rotation of the rod. In another embodiment, four grooves, disposed 90 0 apart about the periphery of an end plug of a rod form a spline. The grooves are configured to receive four, radially inwardly protruding, key members disposed 90 0 apart about the periphery of a sleeve secured to the lower tie plate, thereby to prevent the rotation of the rod. In a further embodiment, a sleeve is secured to an end plug of a rod and includes four elongated slots disposed 90 0 apart about the periphery of the sleeve and configured in width, depth and spacing to receive and mate with four web portions of the lower tie plate of the nuclear fuel assembly, thereby to secure the rod against rotation about its longitudinal axis

  19. Development of nuclear fuel rod inspection technique using ultrasonic resonance phenomenon

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Myung Sun; Lee, Jong Po; Ju, Young Sang [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-11-01

    Acoustic resonance scattering from a nuclear fuel rod in water is analyzed. A new model for the background which is attributed to the interference of reflected wave and diffracted wave is found and here named {sup t}he inherent background{sup .} The resonance spectrum of a fuel rod is obtained by subtracting the inherent background from the scattered pressure. And also analyzed are the effect of material damping of cladding tube and pellet on the resonance spectrum of a fuel rod. The propagation characteristics of circumferential waves which cause the resonances of cladding tube is produced and the appropriate resonance modes for the application to the inspection of assembled fuel rods are selected. The resonance modes are experimentally measured for pre- and post-irradiated fuel rods and the validation of the fuel rod inspection using ultrasonic resonance phenomenon is examined. And thin ultrasonic sensors accessible into the narrow interval (about 2-3mm) between assembled fuel rods are designed and manufactured. 14 refs. (Author).

  20. Nuclear fuel rod with burnable plate and pellet-clad interaction fix

    International Nuclear Information System (INIS)

    Boyle, R.F.

    1987-01-01

    This patent describes a nuclear fuel rod comprising a metallic tubular cladding containing nuclear fuel pellets, the pellets containing enriched uranium-235. The improvement described here comprises: ceramic wafers, each wafter comprising a sintered mixture of gadolinium oxide and uranium dioxide, the uranium oxide having no more uranium-235 than is present in natural uranium dioxide. Each of the wafers is axially disposed between a major portion of adjacent the nuclear fuel pellets, whereby the wafers freeze out volatile fission products produced by the nuclear fuel and prevent interaction of the fission products with the metallic tubing cladding

  1. The out-of-pile test for internal pressure measurement of nuclear fuel rod using LVDT

    Energy Technology Data Exchange (ETDEWEB)

    Son, J. M.; Kim, B. K.; Kim, D. S.; Joo, K. N.; Park, S. J.; Kang, Y. H.; Kim, Y. K.; Yeum, K. I. [KAERI, Taejon (Korea, Republic of)

    2002-05-01

    As a part of the development of instrumentation technologies for the nuclear fuel irradiation test in HANARO(High-flux Advanced Nuclear Application Reactor), the internal pressure measurement technique of the nuclear fuel rod is being developed using LVDT(Linear Variable Differential Transformer). The objectives of this test were to understand the LVDT's characteristics and to study its application techniques for fuel irradiation technology. It will be required to analyze the acquired internal pressure of fuel rod during fuel irradiation test in HANARO. Therefore, the out of pile test system for pressure measurement was developed, and the test with the LVDT at room temperature were performed. This test were implemented in 1 kg/cm{sup 2} increment from 1 kg/cm{sup 2} to 30 kg/cm{sup 2}, and repeated 6 times at same condition. The LVDT's sensitivities were obtained by following two ways, the one by test and the other by calculation from characteristics data. These two sensitivities were compared and analyzed. The calculation method for internal pressure of nuclear fuel rod at specified temperature was also established. The results of the out-of-pile test will be used to predict accurately the internal pressure of fuel rod during irradiation test. And, the well qualified out-of-pile tests are needed to understand the LVDT's detail characteristics at high temperature for the detail design of the fuel irradiation capsule.

  2. Experiments with preirradiated fuel rods in the Nuclear Safety Research Reactor

    International Nuclear Information System (INIS)

    Horiki, O.; Kobayashi, S.; Takariko, I.; Ishijima, K.

    1992-01-01

    In the Nuclear Safety Research Reactor (NSRR) owned and operated by Japan Atomic Energy Research Institute (JAERI), extensive experimental studies on the fuel behavior under reactivity initiated accident (RIA) conditions have been continued since the start of the test program in 1975. Accumulated experimental data were used as the fundamental data base of the Japanese safety evaluation guideline for reactivity initiated events in light water cooled nuclear power plants established by the nuclear safety commission in 1984. All of the data used to establish the guideline were, however, limited to those derived from the tests with fresh fuel rods as test samples because of the lack of experimental facility to handle highly radioactive materials.The guideline, therefore, introduces the peak fuel enthalpy of 85 cal/g which was adopted from the SPERT-CDC data as a provisional failure threshold of preirradiated fuel rod and, says that this value should be revised based on the NSRR experiments in the future. According to the above requirement, new NSRR experimental program with the preirradiated fuel rods as test samples was started in 1989. Test fuel rods are prepared by refabrication of the long-sized fuel rods preirradiated in commercial PWRs and BWRs into short segments and by preirradiation of short-sized test fuel rods in the Japan Material Testing Reactor(JMTR). For the tests with preirradiated fuel rods as test samples, the special experimental capsules, the automatic instrumentation fitting device, the automatic capsule assembling device and the capsule loading device were newly developed. In addition, the existing hot cave was modified to mount the capsule assembling device and the other inspection tools and, a new small iron cell was established adjacent to the cave to store the instrumentation fitting device. (author)

  3. The out-of-pile test for internal pressure measurement of nuclear fuel rod using LVDT

    Energy Technology Data Exchange (ETDEWEB)

    Min, Sohn Jae; Kang, Y. H.; Kim, B. G. [and others

    2001-11-01

    As a part of the development of instrumentation technologies for the nuclear fuel irradiation test in HANARO, the internal pressure measurement technique of the nuclear fuel rod is being developed using LVDT. The objectives of this test were to understand the LVDT's characteristics and to study its application techniques for fuel irradiation technology. It will be required to analyze the acquired internal pressure of fuel rod during fuel irradiation test in HANARO. The out-of-pile test system for pressure measurement was developed, and the test with the LVDT at room temperature(19 .deg. C) were performed. A out-of-pile test were implemented in 1 kg/cm{sup 2} increment from 1 kg/cm{sup 2} to 30 kg/cm{sup 2} and repeated 6 times at each condition. The LVDT's sensitivities were obtained by following two ways, the one by test and the other by calculation from characteristics data. These two sensitivities were compared and analyzed. The calculation method for internal pressure of nuclear fuel rod at specified temperature was also established. This report describes the system configuration, the out-of-pile test procedures, and the results. The results of the out-of-pile test will be used to predict accurately the internal pressure of fuel rod during irradiation test. And, the well qualified out-of-pile tests are needed to understand the LVDT's detail characteristics for the detail design of the fuel irradiation capsule.

  4. System for manipulating radioactive fuel rods within a nuclear fuel assembly

    International Nuclear Information System (INIS)

    Tolino, R.W.; King, W.E.; Blickenderfer, J.L.; Roth, C.H. Jr.

    1987-01-01

    A tool is described for manipulating the peripherally located fuel rods of a fuel assembly so that the rods can be visually inspected. The fuel assembly includes top and bottom nozzles, each of which is connected to a support skeleton, as well as grids, and wherein the rods are retained within the grids and confined between the top and bottom nozzles thereof. It consists of: (a) a fixture that is detachably connectable to one of the nozzles of the fuel assembly. The fixture having holes therein, (b) rotating means pivotally mountable within the holes of the fixture for selectively gripping and rotating the rod, and (c) a displacing means mounted on the fixture for reciprocably displacing the rods within the fuel assembly, including a lifting assembly and a push-down assembly for lifting and pushing down a selected one of the rods, respectively, whereby the rods can be selectively rotated, lifted, and pushed down in order to expose portions of the rods which are normally hidden to visual inspection while the nozzles stay connected to the support skeleton and the rods stay confined between the top and bottom nozzles of the fuel assembly

  5. The use of eddy current testing for nuclear fuel rods cladding evaluation

    International Nuclear Information System (INIS)

    Silva Junior, Silverio F. da; Alencar, Donizete A.; Brito, Mucio Jose D. de

    2007-01-01

    Nuclear fuel rods cladding must be tested after their manufacture and during their operational life. This paper describes a study about the use of eddy current test method as a nondestructive tool for nuclear fuel rods cladding evaluation. The experiments were carried out using two different probes: an external probe and an internal probe. The main goal was to verify the sensitivity of the eddy current test system, to develop calibration and reference standards and to establish the main capabilities and limitations presented by this test method for this application. (author)

  6. Device for replacing the rods of a fuel element of a nuclear reactor

    International Nuclear Information System (INIS)

    Nissel, B.; Kybranz, R.; Will, R.

    1977-01-01

    In order to be able to replace several separate rods (fuel rods or absorber rods), in a fuel element, a special grab is introduced, which consists of several individual gripping devices and is operated by spring loading. (TK) [de

  7. Method for automatic filling of nuclear fuel rod cladding tubes

    International Nuclear Information System (INIS)

    Bezold, H.

    1979-01-01

    Prior to welding the zirconium alloy cladding tubes with end caps, they are automatically filled with nuclear fuel tablets and ceramic insulating tablets. The tablets are introduced into magazine drums and led through a drying oven to a discharging station. The empty cladding tubes are removed from this discharging station and filled with tablets. A filling stamp pushes out the columns of tablets in the magazine tubes of the magazine drum into the cladding tube. Weight and measurement of length determine the filled state of the cladding tube. The cladding tubes are then led to the welding station via a conveyor belt. (DG) [de

  8. Fuel rod pellet loading head

    International Nuclear Information System (INIS)

    Howell, T.E.

    1975-01-01

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

  9. 3D finite element analysis of a nuclear fuel rod with gap elements between the pellet and the cladding

    International Nuclear Information System (INIS)

    Kang, Chang-Hak; Lee, Sung-Uk; Yang, Dong-Yol; Kim, Hyo-Chan; Yang, Yong-Sik

    2016-01-01

    Nuclear fuel rods which comprises an important component of a nuclear power plant are composed of nuclear fuel and cladding. Simulating the nuclear fuel rod using a computer program is the universal method to verify its safety. The computer program used for this is called the fuel performance code. The main objective of this study is to simulate the nuclear fuel rod behavior considering the gap conductance using three-dimensional gap elements. Gap elements are used because, unlike other methods, this approach does not require special methods or other variables such as the Lagrange multiplier. In this work, a nuclear fuel rod has been simulated and the results are compared with the experimental results. (author)

  10. Zirconium-based alloys, nuclear fuel rods and nuclear reactors including such alloys, and related methods

    Science.gov (United States)

    Mariani, Robert Dominick

    2014-09-09

    Zirconium-based metal alloy compositions comprise zirconium, a first additive in which the permeability of hydrogen decreases with increasing temperatures at least over a temperature range extending from 350.degree. C. to 750.degree. C., and a second additive having a solubility in zirconium over the temperature range extending from 350.degree. C. to 750.degree. C. At least one of a solubility of the first additive in the second additive over the temperature range extending from 350.degree. C. to 750.degree. C. and a solubility of the second additive in the first additive over the temperature range extending from 350.degree. C. to 750.degree. C. is higher than the solubility of the second additive in zirconium over the temperature range extending from 350.degree. C. to 750.degree. C. Nuclear fuel rods include a cladding material comprising such metal alloy compositions, and nuclear reactors include such fuel rods. Methods are used to fabricate such zirconium-based metal alloy compositions.

  11. Out pile test of a disassembly tool for the intermediate examination of nuclear fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Jintae; Joung, Chang-Young; Ahn, Sung-Ho; Yang, Tae-Ho; Jang, Seo-Yoon; Park, Seung-Jae [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    The two fuel rod assemblies are assembled with a bayonet coupler, and the non-instrumented fuel rod assembly can be disassembled for intermediate examination. A tool to disassemble the non-instrumented fuel rod assembly from the test rig was developed, and steel wires are connected to the tool to operate release function. In this study, an assembly plug with a quick plug typed bayonet coupler and the accompanying disassembly tool was designed to prevent the interference problem. A test rig mockup was fabricated, and performance test was carried out in the laboratory. And, the out pile test was also carried out in the single channel test loop established in the KAERI. In this study, a modified coupler design to disassemble the non-instrumented fuel rod assembly from the test rig for the intermediate examination was suggested to solve interference problem of previous design. The performance of the modified design was verified by test mockup fabricated with the modified coupler design and accompanied disassembly tool design. Finally, out pile test was carried out in the single channel test loop in the KAERI, and the test rig and the disassembly tool showed good performance and reliability. The developed technique will be useful to the periodic intermediate examination of nuclear fuel rods.

  12. Out pile test of a disassembly tool for the intermediate examination of nuclear fuel rods

    International Nuclear Information System (INIS)

    Hong, Jintae; Joung, Chang-Young; Ahn, Sung-Ho; Yang, Tae-Ho; Jang, Seo-Yoon; Park, Seung-Jae

    2016-01-01

    The two fuel rod assemblies are assembled with a bayonet coupler, and the non-instrumented fuel rod assembly can be disassembled for intermediate examination. A tool to disassemble the non-instrumented fuel rod assembly from the test rig was developed, and steel wires are connected to the tool to operate release function. In this study, an assembly plug with a quick plug typed bayonet coupler and the accompanying disassembly tool was designed to prevent the interference problem. A test rig mockup was fabricated, and performance test was carried out in the laboratory. And, the out pile test was also carried out in the single channel test loop established in the KAERI. In this study, a modified coupler design to disassemble the non-instrumented fuel rod assembly from the test rig for the intermediate examination was suggested to solve interference problem of previous design. The performance of the modified design was verified by test mockup fabricated with the modified coupler design and accompanied disassembly tool design. Finally, out pile test was carried out in the single channel test loop in the KAERI, and the test rig and the disassembly tool showed good performance and reliability. The developed technique will be useful to the periodic intermediate examination of nuclear fuel rods

  13. A model for gap conductance in nuclear fuel rods

    International Nuclear Information System (INIS)

    Loyalka, S.K.

    1982-01-01

    Computation of nuclear reactor fuel behavior under normal and off-normal conditions is influenced by gap conductance models. These models should provide accurate results for heat transfer for arbitrary gap widths and gas mixtures and should be based on considerations of the kinetic theory of gases. There has been considerable progress in the study of heat transfer in a simple gas for arbitrary Knudsen numbers (Kn = l/similar to d, where l is a meanfree-path and similar d is the gap width) in recent years. Using these recent results, a simple expression for heat transfer in a gas mixture (enclosed between parallel plates) for an arbitrary Knudsen number has been constructed, and a new model for gap conductance has been proposed. The latter reproduces the free molecular (small gap, Kn >> 1) and the jump limits (large gaps, Kn << 1) correctly, and it provides fairly accurate results for arbitrary gap widths. The new model is suitable for use in large fuel behavior computer programs

  14. Procyon 1. First prototype worldwide for storage spent nuclear fuel rods

    International Nuclear Information System (INIS)

    Meyering, Manfred

    2010-01-01

    HFH Herbst has designed and built a unique machine for storage of spent highly radioactive nuclear fuel rods within two years for the Swedish SKB. The vehicle (total weight 98 t) can be operated underground without a driver. Herbst was able to bring to this project almost 30 years of experience in the complementation of vehicle projects for the nuclear industry. The Procyon 1 already proved its efficiency impressively in several hundred storage processes and operates with absolute reliability. (orig.)

  15. Fuel rod leak detector

    International Nuclear Information System (INIS)

    Womack, R.E.

    1978-01-01

    A typical embodiment of the invention detects leaking fuel rods by means of a radiation detector that measures the concentration of xenon-133 ( 133 Xe) within each individual rod. A collimated detector that provides signals related to the energy of incident radiation is aligned with one of the ends of a fuel rod. A statistically significant sample of the gamma radiation (γ-rays) that characterize 133 Xe is accumulated through the detector. The data so accumulated indicates the presence of a concentration of 133 Xe appropriate to a sound fuel rod, or a significantly different concentration that reflects a leaking fuel rod

  16. Heat transfer coefficient testing in nuclear fuel rod bundles with mixing vane grids

    International Nuclear Information System (INIS)

    Conner, Michael E.; Smith, L. David III; Holloway, Mary V.; Beasley, Donald E.

    2005-01-01

    An air heat transfer test facility was developed to test the heat transfer downstream of support grids in simulated PWR nuclear fuel rod bundles. The goal of this testing is to study the single-phase heat transfer coefficients downstream of grids with mixing vanes in a square-pitch rod bundle. The technique developed utilizes fully-heated grid spans and a specially designed thermocouple holder that can be moved axially down the rod bundle and aximuthally within a test rod. From this testing, the axial and aximuthally varying heat transfer coefficient can be determined. Different grid designs are tested and compared to determine the heat transfer enhancement associated with key grid features such as mixing vanes. (author)

  17. Failed fuel rod detector

    Energy Technology Data Exchange (ETDEWEB)

    Uchida, Katsuya; Matsuda, Yasuhiko

    1984-05-02

    The purpose of the project is to enable failed fuel rod detection simply with no requirement for dismantling the fuel assembly. A gamma-ray detection section is arranged so as to attend on the optional fuel rods in the fuel assembly. The fuel assembly is adapted such that a gamma-ray shielding plate is detachably inserted into optional gaps of the fuel rods or, alternatively, the fuel assembly can detachably be inserted to the gamma-ray shielding plate. In this way, amount of gaseous fission products accumulated in all of the plenum portions in the fuel rods as the object of the measurement can be determined without dismantling the fuel assembly. Accordingly, by comparing the amounts of the gaseous fission products, the failed fuel rod can be detected.

  18. Fuel rod fixing system

    International Nuclear Information System (INIS)

    Christiansen, D.W.

    1982-01-01

    This is a reusable system for fixing a nuclear reactor fuel rod to a support. An interlock cap is fixed to the fuel rod and an interlock strip is fixed to the support. The interlock cap has two opposed fingers, which are shaped so that a base is formed with a body part. The interlock strip has an extension, which is shaped so that this is rigidly fixed to the body part of the base. The fingers of the interlock cap are elastic in bending. To fix it, the interlock cap is pushed longitudinally on to the interlock strip, which causes the extension to bend the fingers open in order to engage with the body part of the base. To remove it, the procedure is reversed. (orig.) [de

  19. Fuel rod attachment system

    International Nuclear Information System (INIS)

    Christiansen, D.W.

    1982-01-01

    A reusable system for removably attaching a nuclear reactor fuel rod to a support member. A locking cap is secured to the fuel rod and a locking strip is fastened to the support member or vice versa. The locking cap has two opposing fingers and shaped to form a socket having a body portion. The locking strip has an extension shaped to rigidly attach to the socket's body portion. The locking cap's fingers are resiliently deflectable. For attachment, the locking cap is longitudinally pushed onto the locking strip causing the extension to temporarily deflect open the fingers to engage the socket's body portion. For removal, the process is reversed. In an alternative embodiment, the cap is rigid and the strip is transversely resiliently compressible. (author)

  20. Heat Transfer Enhancement By Three-Dimensional Surface Roughness Technique In Nuclear Fuel Rod Bundles

    Science.gov (United States)

    Najeeb, Umair

    This thesis experimentally investigates the enhancement of single-phase heat transfer, frictional loss and pressure drop characteristics in a Single Heater Element Loop Tester (SHELT). The heater element simulates a single fuel rod for Pressurized Nuclear reactor. In this experimental investigation, the effect of the outer surface roughness of a simulated nuclear rod bundle was studied. The outer surface of a simulated fuel rod was created with a three-dimensional (Diamond-shaped blocks) surface roughness. The angle of corrugation for each diamond was 45 degrees. The length of each side of a diamond block is 1 mm. The depth of each diamond block was 0.3 mm. The pitch of the pattern was 1.614 mm. The simulated fuel rod had an outside diameter of 9.5 mm and wall thickness of 1.5 mm and was placed in a test-section made of 38.1 mm inner diameter, wall thickness 6.35 mm aluminum pipe. The Simulated fuel rod was made of Nickel 200 and Inconel 625 materials. The fuel rod was connected to 10 KW DC power supply. The Inconel 625 material of the rod with an electrical resistance of 32.3 kO was used to generate heat inside the test-section. The heat energy dissipated from the Inconel tube due to the flow of electrical current flows into the working fluid across the rod at constant heat flux conditions. The DI water was employed as working fluid for this experimental investigation. The temperature and pressure readings for both smooth and rough regions of the fuel rod were recorded and compared later to find enhancement in heat transfer coefficient and increment in the pressure drops. Tests were conducted for Reynold's Numbers ranging from 10e4 to 10e5. Enhancement in heat transfer coefficient at all Re was recorded. The maximum heat transfer co-efficient enhancement recorded was 86% at Re = 4.18e5. It was also observed that the pressure drop and friction factor increased by 14.7% due to the increased surface roughness.

  1. ORNL capability to conduct post irradiation examination of full-length commercial nuclear fuel rods

    International Nuclear Information System (INIS)

    Spellman, Donald J.

    2007-01-01

    Hot cells at Oak Ridge National Laboratory (ORNL) are nearing completion of a multi-year upgrade program to implement 21. century capabilities to meet the examination demands for higher burnup fuels and the future demands that will come from fuel recycling programs. Fuel reliability and zero tolerance for fuel failure is more than an industry goal. Fuel reliability is becoming a requirement that supports the renaissance of nuclear power generation. Thus, fuel development and management of new forms of waste that will come from programs such as the Global Nuclear Energy Partnership (GNEP) will require extensive use of the flexible, high-quality, technically advanced hot cells at ORNL. ORNL has the capability to perform post irradiation examination (PIE) of irradiated commercial nuclear fuel rods and the management structure to ensure a timely, cost-effective result. ORNL can: 1) Handle the transportation issues, 2) Perform macroscopic fuel rod examinations, 3) Perform microscopic fuel and clad examinations, and 4) Manage legacy material and waste disposal issues from PIE activities. All four of these items will be managed in a way that allows the customer day-to-day access to the results and data. Hot cell examination equipment that is necessary to determine the characteristics and performance of irradiated materials must operate in a hostile environment and is subject to long-term degradation that may result in reliability and quality assurance (QA) issues. ORNL has modernized its hot cell nuclear fuel examination equipment, installing state-of-the-art automated examination equipment and data gathering capabilities. ORNL is planning a major commitment to nuclear fuel examination and development, and future improvements will continue to be made over the next few years. (author)

  2. Conservative performance analysis of a PWR nuclear fuel rod using the FRAPCON code

    Energy Technology Data Exchange (ETDEWEB)

    Oliveira, Fabio Branco Vaz de; Sabundjian, Gaiane, E-mail: fabio@ipen.br, E-mail: gdjian@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2015-07-01

    In this paper, some of the preliminary results of the sensitivity and conservative analysis of a hypothetical pressurized water reactor fuel rod are presented, using the FRAPCON code as a basic and preparation tool for the future transient analysis, which will be carried out by the FRAPTRAN code. Emphasis is given to the evaluation of the cladding behavior, since it is one of the critical containment barriers of the fission products, generated during fuel irradiation. Sensitivity analyses were performed by the variation of the values of some parameters, which were mainly related with thermal cycle conditions, and taking into account an intermediate value between the realistic and conservative conditions for the linear heat generation rate parameter, given in literature. Time lengths were taken from typical nuclear power plant operational cycle, adjusted to the obtention of a chosen burnup. Curves of fuel and cladding temperatures, and also for their mechanical and oxidation behavior, as a function of the reactor operation's time, are presented for each one of the nodes considered, over the nuclear fuel rod. Analyzing the curves, it was possible to observe the influence of the thermal cycle on the fuel rod performance, in this preliminary step for the accident/transient analysis. (author)

  3. Spacers for fuel rod clusters

    International Nuclear Information System (INIS)

    Jabsen, F.S.

    1978-01-01

    The proposition deals with the fixing of nuclear fuel element rods in a grid which consists of a number of crossed Zy-plates which form cells. The rectangular cells have projections which serve as spacers for the fuel rods. According to the invention there are additional butt straps which can be moved in such a way that insertion and extraction of the fuel rods can be done without obstruction and they can be spring-loaded hold in their final position. (UWI) [de

  4. Method and system for the production of fuel rods for nuclear reactors

    International Nuclear Information System (INIS)

    Meinhardt, V.; Schultz, R.; Gall, A.

    1975-01-01

    The invention deals with a method to produce nuclear reactor fuel rods which contain the fuel in a cladding tube and in which an alkali metal is put in the space between the fuel and cladding for better heat transfer from the fuel to the cladding. The bubble-free filling-in of the alkali metal causes difficulties; the invention suggests to carry out the filling-in of fuel and alkali metal in a horizontal position of the cladding to rinse through the free space firstly with alkali metal, then to fill with fuel and then with the alkali metal taking account of necessary space for the fission gas volume. Equipment to carry out this process in which all operations are carried out in a vacuum box is described. (UWI) [de

  5. Prototypical spent nuclear fuel rod consolidation equipment: Phase 2, Final design report: Volume 4, Appendices: Part 3

    International Nuclear Information System (INIS)

    Ciez, A.P.

    1987-01-01

    The purpose of this manual is to provide assembly, installation, operation, maintenance, and off-normal recovery procedures for the Consolidation Equipment. The Consolidation System is a horizontal, dry system capable of processing one Pressurized Water Reactor (PWR) fuel assembly or one Boiling Water Reactor (BWR) fuel assembly at a time. The system will process all spent PWR and BWR fuels from the commercial US nuclear power reactor industry. Component changeouts for various fuel types have been minimized to reduce costs, required in-cell module storage space, and to increase efficiency by decreasing set-up time between fuel consolidation campaigns. The most important feature of the Westinghouse system is the ability to control the fuel rods at all times during the consolidation process from rod extraction, through canister loading. This features assures that the rods from two PWR fuel assemblies or four BWR fuel assemblies (minimum) can be loaded into one consolidated rods canister

  6. Sturdy on Orbital TIG Welding Properties for Nuclear Fuel Test Rod

    Energy Technology Data Exchange (ETDEWEB)

    Joung, Changyoung; Hong, Jintae; Kim, Kahye; Huh, Sungho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    We developed a precision TIG welding system that is able to weld the seam between end-caps and a fuel cladding tube for the nuclear fuel test rod and rig. This system can be mainly classified into an orbital TIG welder (AMI, M-207A) and a pressure chamber. The orbital TIG welder can be independently used, and it consists of a power supply unit, a microprocessor, water cooling unit, a gas supply unit and an orbital weld head. In this welder, the power supply unit mainly supplies GTAW power for a welding specimen and controls an arc starting of high frequency, supping of purge gas, arc rotation through the orbital TIG welding head, and automatic timing functions. In addition, the pressure chamber is used to make the welded surface of the cladding specimen clean with the inert gas filled inside the chamber. To precisely weld the cladding tube, a welding process needs to establish a schedule program for an orbital TIG welding. Therefore, the weld tests were performed on a cladding tube and dummy rods under various conditions. This paper describes not only test results on parameters of the purge gas flow rates and the chamber gas pressures for the orbital TIG welding, but also test results on the program establishment of an orbital TIG welding system to weld the fuel test rods. Various welding tests were performed to develop the orbital TIG welding techniques for the nuclear fuel test rod. The width of HAZ of a cladding specimen welded with the identical power during an orbital TIG welding cycle was continuously increased from a welded start-point to a weld end-point because of heat accumulation. The welding effect of the PGFR and CGP shows a relatively large difference for FSS and LSS. Each hole on the cladding specimens was formed in the 1bar CGP with the 20L/min PGFR but not made in the case of the PGFR of 10L/min in the CGP of 2bar. The optimum schedule program of the orbital TIG welding system to weld the nuclear fuel test rod was established through the program

  7. Sturdy on Orbital TIG Welding Properties for Nuclear Fuel Test Rod

    International Nuclear Information System (INIS)

    Joung, Changyoung; Hong, Jintae; Kim, Kahye; Huh, Sungho

    2014-01-01

    We developed a precision TIG welding system that is able to weld the seam between end-caps and a fuel cladding tube for the nuclear fuel test rod and rig. This system can be mainly classified into an orbital TIG welder (AMI, M-207A) and a pressure chamber. The orbital TIG welder can be independently used, and it consists of a power supply unit, a microprocessor, water cooling unit, a gas supply unit and an orbital weld head. In this welder, the power supply unit mainly supplies GTAW power for a welding specimen and controls an arc starting of high frequency, supping of purge gas, arc rotation through the orbital TIG welding head, and automatic timing functions. In addition, the pressure chamber is used to make the welded surface of the cladding specimen clean with the inert gas filled inside the chamber. To precisely weld the cladding tube, a welding process needs to establish a schedule program for an orbital TIG welding. Therefore, the weld tests were performed on a cladding tube and dummy rods under various conditions. This paper describes not only test results on parameters of the purge gas flow rates and the chamber gas pressures for the orbital TIG welding, but also test results on the program establishment of an orbital TIG welding system to weld the fuel test rods. Various welding tests were performed to develop the orbital TIG welding techniques for the nuclear fuel test rod. The width of HAZ of a cladding specimen welded with the identical power during an orbital TIG welding cycle was continuously increased from a welded start-point to a weld end-point because of heat accumulation. The welding effect of the PGFR and CGP shows a relatively large difference for FSS and LSS. Each hole on the cladding specimens was formed in the 1bar CGP with the 20L/min PGFR but not made in the case of the PGFR of 10L/min in the CGP of 2bar. The optimum schedule program of the orbital TIG welding system to weld the nuclear fuel test rod was established through the program

  8. Operational experience gained with the failed fuel rod detection system in nuclear power plants

    International Nuclear Information System (INIS)

    Boehm, H.H.; Forch, H.

    1985-01-01

    Brown Boveri Reaktor GmbH together with Krautkramer Company developed such a FAILED FUEL ROD DETECTION SYSTEM (FFRDS) which allows to located defective fuel rods without dismantling the fuel assembly or pulling of individual rods. Since 1979 the FFRDS is employed successfully in various nuclear power plants in Europe, USA, Japan, and Korea. The short inspection time and the high reliability of the method make the FFRDS a true competitor to the sipping method. In this paper the authors discuss the method and the design of the system, the equipment set-up, its features and the experience gained so far. The system has been performed and automated to such an extent that within a short installation period series of fuel assemblies can be tested with relatively short intervals of time (5 minutes for BWR and 7 minutes for PWR fuel assemblies per side). The ability of the system for deployment under various conditions and the experience gained during the past six years have made this system universally applicable and highly sensitive to the requirements of NDT during outages and for transport of FAs to intermediate storage facilities. Comparison of FFRDS to conventional sipping has indicated in several instances that the FFRDS is superior to the latter technique

  9. External attachment of titanium sheathed thermocouples to zirconium nuclear fuel rods for the LOFT reactor

    International Nuclear Information System (INIS)

    Welty, R.K.

    1980-01-01

    The Exxon Nuclear Company, Inc., acting as a Subcontractor to EG and G Idaho Inc., Idaho National Engineering Laboratory, Idaho Falls, Idaho, has developed a welding process to attach titanium sheathed thermocouples to the outside of the zircaloy clad fuel rods. The fuel rods and thermocouples are used to test simulated loss-of-coolant accident (LOCA) conditions in a pressurized water reactor (LOFT Reactor, Idaho National Laboratory). A laser beam was selected as the optimum welding process because of the extremely high energy input per unit volume that can be achieved allowing local fusion of a small area irrespective of the difference in material thickness to be joined. A commercial pulsed laser and energy control system was installed along with specialized welding fixtures. Laser room facility requirements and tolerances were established. Performance qualifications, and detailed welding procedures were also developed. Product performance tests were conducted to assure that engineering design requirements could be met on a production basis

  10. Results of the first nuclear blowdown test on single fuel rods (LOC-11 Series in PBF)

    International Nuclear Information System (INIS)

    Larson, J.R.; Evans, D.R.; McCardell, R.K.

    1978-01-01

    This paper presents results of the first nuclear blowdown tests (LOC-11A, LOC-11B, LOC-11C) ever conducted. The Loss-of-Coolant Accident (LOCA) Test Series is being conducted in the Power Burst Facility (PBF) reactor at the Idaho National Engineering Laboratory, near Idaho Falls, Idaho, for the Nuclear Regulatory Commission. The objective of the LOC-11 tests was to obtain data on the behavior of pressurized and unpressurized rods when exposed to a blowdown similar to that expected in a pressurized water reactor (PWR) during a hypothesized double-ended cold-leg break. The data are being used for the development and verification of analytical models that are used to predict coolant and fuel rod pressure during a LOCA in a PWR

  11. Method for verifying the pressure in a nuclear reactor fuel rod

    International Nuclear Information System (INIS)

    Jones, W.J.

    1979-01-01

    Disclosed is a method of accurately verifying the pressure contained in a sealed pressurized fuel rod by utilizing a pressure balance measurement technique wherein an end of the fuel rod extends through and is sealed in a wall of a small chamber. The chamber is pressurized to the nominal (desired) fuel rod pressure and the fuel rod is then pierced to interconnect the chamber and fuel rod. The deviation of chamber pressure is noted. The final combined pressure of the fuel rod and drill chamber is substantially equal to the nominal rod pressure; departure of the combined pressure from nominal is in direct proportion to departure of rod pressure from nominal. The maximum error in computing the rod pressure from the deviation of the combined pressure from nominal is estimated at plus or minus 3.0 psig for rod pressures within the specified production limits. If the rod pressure is corrected for rod void volume using a digital printer data record, the accuracy improves to about plus or minus 2.0 psig

  12. Tomography on nuclear fuel rods in the nuclear power plant of Dodewaard

    International Nuclear Information System (INIS)

    Tanke, R.H.J.; Jaspers, J.E.; Gaalman, P.A.M.

    1990-01-01

    This report discusses the feasibility of using emission tomography on fuel rods in the Dodewaard reactor. The tomography can be used to increase the efficiency of the use of fissionable material. (R.A.B.). 4 refs.; 17 figs.; 1 tab

  13. PWR Fuel licensing in France - from design to reprocessing: licensing of nuclear PWR fuel rod design to satisfy with criteria for normal and abnormal fuel operation in France

    International Nuclear Information System (INIS)

    Beraha, R.

    1999-01-01

    In this lecture are presented: French regulatory context; Current fuel management methods; Request from the french operator EdF; Most recent actions of the french Nuclear safety authority; Fuel assemblies deformations (impact of high burn-up; investigations during reactor's exploitation; control rods drop off times)

  14. Gas pressure and gas purity analyzing device in nuclear fuel rod

    International Nuclear Information System (INIS)

    Mizutani, Chihiro; Hasegawa, Toru.

    1996-01-01

    The present invention provides a device for measuring and analyzing a pressure and a purity of a helium gas sealed in a BWR type nuclear fuel rod. Namely, a portion between a rotational shaft of an electromotive drill for perforating the fuel rod and a vacuum chamber is sealed with a magnetic fluid sealing material so that error factors can be recognized before and after the destruction detection (perforation) of a fuel rod. With such procedures, involving of an atmospheric air from the drill rotational shaft upon perforation can be eliminated. As a result, accuracy for the measurement can be improved. In addition, a filter is disposed to a pipeline connecting the vacuum chamber and the measuring system. With such a constitution, scattering of cutting dusts to the measuring system, troubles due to damages of a stop valve can be reduced. As a result, the efficiency of the measurement is improved. Further, a plurality kinds of gas collecting vessel having different capacities are connected in parallel to the pipeline of the measuring system. Then, the gas collecting vessels can be used selectively. As a result, the device can cope with a gas pressure over a wide range. (I.S.)

  15. Fuel rod technology

    International Nuclear Information System (INIS)

    Bezold, H.; Romeiser, H.J.

    1979-07-01

    By extensive mechanization and automation of the fuel rod production, also at increasing production numbers, an efficient production shall be secured, simultaneously corresponding to the high quality standard of the fuel rods. The works done up to now concentrated on the lay out of a rough concept for a mechanized production course. Detail-studies were made for the problems of fuel rod humidity, filling and resistance welding. Further promotion of this project and thus further report will be stopped, since the main point of these works is the production technique. (orig.) [de

  16. Three-dimensional FE analysis of the thermal-mechanical behaviors in the nuclear fuel rods

    International Nuclear Information System (INIS)

    Jiang Yijie; Cui Yi; Huo Yongzhong; Ding Shurong

    2011-01-01

    Highlights: → We establish three-dimensional finite element models for nuclear fuel rods. → The thermal-mechanical behaviors at the initial stage of burnup are obtained. → Several parameters on the in-pile performances are investigated. → The parameters have remarkable effects on the in-pile behaviors. → This study lays a foundation for optimal design and irradiation safety. - Abstract: In order to implement numerical simulation of the thermal-mechanical behaviors in the nuclear fuel rods, a three-dimensional finite element model is established. The thermal-mechanical behaviors at the initial stage of burnup in both the pellet and the cladding are obtained. Comparison of the obtained numerical results with those from experiments validates the developed finite element model. The effects of the constraint conditions, several operation and structural parameters on the thermal-mechanical performances of the fuel rod are investigated. The research results indicate that: (1) with increasing the heat generation rates from 0.15 to 0.6 W/mm 3 , the maximum temperature within the pellet increases by 99.3% and the maximum radial displacement at the outer surface of the pellet increases by 94.3%. And the maximum Mises stresses in the cladding all increase; while the maximum values of the first principal stresses within the pellet decrease as a whole; (2) with increasing the heat transfer coefficients between the cladding and the coolant, the internal temperatures reduce and the temperature gradient remains similar; when the heat transfer coefficient is lower than a critical value, the temperature change is sensitive to the heat transfer coefficient. The maximum temperature increases only 7.13% when h changes from 0.5 W/mm 2 K to 0.01 W/mm 2 K, while increases up to 54.7% when h decreases from 0.01 W/mm 2 K to 0.005 W/mm 2 K; (3) the initial gap sizes between the pellet and the cladding significantly affect the thermal-mechanical behaviors in the fuel rod; when the

  17. Prototypical spent nuclear nuclear fuel rod consolidation equipment, Phase 2: Final design report: Volume 2, Appendices: Part 1

    International Nuclear Information System (INIS)

    Ciez, A.P.

    1987-01-01

    The purpose of this specification is to establish functional and design requirements for the Prototypical Spent Nuclear Fuel Rod Consolidation System. The Department of Energy-Idaho Operations Office (DOE-ID) is responsible for the implementation of the Prototypic Dry Rod Consolidation Demonstration Project. This program is to develop and demonstrate a fully qualified, licensable, cost-effective, dry spent fuel rod consolidation system by July 1989. The work is divided into four phases as follows: Phase I--Preliminary Design, Phase II--Final Design Option, Phase III--Fabrication and System Checkout Option, and Phase IV--Installation and Hot Demonstration Option. This specification is part of the Phase II effort. The objectives of this specification are to provide functional and design requirements for the Prototypical Spent Nuclear Fuel Rod Consolidation equipment; establish specific tool and subsystem requirements such that the integrated and overall system requirements are satisfied; and establish positioning, envelope and size interface control requirements for each tool or subsystem such that the individual components will interface properly with the overall system design

  18. Development of Mechanical Sealing and Laser Welding Technology to Instrument Thermocouple for Nuclear Fuel Test Rod

    International Nuclear Information System (INIS)

    Joung, Chang-Young; Ahn, Sung-Ho; Hong, Jin-Tae; Kim, Ka-Hye; Huh, Sung-Ho

    2015-01-01

    Zircaloy-4 of the nuclear fuel test rod, AISI 316L of the mechanical sealing parts, and the MI (mineral insulated) cable at a thermocouple instrumentation are hetero-metals, and are difficult to weld to dissimilar materials. Therefore, a mechanical sealing method to instrument the thermocouple should be conducted using two kinds of sealing process as follows: One is a mechanical sealing process using Swagelok, which is composed of sealing components that consists of an end-cap, a seal tube, a compression ring and a Swagelok nut. The other is a laser welding process used to join a seal tube, and an MI cable, which are made of the same material. The mechanical sealing process should be sealed up with the mechanical contact compressed by the strength forced between a seal tube and an end-cap, and the laser welding process should be conducted to have no defects on the sealing area between a seal tube and an MI cable. Therefore, the mechanical sealing and laser welding techniques need to be developed to accurately measure the centerline temperature of the nuclear fuel test rod in an experimental reactor. The mechanical sealing and laser welding tests were conducted to develop the thermocouple instrumentation techniques for the nuclear fuel test rod. The optimum torque value of a Swagelok nut to seal the mechanical sealing part between the end-cap and seal tube was established through various torque tests using a torque wrench. The optimum laser welding conditions to seal the welding part between a seal tube and an MI cable were obtained through various welding tests using a laser welding system

  19. Development of Mechanical Sealing and Laser Welding Technology to Instrument Thermocouple for Nuclear Fuel Test Rod

    Energy Technology Data Exchange (ETDEWEB)

    Joung, Chang-Young; Ahn, Sung-Ho; Hong, Jin-Tae; Kim, Ka-Hye; Huh, Sung-Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    Zircaloy-4 of the nuclear fuel test rod, AISI 316L of the mechanical sealing parts, and the MI (mineral insulated) cable at a thermocouple instrumentation are hetero-metals, and are difficult to weld to dissimilar materials. Therefore, a mechanical sealing method to instrument the thermocouple should be conducted using two kinds of sealing process as follows: One is a mechanical sealing process using Swagelok, which is composed of sealing components that consists of an end-cap, a seal tube, a compression ring and a Swagelok nut. The other is a laser welding process used to join a seal tube, and an MI cable, which are made of the same material. The mechanical sealing process should be sealed up with the mechanical contact compressed by the strength forced between a seal tube and an end-cap, and the laser welding process should be conducted to have no defects on the sealing area between a seal tube and an MI cable. Therefore, the mechanical sealing and laser welding techniques need to be developed to accurately measure the centerline temperature of the nuclear fuel test rod in an experimental reactor. The mechanical sealing and laser welding tests were conducted to develop the thermocouple instrumentation techniques for the nuclear fuel test rod. The optimum torque value of a Swagelok nut to seal the mechanical sealing part between the end-cap and seal tube was established through various torque tests using a torque wrench. The optimum laser welding conditions to seal the welding part between a seal tube and an MI cable were obtained through various welding tests using a laser welding system.

  20. An electrical simulator of a nuclear fuel rod cooled by nucleate boiling

    International Nuclear Information System (INIS)

    Costa, Antonio Carlos Lopes da; Machado, Luiz; Koury, Ricardo Nicolau Nassar; Passos, Julio Cesar

    2009-01-01

    This study investigates an electrical heated test section designed to simulate a nuclear fuel rod. This simulator comprises a stainless steel vertical tube, with length and outside diameter of 600 mm and 10 mm, respectively, inside which there is a high power electrical resistor. The heat generated is removed by means of enhanced confined subcooled nucleate boiling of water in an annular space containing 153 small metal inclined discs. The tests were performed under electrical power and pressure up to 48 kW and 40 bar, respectively. The results show that the experimental boiling heat transfer coefficients are in good agreement with those calculated using the Jens-Lottes correlation. (author)

  1. An electrical simulator of a nuclear fuel rod cooled by nucleate boiling

    Energy Technology Data Exchange (ETDEWEB)

    Costa, Antonio Carlos Lopes da [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)], e-mail: aclc@cdtn.br; Machado, Luiz; Koury, Ricardo Nicolau Nassar [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Dept. de Engenharia Mecanica], e-mail: luizm@demec.ufmg.br; Bonjour, Jocelyn [CETHIL, UMR5008, CNRS, INSA-Lyon (France)], e-mail: jocelyn.bonjour@insa-lyon.fr; Passos, Julio Cesar [Universidade Federal de Santa Catarina (UFSC), Florianopolis, SC (Brazil). Dept. de Engenharia Mecanica. LEPTEN/Boiling], e-mail: jpassos@emc.ufsc.br

    2009-07-01

    This study investigates an electrical heated test section designed to simulate a nuclear fuel rod. This simulator comprises a stainless steel vertical tube, with length and outside diameter of 600 mm and 10 mm, respectively, inside which there is a high power electrical resistor. The heat generated is removed by means of enhanced confined subcooled nucleate boiling of water in an annular space containing 153 small metal inclined discs. The tests were performed under electrical power and pressure up to 48 kW and 40 bar, respectively. The results show that the experimental boiling heat transfer coefficients are in good agreement with those calculated using the Jens-Lottes correlation. (author)

  2. Simulation of leaking fuel rods

    International Nuclear Information System (INIS)

    Hozer, Z.

    2006-01-01

    The behaviour of failed fuel rods includes several complex phenomena. The cladding failure initiates the release of fission product from the fuel and in case of large defect even urania grains can be released into the coolant. In steady state conditions an equilibrium - diffusion type - release is expected. During transients the release is driven by a convective type leaching mechanism. There are very few experimental data on leaking WWER fuel rods. For this reason the activity measurements at the nuclear power plants provide very important information. The evaluation of measured data can help in the estimation of failed fuel rod characteristics and the prediction of transient release dynamics in power plant transients. The paper deals with the simulation of leaking fuel rods under steady state and transient conditions and describes the following new results: 1) A new algorithm has been developed for the simulation of leaking fuel rods under steady state conditions and the specific parameters of the model for the Paks NPP has been determined; 2) The steady state model has been applied to calculation of leaking fuel characteristics using iodine and noble gas activity measurement data; 3) A new computational method has been developed for the simulation of leaking fuel rods under transient conditions and the specific parameters for the Paks NPP has been determined; 4) The transient model has been applied to the simulation of shutdown process at the Paks NPP and for the prediction of the time and magnitude of 123 I activity peak; 5) Using Paks NPP data a conservative value has been determined for the upper limit of the 123 I release from failed fuel rods during transients

  3. Fuel enrichment and temperature distribution in nuclear fuel rod in (D-T) driven hybrid reactor system

    Energy Technology Data Exchange (ETDEWEB)

    Osman, Ypek [Suleyman Demirel Universitesi Muhendislik-Mimarlyk Fakultesi, Isparta (Turkey)

    2001-07-01

    In this study, melting point of the fuel rod and temperature distribution in nuclear fuel rod are investigated for different coolants under various first wall loads (P{sub w}, =5, 6, 7, 8, 9, and 10 MWm{sup -2}) in Fusion-Fission reactor fueled with 50%LWR +50%CANDU. The fusion source of neutrons of 14.1 MeV is simulated by a movable target along the main axis of cylindrical geometry as a line source. In addition, the fusion chamber was thought as a cylindrical cavity with a diameter of 300 cm that is comparatively small value. The fissile fuel zone is considered to be cooled with four different coolants, gas, flibe (Li{sub 2}BeF{sub 4}), natural lithium (Li), and eutectic lithium (Li{sub 17}Pb{sub 83}). Investigations are observed during 4 years for discrete time intervals of{delta}t= 0.5 month and by a plant factor (PF) of 75%. Volumetric ratio of coolant-to fuel is 1:1, 45.515% coolant, 45.515% fuel, 8.971% clad, in fuel zone. (author)

  4. FY15 Status Report: CIRFT Testing of Spent Nuclear Fuel Rods from Boiler Water Reactor Limerick

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wang, Hong [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jiang, Hao [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-06-01

    The objective of this project is to perform a systematic study of used nuclear fuel (UNF, also known as spent nuclear fuel [SNF]) integrity under simulated transportation environments using the Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT) hot-cell testing technology developed at Oak Ridge National Laboratory (ORNL) in August 2013. Under Nuclear Regulatory Commission (NRC) sponsorship, ORNL completed four benchmark tests, four static tests, and twelve dynamic or cycle tests on H. B. Robinson (HBR) high burn-up (HBU) fuel. The clad of the HBR fuels was made of Zircaloy-4. Testing was continued in fiscal year (FY) 2014 using Department of Energy (DOE) funds. The additional CIRFT was conducted on three HBR rods (R3, R4, and R5) in which two specimens failed and one specimen was tested to over 2.23 10⁷ cycles without failing. The data analysis on all the HBR UNF rods demonstrated that it is necessary to characterize the fatigue life of the UNF rods in terms of (1) the curvature amplitude and (2) the maximum absolute of curvature extremes. The maximum extremes are significant because they signify the maximum of tensile stress for the outer fiber of the bending rod. CIRFT testing has also addressed a large variation in hydrogen content on the HBR rods. While the load amplitude is the dominant factor that controls the fatigue life of bending rods, the hydrogen content also has an important effect on the lifetime attained at each load range tested. In FY 15, ten SNF rod segments from BWR Limerick were tested using ORNL CIRFT, with one under static and nine dynamic loading conditions. Under static unidirectional loading, a moment of 85 N·m was obtained at maximum curvature 4.0 m⁻¹. The specimen did not show any sign of failure in three repeated loading cycles to almost same maximum curvature. Ten cyclic tests were conducted with amplitude varying from 15.2 to 7.1 N·m. Failure was observed in nine of the tested rod specimens. The cycles to failure were

  5. Refabricated and instrumented fuel rods

    International Nuclear Information System (INIS)

    Silberstein, K.

    2005-01-01

    Nuclear Fuel for power reactors capabilities evaluation is strongly based on the intimate knowledge of its behaviour under irradiation. This knowledge can be acquired from refabricated and instrumented fuel rods irradiated at different levels in commercial reactors. This paper presents the development and qualification of a new technique called RECTO related to a double-instrumented rod re-fabrication process developed by CEA/LECA hot laboratory facility at CADARACHE. The technique development includes manufacturing of the properly dimensioned cavity in the fuel pellet stack to house the thermocouple and the use of a newly designed pressure transducer. An analytic irradiation of such a double-instrumented fuel rod will be performed in OSIRIS test reactor starting October 2004. (Author)

  6. Method of inserting fuel rod

    International Nuclear Information System (INIS)

    Kamimoto, Shuji; Imoo, Makoto; Tsuchida, Kenji.

    1991-01-01

    The present invention concerns a method of inserting a fuel rod upon automatic assembling, automatic dismantling and reassembling of a fuel assembly in a light water moderated reactor, as well as a device and components used therefor. That is, a fuel rod is inserted reliably to an aimed point of insertion by surrounding the periphery of the fuel rod to be inserted with guide rods, and thereby suppressing the movement of the fuel rod during insertion. Alternatively, a fuel rod is inserted reliably to a point of insertion by inserting guide rods at the periphery of the point of insertion for the fuel rod to be inserted thereby surrounding the point of insertion with the guide rods or fuel rods. By utilizing fuel rods already present in the fuel assembly as the guide rods described above, the fuel rod can be inserted reliably to the point of insertion with no additional devices. Dummy fuel rods are previously inserted in a fuel assembly which are then utilized as the above-mentioned guide rods to accurately insert the fuel rod to the point of insertion. (I.S.)

  7. Detection of the departure from nucleate boiling (DNB) in nuclear fuel rod simulators

    Energy Technology Data Exchange (ETDEWEB)

    Mesquita, Amir Z.; Rezende, Hugo C.; Santos, Andre Augusto C.; Silva, Vitor Vasconcelos A.; Campolina, Daniel de Almeida M., E-mail: amir@cdtn.br, E-mail: hcr@cdtn.br, E-mail: aacs@cdtn.br, E-mail: vitors@cdtn.br, E-mail: campolina@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/UFMG), Belo Horizonte, MG (Brazil). Servico de Tecnologia de Reatores; Palma, Daniel Artur P., E-mail: dapalma@cnen.gov.br [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)

    2013-07-01

    In the thermal hydraulic experiments to determinate parameters of heat transfer, where fuel rod simulators are heated by electric current, the preservation of the simulators are essential when the heat flux goes to the critical point. One of the most important limits in the design of cooling water reactors is the condition in which the heat transfer coefficient by boiling in the core deteriorates itself. The departure from nucleate boiling (DNB) happens in the area of low steam quality when there is nucleus formation of bubbles. This result in a departure from nucleate boiling in which steam bubbles no longer break away from the solid surface of the channel, bubbles dominate the channel or surface, and the heat flux dramatically decreases. Vapor essentially insulates the bulk liquid from the hot surface. At this time, the small increase in the heat flux or in the inlet temperature of the cooler in the core, or the small decrease in the inlet flux of cooling, results in changes in the heat transfer mechanism. This causes increases in the surface temperature of the fuel elements causing failures at the fuel (burnout). This paper describes the experiments conducted to detection of critical heat flux in nuclear fuel element simulators carried out in the thermal-hydraulic laboratory of Nuclear Technology Development Centre (CDTN). It is concluded that the use of displacement transducer is the most efficient technique for detecting of critical heat flux in nuclear simulators heated by electric current in open pool. (author)

  8. Detection of the departure from nucleate boiling (DNB) in nuclear fuel rod simulators

    International Nuclear Information System (INIS)

    Mesquita, Amir Z.; Rezende, Hugo C.; Santos, Andre Augusto C.; Silva, Vitor Vasconcelos A.; Campolina, Daniel de Almeida M.

    2013-01-01

    In the thermal hydraulic experiments to determinate parameters of heat transfer, where fuel rod simulators are heated by electric current, the preservation of the simulators are essential when the heat flux goes to the critical point. One of the most important limits in the design of cooling water reactors is the condition in which the heat transfer coefficient by boiling in the core deteriorates itself. The departure from nucleate boiling (DNB) happens in the area of low steam quality when there is nucleus formation of bubbles. This result in a departure from nucleate boiling in which steam bubbles no longer break away from the solid surface of the channel, bubbles dominate the channel or surface, and the heat flux dramatically decreases. Vapor essentially insulates the bulk liquid from the hot surface. At this time, the small increase in the heat flux or in the inlet temperature of the cooler in the core, or the small decrease in the inlet flux of cooling, results in changes in the heat transfer mechanism. This causes increases in the surface temperature of the fuel elements causing failures at the fuel (burnout). This paper describes the experiments conducted to detection of critical heat flux in nuclear fuel element simulators carried out in the thermal-hydraulic laboratory of Nuclear Technology Development Centre (CDTN). It is concluded that the use of displacement transducer is the most efficient technique for detecting of critical heat flux in nuclear simulators heated by electric current in open pool. (author)

  9. Thermal performance of the nuclear fuel rods submitted to angular variation of the heat exchanger coefficients

    International Nuclear Information System (INIS)

    Carvalho, A.M.M. de.

    1984-01-01

    Generally, LMFBR fuel rods consist of fuel pellets encapsulated in cladding tubes. These tubes are wrapped by a helical wire, working as a spacer. Distortions in the rod temperature distribution and in the external heat flux can be generated by angular variations in the local heat transfer coefficients due to the wire, by excentricity between pellet and clad or by ovalization of the cladding tube. Also, the temperature distributions can be affected by fuel densification, reestructuring and swelling. The present work consists of the development of a computer code in order to analyse the fuel rod performance as function of geometrical and operational effects, in steady state regime. (Author) [pt

  10. Proceedings of the specialist meeting on nuclear fuel and control rods: operating experience, design evolution and safety aspects

    International Nuclear Information System (INIS)

    1997-01-01

    Design and management of nuclear fuel has undergone a strong evolution process during past years. The increase of the operating cycle length and of the discharge burnup has led to the use of more advanced fuel designs, as well as to the adoption of fuel efficient operational strategies. The analysis of recent operational experience highlighted a number of issues related to nuclear fuel and control rod events raising concerns about the safety aspects of these new designs and operational strategies, which led to the organisation of this Specialists Meeting on fuel and control rod issues. The meeting was intended to provide a forum for the exchange of information on lessons learned and safety concern related to operating experience with fuel and control rods (degradation, reliability, experience with high burnup fuel, and others). After an opening session 6 papers), this meeting was subdivided into four sessions: Operating experience and safety concern (technical session I - 6 papers), Fuel performance and operational issues (technical session II - 7 papers), Control rod issues (technical session III - 9 papers), Improvement of fuel design (technical session IV.A - 4 papers), Improvement on fuel fabrication and core management (technical session IV.B - 6 papers)

  11. Process and equipment for pressure build-up in nuclear reactor fuel rods

    International Nuclear Information System (INIS)

    Heer, W.F.; Carli, E.V. de.

    1976-01-01

    The equipment makes possible the build-up of inert gas pressure in a filled and closed fuel can, i.e. in a complete fuel rod. Handling is simple, it is suitable for mass production and only causes low processing costs. The quality, e.g. the degree of purity of the contents of the rod, remains unchangedin processing. The equipment consists of a vacuum-tight space, into which the equally vacuum tight fuel rod is introduced, and can be fixed so that its position can be reproduced unmistakeably. The vacuum space contains a connection for the inert gases and a laser arrangement. After inserting a fuel rod into the facility, this is evacuated and the fuel can has a hole bored in it by a laser beam. After fast equalisation of pressure, an inert gas at the required pressure is introduced into the chamber and the fuel rod. After the filling process is completed, the fuel can is closed again with the same laser beam. The quality of the seal obtained, i.e the leak-tightness of the fuel can, can be checked after reduction of the inert gas pressure and before taking out the fuel rod, by repeated evacuation of the chamber. Laser light energies between 13,000 and 110,000 Joule/sq cm are sufficient. Optimum results were obtained for a Zircaloy fuel can with about 52,000 Joule/sq cm. (TK) [de

  12. Drying of encapsulated parts (nuclear fuel rods) in applying vacuum, by introducing dehydratings, vacuum, and filling with an inert gas

    International Nuclear Information System (INIS)

    Johnson, C.R.

    1976-01-01

    This invention concerns a decontamination technique, in particular a process and equipment for extracting the water contained in fuel rods and other similar components of a nuclear reactor. The extraction of the contaminants contained in the fuel rods is carried out by a standard method by drilling a small hole in the surface of the cladding and applying a vacuum to bleed the rod of its impurities (moisture and gas). The invention consists for example in applying a vacuum at the hole drilled in the cladding to extract the contaminants and introducing spirit into the rod through the same orifice. The spirit absorbs the remaining liquid and other impurities. The spirit charged with the impurities is then pumped out by the same aperture by means of a regulated atmosphere inside a closed receptacle. This receptacle is then filled with an inert gas cooled to ambient temperature. The rods are then pressurised and the small orifice is sealed [fr

  13. Contribution to numerical and mechanical modelling of pellet-cladding interaction in nuclear reactor fuel rod

    International Nuclear Information System (INIS)

    Retel, V.

    2002-12-01

    Pressurised water reactor fuel rods (PWR) are the place of nuclear fission, resulting in unstable and radioactive elements. Today, the mechanical loading on the cladding is harder and harder and is partly due to the fuel pellet movement. Then, the mechanical behaviour of the cladding needs to be simulated with models allowing to assess realistic stress and strain fields for all the running conditions. Besides, the mechanical treatment of the fuel pellet needs to be improved. The study is part of a global way of improving the treatment of pellet-cladding interaction (PCI) in the 1D finite elements EDF code named CYRANO3. Non-axisymmetrical multidirectional effects have to be accounted for in a context of unidirectional axisymmetrical finite elements. The aim of this work is double. Firstly a model simulating the effect of stress concentration on the cladding, due to the opening of the radial cracks of fuel, had been added in the code. Then, the fragmented state of fuel material has been taken into account in the thermomechanical calculation, through a model which led the strain and stress relaxation in the pellet due to the fragmentation, be simulated. This model has been implemented in the code for two types of fuel behaviour: elastic and viscoplastic. (author)

  14. Vibrational characteristics and wear of fuel rods

    International Nuclear Information System (INIS)

    Schmugar, K.L.

    1977-01-01

    Fuel rod wear, due to vibration, is a continuing concern in the design of liquid-cooled reactors. In my report, the methodology and models that are used to predict fuel rod vibrational response and vibratory wear, in a light water reactor environment, are discussed. This methodology is being followed at present in the design of Westinghouse Nuclear Fuel. Fuel rod vibrations are expressed as the normal bending modes, and sources of rod vibration are examined with special emphasis on flow-induced mechanisms in the stable flow region. In a typical Westinghouse PWR fuel assembly design, each fuel rod is supported at multiple locations along the rod axis by a square-shaped 'grid cell'. For a fuel rod /grid support system, the development of small oscillatory motions, due to fluid flow at the rod/grid interface, results in material wear. A theoretical wear mode is developed using the Archard Theory of Adhesive Wear as the basis. Without question certainty, fretting wear becomes a serious problem if it progresses to the stage where the fuel cladding is penetrated and fuel is exposed to the coolant. Westinghouse fuel is designed to minimize fretting wear by limiting the relative motion between the fuel rod and its supports. The wear producing motion between the fuel rod and its supports occurs when the vibration amplitude exceeds the slippage threshold amplitude

  15. CFD analysis of multiphase coolant flow through fuel rod bundles in advanced pressure tube nuclear reactors

    International Nuclear Information System (INIS)

    Catana, A.; Turcu, I.; Prisecaru, I.; Dupleac, D.; Danila, N.

    2010-01-01

    The key component of a pressure tube nuclear reactor core is pressure tube filled with a stream of fuel bundles. This feature makes them suitable for CFD thermal-hydraulic analysis. A methodology for CFD analysis applied to pressure tube nuclear reactors is presented in this paper, which is focused on advanced pressure tube nuclear reactors. The complex flow conditions inside pressure tube are analysed by using the Eulerian multiphase model implemented in FLUENT CFD computer code. Fuel rods in these channels are superheated but the liquid is under high pressure, so it is sub-cooled in normal operating conditions on most of pressure tube length. In the second half of pressure tube length, the onset of boiling occurs, so the flow consists of a gas liquid mixture, with the volume of gas increasing along the length of the channel in the direction of the flow. Limited computer resources enforced us to use CFD analysis for segments of pressure tube. Significant local geometries (junctions, spacers) were simulated. Main results of this work are: prediction of main thermal-hydraulic parameters along pressure tube including CHF evaluation through fuel assemblies. (authors)

  16. Nuclear fuels

    International Nuclear Information System (INIS)

    Gangwani, Saloni; Chakrabortty, Sumita

    2011-01-01

    Nuclear fuel is a material that can be consumed to derive nuclear energy, by analogy to chemical fuel that is burned for energy. Nuclear fuels are the most dense sources of energy available. Nuclear fuel in a nuclear fuel cycle can refer to the fuel itself, or to physical objects (for example bundles composed of fuel rods) composed of the fuel material, mixed with structural, neutron moderating, or neutron reflecting materials. Long-lived radioactive waste from the back end of the fuel cycle is especially relevant when designing a complete waste management plan for SNF. When looking at long-term radioactive decay, the actinides in the SNF have a significant influence due to their characteristically long half-lives. Depending on what a nuclear reactor is fueled with, the actinide composition in the SNF will be different. The following paper will also include the uses. advancements, advantages, disadvantages, various processes and behavior of nuclear fuels

  17. The State of the Art of the Decladding Method for the Spent Nuclear Fuel Rods

    International Nuclear Information System (INIS)

    Jung, Jae Hoo; Park, B. S.; Kim, K. H.; Hong, D. H.; Yoon, J. S.; Lee, H. J.; Kim, S. H.; Song, T. G.; Lim, K. M.; Lee, J. K.

    2006-12-01

    Our country's energy consumption is increasing day after day even though it relies on imports for more than 95 percent of its energy needs. In this circumstances, the atomic energy is a promising alternative to solve the problem of an energy security and an environmental preservation simultaneously. However, nuclear power produces spent fuel which is a highly radioactive waste. For a reliable and effective management of the spent fuel, the ACP(Advanced Spent Fuel Conditioning Process) is being developed at the KAERI. As a first state equipment of the ACP, a decladding machine is used to separate spent fuel rod into the UO 2 pellets and hulls. This technical report aims to analyze existing decladding methods, and then, find a suitable decladding mechanism for the ACP. Many studies on the decladding of spent fuel can be categorized two approaches: chemical approach and mechanical one. In this report, we concentrated on the mechanical decladding approach. We developed engineering scale decladding device(20 kgHM/batch) and evaluated the performance through the verification experiments. We expect that this technical report helps in developing a scale-up equipment and technology

  18. Process Management Development for Quality Monitoring on Resistance Weldment of Nuclear Fuel Rods

    Energy Technology Data Exchange (ETDEWEB)

    Na, Tae Hyung; Yang, Kyung Hwan; Kim, In Kyu [KEPCO, Daejeon (Korea, Republic of)

    2016-05-15

    The current, welding force, and displacement are displayed on the indicator during welding. However, real-time quality control is not performed. Due to the importance of fuel rod weldment, many studies on welding procedures have been conducted. However, there are not enough studies regarding weldment quality evaluation. On the other hand, there are continuous studies on the monitoring and control of welding phenomena. In resistance welding, which is performed in a very short time, it is important to find the process parameters that well represent the weld zone formation and the welding process. In his study, Gould attempted to analyze melt zone formation using the finite difference method. Using the artificial neural network, Javed and Sanders, Messler Jr et al., Cho and Rhee, Li and Gong et al. estimated the size of the melt zone by mapping a nonlinear functional relation between the weldment and the electrode head movement, which is a typical welding process parameter. Applications of the artificial intelligence method include fuzzy control using electrode displacement, fuzzy control using the optimal power curve, neural network control using the dynamic resistance curve, fuzzy adaptive control using the optimal electrode curve, etc. Therefore, this study induced quality factors for the real-time quality control of nuclear fuel rod end plug weldment using instantaneous dynamic resistance (IDR), which incorporates the instantaneous value of secondary current and voltage of the transformer, and using instantaneous dynamic force (IDF), obtained real-time during welding.

  19. Nodal methods for calculating nuclear reactor transients, control rod patterns, and fuel pin powers

    International Nuclear Information System (INIS)

    Cho, Byungoh.

    1990-01-01

    Nodal methods which are used to calculate reactor transients, control rod patterns, and fuel pin powers are investigated. The 3-D nodal code, STORM, has been modified to perform these calculations. Several numerical examples lead to the following conclusions: (1) By employing a thermal leakage-to-absorption ratio (TLAR) approximation for the spatial shape of the thermal fluxes for the 3-D Langenbuch-Maurer-Werner (LMW) and the superprompt critical transient problems, the convergence of the conventional two-group scheme is accelerated. (2) By employing the steepest-ascent hill climbing search with heuristic strategies, Optimum Control Rod Pattern Searcher (OCRPS) is developed for solving control rod positioning problem in BWRs. Using the method of approximation programming the objective function and the nuclear and thermal-hydraulic constraints are modified as heuristic functions that guide the search. The test calculations have demonstrated that, for the first cycle of the Edwin Hatch Unit number-sign 2 reactor, OCRPS shows excellent performance for finding a series of optimum control rod patterns for six burnup steps during the operating cycle. (3) For the modified two-dimensional EPRI-9R problem, the least square second-order polynomial flux expansion method was demonstrated to be computationally about 30 times faster than a fine-mesh finite difference calculation in order to achieve comparable accuracy for pin powers. The basic assumption of this method is that the reconstructed flux can be expressed as a product of an assembly form function and a second-order polynomial function

  20. A numerical solution model of the rewetting of a nuclear fuel rod

    International Nuclear Information System (INIS)

    Braz Filho, F.A.

    1984-01-01

    The study of thermal behaviour of a nuclear reactor fuel rod during the reflooding phase of the loss-of-coolant accident (LOCA) is presented. A mathematical model and a numerical scheme were proposed in order to solve the bidimensional heat conduction equation in cylindrical coordinates. The phenomenon of reflooding is not completely understood. One of the main difficulties is to estimate the heat transfer coefficient (h). For this reason two different models were elaborated: in the first three regions are considered and in each region h is considered constant; in the second the h profile is adjusted according to the boiling curve. The three region model yields satisfactory results at high and low mass flows while the 'boiling curve' model yields reasonable at low flows. (Author) [pt

  1. Thermal performance of a buried nuclear waste storage container storing a hybrid mix of PWR and BWR spent fuel rods

    International Nuclear Information System (INIS)

    Johnson, G.L.

    1988-09-01

    Lawrence Livermore National Laboratory will design, model, and test nuclear waste packages for use at the Nevada Nuclear Waste Storage Repository at Yucca Mountain, Nevada. One such package would store lightly packed spent fuel rods from both pressurized and boiling water reactors. The storage container provides the primary containment of the nuclear waste and the spent fuel rod cladding provides secondary containment. A series of transient conduction and radiation heat transfer analyses was run to determine for the first 1000 yr of storage if the temperature of the tuff at the borehole wall ever falls below 97/degree/C and whether the cladding of the stored spent fuel ever exceeds 350/degree/C. Limiting the borehole to temperatures of 97/degree/C or greater helps minimize corrosion by assuring that no condensed water collects on the container. The 350/degree/C cladding limit minimizes the possibility of creep-related failure in the spent fuel rod cladding. For a series of packages stored in a 8 x 30 m borehole grid where each package contains 10-yr-old spent fuel rods generating 4.74 kW or more, the borehole wall stays above 97/degree/C for the full 1000-yr analysis period

  2. SCORE-EVET: a computer code for the multidimensional transient thermal-hydraulic analysis of nuclear fuel rod arrays

    International Nuclear Information System (INIS)

    Benedetti, R.L.; Lords, L.V.; Kiser, D.M.

    1978-02-01

    The SCORE-EVET code was developed to study multidimensional transient fluid flow in nuclear reactor fuel rod arrays. The conservation equations used were derived by volume averaging the transient compressible three-dimensional local continuum equations in Cartesian coordinates. No assumptions associated with subchannel flow have been incorporated into the derivation of the conservation equations. In addition to the three-dimensional fluid flow equations, the SCORE-EVET code ocntains: (a) a one-dimensional steady state solution scheme to initialize the flow field, (b) steady state and transient fuel rod conduction models, and (c) comprehensive correlation packages to describe fluid-to-fuel rod interfacial energy and momentum exchange. Velocity and pressure boundary conditions can be specified as a function of time and space to model reactor transient conditions such as a hypothesized loss-of-coolant accident (LOCA) or flow blockage

  3. Inspection of domestic nuclear fuel rods using neutron radiography at the Tehran research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Dastjerdi, Mohammad Hosein Choopan; Khalafi, Hossein; Kasesaz, Yaser [Nuclear Science and Technology Research Institute, Tehran (Iran, Islamic Republic of); Movafeghi, Amir

    2016-11-01

    Three unused domestic fuel rods were investigated qualitatively and quantitatively by means of thermal neutron radiography. The neutron radiography tests were performed by the image plate method at Tehran research reactor in order to check the fuel properties. The pellets of these three fuel rods contained three different U-235 enrichments and different sizes that were filled into a zircalloy tube. In the qualitative investigations, the difference in size and enrichment between the pellets and the gaps between them were obviously recognized in the image of the fuel rods. In the quantitative investigations, data of the pellets compositions, their sizes (lengths and diameters) and the gaps between them were extracted from obtained images. It was found that the measured data and the manufacturer's specifications are in good agreement.

  4. Inspection of domestic nuclear fuel rods using neutron radiography at the Tehran research reactor

    International Nuclear Information System (INIS)

    Dastjerdi, Mohammad Hosein Choopan; Khalafi, Hossein; Kasesaz, Yaser; Movafeghi, Amir

    2016-01-01

    Three unused domestic fuel rods were investigated qualitatively and quantitatively by means of thermal neutron radiography. The neutron radiography tests were performed by the image plate method at Tehran research reactor in order to check the fuel properties. The pellets of these three fuel rods contained three different U-235 enrichments and different sizes that were filled into a zircalloy tube. In the qualitative investigations, the difference in size and enrichment between the pellets and the gaps between them were obviously recognized in the image of the fuel rods. In the quantitative investigations, data of the pellets compositions, their sizes (lengths and diameters) and the gaps between them were extracted from obtained images. It was found that the measured data and the manufacturer's specifications are in good agreement.

  5. Determination of the vibration characteristics of nuclear fuel rods in a fluid flow using multiphysics computation

    International Nuclear Information System (INIS)

    Sbragio, Ricardo

    1999-01-01

    The determination of natural frequencies and displacement Power Spectrum Density (PSD) of fuel rods in a fluid using Computational Fluid Dynamics and Finite Element Methods is presented. The rods are modeled as slender beams subjected to small displacements in a fluid using three-dimensional mesh. The incompressible Navier-Stokes and linear momentum balance equations are solved simultaneously using Spectrum code. Two examples from literature are analyzed. The first consists in one rod in a fluid. The excitation is an impulse force at the rod central node. The second example is a two rod system in a fluid. In this case, the excitation force is random and is determined from a PSD. (author)

  6. 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

  7. Nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    Hindle, E.D.

    1984-01-01

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

  8. Process and device for fastening and removing fuel-absorber rods in fuel elements of nuclear reactors

    International Nuclear Information System (INIS)

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

    1980-01-01

    This is concerned with an improvement of the fixing of absorber rods in a nuclear reactor. It is important that the rod should not be damaged during removal from the reactor, and that no particles of material are shed during this process. According to the invention, the rod has a stalk which is pressed into a hole in the star shaped arms and welded in. During removal, the stalk is broken at a preferred position. Details of construction are described. (UWI) [de

  9. Investigating mechanical behavior and radiation resistant of fuel rods clad in nuclear power plant

    International Nuclear Information System (INIS)

    Sedgh Kerdar, A.

    1999-01-01

    interstitials in metal lattice under irradiation causes increased strength and hardness but decreases ductility in metals.The increase in strength and hardness depends on obstacles that prevent the motion of dislocations. The clustering of point defects are responsible for these changes. Irradiation also induces instabilities in phases due to enhancement of diffusion, solute segregation, precipitate formation, order- disorder transformation and resolution of small precipitates. From the microscopic point of view accumulation of vacancies accompanied by formation of He and H 2 gases under irradiation cause an increase in volume which results in swelling and eventually ends up with embrittlement of metals. This subject was described in chapter three Zirconium and its alloys are the best structural materials for fuel cladding of BWR and PWR reactors core. The working condition in the core of nuclear reactor are very serve, respect temperature and radiation dose. It should be realized that, if fuel cladding receive damage and get cracked, the first cooling cycle and the maine equipment will be contaminated with active materials which cause additional environmental problems. Furthermore, replacement of fuel rods are very costly. Therefore, for increasing life time of fuel cladding and minimizing damage, the effect of radiation and heat on Zirconium and its alloys must be investigated. This subject was described in chapter four.The mechanical behavior and radiation resistant of fuel cladding in PWR reactor (specifically WWER ) have been investigated which is described in chapter five. Result, discussion and final conclusion are summarized in last chapter and also several points for improvement have been offered

  10. Utilization of carbon/carbon composites in nuclear simulation fuel rods

    International Nuclear Information System (INIS)

    Polidoro, H.A.; Otani, S.; Rezende, M.C.; Ferreira, S.R.; Otani, C.

    1988-01-01

    Thermo-hydraulic problems, in nuclear plants are normally analysed by using electrically heated rods. Carbon/carbon composites were used to make heating elements for testing by indirect heating up to a heat flux of 100 W/cm 2 . It is easy to verify that this value can be exceed if the choice of the complementary materials for insulator and cladding were improved. The swaging process used to reduce the cladding diameter prevented the fabrication of graphite heater rods. (author) [pt

  11. Method and apparatus for the production of a nuclear fuel rod

    International Nuclear Information System (INIS)

    Ballard, A.S.; Cooper, R.G.; Davis, D.E.

    1975-01-01

    The method designs the manufacture of e.g. rod-shaped fuel element fillings in which fuel particles are suspended within a liquid and solidifiable binder such as graphite powder in pitch. The fuel particles are filled into cavities whose cross-sections correspond to those of the fuel rods. After closing with a covering plate, a piston exerts a force from below on it until its solidification. To follow, the liquid binder is injected through lower openings in the cavities. Due to the lubricity of the binder, the cavities are heated to 150 to 175 0 C, the packing of particles are homogenized. This procedure is further supported by the constant pressure of the pistons. Excess binder and air can flow out through openings in the covering plate. After cooling and solidification of the binder as well as after removal of the covering plate, the piston thrusts out the formed bodies or fuel rods from the cavities by an upwards movement. (DG/LH) [de

  12. Starting Point, Keys and Milestones of a Computer Code for the Simulation of the Behaviour of a Nuclear Fuel Rod

    Directory of Open Access Journals (Sweden)

    Armando C. Marino

    2011-01-01

    Full Text Available The BaCo code (“Barra Combustible” was developed at the Atomic Energy National Commission of Argentina (CNEA for the simulation of nuclear fuel rod behaviour under irradiation conditions. We present in this paper a brief description of the code and the strategy used for the development, improvement, enhancement, and validation of a BaCo during the last 30 years. “Extreme case analysis”, parametric (or sensitivity, probabilistic (or statistic analysis plus the analysis of the fuel performance (full core analysis are the tools developed in the structure of BaCo in order to improve the understanding of the burnup extension in the Atucha I NPP, and the design of advanced fuel elements as CARA and CAREM. The 3D additional tools of BaCo can enhance the understanding of the fuel rod behaviour, the fuel design, and the safety margins. The modular structure of the BaCo code and its detailed coupling of thermo-mechanical and irradiation-induced phenomena make it a powerful tool for the prediction of the influence of material properties on the fuel rod performance and integrity.

  13. Thermal performance of a buried nuclear waste storage container storing a hybrid mix of PWR and BWR spent fuel rods

    International Nuclear Information System (INIS)

    Johnson, G.L.

    1991-11-01

    Lawrence Livermore National Laboratory will design, model, and test nuclear waste packages for use at the Nevada Nuclear Waste Storage Repository at Yucca Mountain, Nevada. On such package would store tightly packed spent fuel rods from both pressurized and boiling water reactors. The storage container provides the primary containment of the nuclear waste and the spent fuel rod cladding provides secondary containment. A series of transient conduction and radiation heat transfer analyses was run to determine for the first 1000 yr of storage if the temperature of the tuff at the borehole wall ever falls below 97 degrees C and whether the cladding of the stored spent fuel ever exceeds 350 degrees C. Limiting the borehole to temperatures of 97 degrees C or greater helps minimize corrosion by assuring that no condensed water collects on the container. The 350 degrees C cladding limit minimizes the possibility of creep- related failure in the spent fuel rod cladding. For a series of packages stored in a 8 x 30 m borehole grid where each package contains 10-yr-old spent fuel rods generating 4.74 kW or more, the borehole wall stays above 97 degrees C for the full 10000-yr analysis period. For the 4.74-kW load, the peak cladding temperature rises to just below the 350 degrees C limit about 4 years after emplacement. If the packages are stored using the spacing specified in the Site Characterization Plan (15 ft x 126 ft), a maximum of 4.1 kW per container may be stored. If the 0.05-m-thick void between the container and the borehole wall is filled with loosely packed bentonite, the peak cladding temperature rises more than 40 degrees C above the allowed cladding limit. In all cases the dominant heat transfer mode between container components is thermal radiation

  14. Automatic system of welding for nuclear fuel rods; Sistema automatico de soldadura para barras de combustible nuclear

    Energy Technology Data Exchange (ETDEWEB)

    Romero G, M; Romero C, J [Instituto Nacional de Investigaciones Nucleares, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

    1998-07-01

    The welding process of nuclear fuel must be realized in an inert gas environment (He) and constant flow of this. In order to reach these conditions it is necessary to do vacuum at the chamber and after it is pressurized with the noble gas (purge) twice in the welding chamber. The purge eliminates impurities that can provoke oxidation in the weld. Once the conditions for initiating the welding are gotten, it is necessary to draw a graph of the flow parameters, pressure, voltage and arc current and to analyse those conditions in which have been carried out the weld. The rod weld must be free of possible pores or cracks which could provoke rod leaks, so reducing the probability of these failures should intervene mechanical and metallurgical factors. Automatizing the process it allows to do reliable welding assuring that conditions have been performed, reaching a high quality welding. Visually it can be observed the welding process by means of a mimic which represents the welding system. There are the parameters acquired such as voltage, current, pressure and flow during the welding arc to be analysed later. (Author)

  15. Apparatus for loading fuel pellets in fuel rods

    International Nuclear Information System (INIS)

    Tedesco, R.J.

    1976-01-01

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

  16. Quivers For Special Fuel Rods-Disposal Of Special Fuel Rods In CASTOR V Casks

    International Nuclear Information System (INIS)

    Bannani, Amin; Cebula, Wojciech; Buchmuller, Olga; Huggenberg, Roland; Helmut Kuhl

    2015-01-01

    While GNS casks of the CASTOR family are a suitable means to transfer fuel assemblies (FA) from the NPP to an interim dry storage site, Germanys phase-out of nuclear energy has triggered the demand for an additional solution to dispose of special fuel rods (SFR), normally remaining in the fuel pond until the final shutdown of the NPP. SFR are fuel rods that had to be removed from fuel assemblies mainly due to their special condition, e. g. damages in the cladding of the fuel rods which may have occurred during reactor operations. SFR are usually stored in the spent fuel pond after they are removed from the FA. The quiver for special fuel rods features a robust yet simple design, with a high mechanical stability, a reliable leak-tightness and large safety margins for future requirements on safety analysis. The quiver for special fuel rods can be easily adapted to a large variety of different damaged fuel rods and tailored to the specific need of the customer. The quiver for special fuel rods is adaptable e.g. in length and diameter for use in other types of transport and storage casks and is applicable in other countries as well. The overall concept presented here is a first of its kind solution for the disposal of SFRs via Castor V-casks. This provides an important precondition in achieving the status 'free from nuclear fuel' of the shut down German NPPs

  17. Quivers For Special Fuel Rods-Disposal Of Special Fuel Rods In CASTOR V Casks

    Energy Technology Data Exchange (ETDEWEB)

    Bannani, Amin; Cebula, Wojciech; Buchmuller, Olga; Huggenberg, Roland [GNS, Essen (Germany); Helmut Kuhl [WTI, Julich (Germany)

    2015-05-15

    While GNS casks of the CASTOR family are a suitable means to transfer fuel assemblies (FA) from the NPP to an interim dry storage site, Germanys phase-out of nuclear energy has triggered the demand for an additional solution to dispose of special fuel rods (SFR), normally remaining in the fuel pond until the final shutdown of the NPP. SFR are fuel rods that had to be removed from fuel assemblies mainly due to their special condition, e. g. damages in the cladding of the fuel rods which may have occurred during reactor operations. SFR are usually stored in the spent fuel pond after they are removed from the FA. The quiver for special fuel rods features a robust yet simple design, with a high mechanical stability, a reliable leak-tightness and large safety margins for future requirements on safety analysis. The quiver for special fuel rods can be easily adapted to a large variety of different damaged fuel rods and tailored to the specific need of the customer. The quiver for special fuel rods is adaptable e.g. in length and diameter for use in other types of transport and storage casks and is applicable in other countries as well. The overall concept presented here is a first of its kind solution for the disposal of SFRs via Castor V-casks. This provides an important precondition in achieving the status 'free from nuclear fuel' of the shut down German NPPs.

  18. The effects of off-center pellets on the temperature distribution and the heat flux distribution of fuel rods in nuclear reactors

    International Nuclear Information System (INIS)

    Peng Muzhang; Xing Jianhua

    1986-01-01

    This paper analyzes the effects of off-center pellets on the steady state temperature distribution and heat flux distribution of fuel rods in the nuclear reactors, and derives the dimensionless temperature distribution relationships and the dimensionless heat flux distribution relationship from the fuel rods with off-center pellets. The calculated results show that the effects of off-center will result in not only deviations of the highest temperature placement in the fuel pellets, but also the circumferentially nonuniform distributions of the temperatures and heat fluxes of the fuel rod surfaces

  19. The high temperature out-of-pile test of LVDT for internal pressure measurement of nuclear fuel rod

    Energy Technology Data Exchange (ETDEWEB)

    Son, J. M.; Kim, B. K.; Kim, D. S.; Yoon, K. B.; Sin, Y. T.; Park, S. J.; Kang, Y. H. [KAERI, Taejon (Korea, Republic of)

    2002-10-01

    As a part of the development of instrumentation technologies for the nuclear fuel irradiation test in HANARO(High-flux Advanced Nuclear Application Reactor), the internal pressure measurement technique of the nuclear fuel rod is being developed using LVDT(Linear Variable Differential Transformer). As the results of out-of-pile test at room temperature, it was concluded that the well qualified out-of-pile tests were needed to understand the LVDT's detail characteristics at high temperature for the detail design of the fuel irradiation capsule, because LVDT is very sensitive to variation of temperature. Therefore, the high temperature out-of-pile test system for pressure measurement was developed, and this test was performed under the temperature condition between room temperature and 300 .deg. C increasing the pressure from 0 bar to 30 bar. The LVDT's high temperature characteristics and temperature sensitivity of LVDT were analyzed through this experiment. Based on the result of this test, the method for the application of LVDT at high temperature was introduced. It is known that the results will be used to predict accurately the internal pressure of fuel rod during irradiation test.

  20. Inspecting method for fuel rods

    International Nuclear Information System (INIS)

    Watanabe, Masaaki; Kogure, Sumio.

    1976-01-01

    Purpose: To precisely detect the response of flaw in clad tube and submerged fuel pellets from a relationship between the surface of fuel rod and internal signal. Constitution: Ultrasonic reflected waves from the surface of fuel rods and the interior are detected and either one of fuel rod or ultrasonic flaw detecting contact is rotated to thereby precisely detect the response of the flaw of clad tube and submerged fuel pellets from a relationship between said surface and the interior. It will be noted that the ultrasonic flaw detecting contact used is of the line-focus type, the incident angle of ultrasonic wave from the ultrasonic flaw detecting contact relative to the fuel rod is the angle of skew, that is, the ultrasonic flaw detecting contact is not perpendicular to a center axis of the fuel rod but is slightly displace. That is, the use of the aforesaid contact may facilitate discrimination between the surface flaw of the fuel rod and the response of submergence, and in addition, the employment of the aforesaid incident angle makes it hard to receive reflected waves from the surface of the fuel rod which is great in terms of energy to facilitate discrimination of waves responsive to submergence. (Kawakami, Y.)

  1. Prototypical spent nuclear fuel rod consolidation equipment: Phase 2, Final design report: Volume 1, Detailed design

    International Nuclear Information System (INIS)

    Blissell, W.H.; Ciez, A.P.; Goedicke, F.E.; Bessko, C.

    1987-01-01

    This document describes the Westinghouse Final Design for the Prototypical Spent Fuel Consolidation Equipment Demonstration Project. This design represents a fully qualified, licensable, cost effective spent fuel rod consolidation system. As a result of significant concerns raised by DOE and its Technical Review Committee during the 30% Design Review, significant changes were made to the original Preliminary Design resulting from Phase I activities. These changes focused on increased automation, end fitting removal, the rod pulling process and the need to maintain the consolidation canisters as clean as possible. As a result of these changes, the new system is greatly enhanced with a much greater probability of meeting or exceeding the project functional requirements. As a result of delays in resolving cost and contractual differences, additional bench testing was not conducted during Phase II. It is however our belief that the current design exceeds the 90% confidence level required by DOE because of the confidence gained from the Phase I tests, the additional engineering detail completed and the fact that our rod pulling tool has been demonstrated in a similar application at Oconee while our ID tube cutter is a modified (mounting method only) off-the-shelf design. 7 refs., 49 figs., 36 tabs

  2. Optimum nuclear design of target fuel rod for Mo-99 production in HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Myung Hyun [Kyung Hee University, Seoul (Korea)

    1998-04-01

    Nuclear target design for Mo-99 production in HANARO was performed, KAERI proposed target design was analyzed and its feasibility was shown. Three commercial target designs of Cintichem, ANL and KAERI were tested for the HANARO irradiation an d they all satisfied with design specification. A parametric study was done for target design options and Mo-99 yields ratio and surface heat flux were compared. Tested parameters were target fuel thickness, irradiation location, target axial length, packing density of powder fuel, size of target radius, target geometry, fuel enrichment, fuel composition, and cladding material. Optimized target fuel was designed for both LEU and HEU options. (author). 17 refs., 33 figs., 42 tabs.

  3. Design of a nuclear fuel rod support grid using axiomatic design

    International Nuclear Information System (INIS)

    Song, Kee Nam; Yoon, Kyung Ho; Kang, Byung Soo; Park, Gyung Jin; Choi, Sung Kyoo

    2002-01-01

    Recently, much attention is imposed on the design of the fuel assemblies in the Pressurized Light Water Reactor (PWR). Spacer grid is one of the main structural components in a fuel assembly. It supports fuel rods, guides cooling water, and maintains a coolable geometry from the external impact loads. In this research, a new shape of the spacer grid is designed by the axiomatic approach. The Independence axiom is utilized for the design. For conceptual design, functional requirements (FRs) are defined and corresponding design parameters (DPs) are found to satisfy FRs in sequence. Overall configuration and shapes are determined in this process. Detail design is carried out based on the result of the axiomatic design. For the detail design, the system performances are evaluated by using linear and nonlinear finite element analysis. The dimensions are determined by optimization. Some commercial codes are utilized for the analysis and design

  4. DIONISIO 2.0: New version of the code for simulating a whole nuclear fuel rod under extended irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Soba, Alejandro, E-mail: soba@cnea.gov.ar; Denis, Alicia

    2015-10-15

    Highlights: • A new version of the DIONISIO code is developed. • DIONISIO is devoted to simulating the behavior of a nuclear fuel rod in operation. • The formerly two-dimensional simulation of a pellet-cladding segment is now extended to the whole rod length. • An acceptable and more realistic agreement with experimental data is obtained. • The prediction range of our code is extended up to average burnup of 60 MWd/kgU. - Abstract: The version 2.0 of the DIONISIO code, that incorporates diverse new aspects, has been recently developed. One of them is referred to the code architecture that allows taking into account the axial variation of the conditions external to the rod. With this purpose, the rod is divided into a number of axial segments. In each one the program considers the system formed by a pellet and the corresponding cladding portion and solves the numerous phenomena that take place under the local conditions of linear power and coolant temperature, which are given as input parameters. To do this a bi-dimensional domain in the r–z plane is considered where cylindrical symmetry and also symmetry with respect to the pellet mid-plane are assumed. The results obtained for this representative system are assumed valid for the complete segment. The program thus produces in each rod section the values of the temperature, stress, strain, among others as outputs, as functions of the local coordinates r and z. Then, the general rod parameters (internal rod pressure, amount of fission gas released, pellet stack elongation, etc.) are evaluated. Moreover, new calculation tools designed to extend the application range of the code to high burnup, which were reported elsewhere, have also been incorporated to DIONISIO 2.0 in recent times. With these improvements, the code results are compared with some 33 experiments compiled in the IFPE data base, that cover more than 380 fuel rods irradiated up to average burnup levels of 40–60 MWd/kgU. The results of these

  5. DIONISIO 2.0: New version of the code for simulating a whole nuclear fuel rod under extended irradiation

    International Nuclear Information System (INIS)

    Soba, Alejandro; Denis, Alicia

    2015-01-01

    Highlights: • A new version of the DIONISIO code is developed. • DIONISIO is devoted to simulating the behavior of a nuclear fuel rod in operation. • The formerly two-dimensional simulation of a pellet-cladding segment is now extended to the whole rod length. • An acceptable and more realistic agreement with experimental data is obtained. • The prediction range of our code is extended up to average burnup of 60 MWd/kgU. - Abstract: The version 2.0 of the DIONISIO code, that incorporates diverse new aspects, has been recently developed. One of them is referred to the code architecture that allows taking into account the axial variation of the conditions external to the rod. With this purpose, the rod is divided into a number of axial segments. In each one the program considers the system formed by a pellet and the corresponding cladding portion and solves the numerous phenomena that take place under the local conditions of linear power and coolant temperature, which are given as input parameters. To do this a bi-dimensional domain in the r–z plane is considered where cylindrical symmetry and also symmetry with respect to the pellet mid-plane are assumed. The results obtained for this representative system are assumed valid for the complete segment. The program thus produces in each rod section the values of the temperature, stress, strain, among others as outputs, as functions of the local coordinates r and z. Then, the general rod parameters (internal rod pressure, amount of fission gas released, pellet stack elongation, etc.) are evaluated. Moreover, new calculation tools designed to extend the application range of the code to high burnup, which were reported elsewhere, have also been incorporated to DIONISIO 2.0 in recent times. With these improvements, the code results are compared with some 33 experiments compiled in the IFPE data base, that cover more than 380 fuel rods irradiated up to average burnup levels of 40–60 MWd/kgU. The results of these

  6. Fabrication of preliminary fuel rods for SFR

    International Nuclear Information System (INIS)

    Kim, Sun Ki; Oh, Seok Jin; Ko, Young Mo; Woo, Youn Myung; Kim, Ki Hwan

    2012-01-01

    Metal fuels was selected for fueling many of the first reactors in the US, including the Experimental Breeder Reactor-I (EBR-I) and the Experimental Breeder Reactor-II (EBR-II) in Idaho, the FERMI-I reactor, and the Dounreay Fast Reactor (DFR) in the UK. Metallic U.Pu.Zr alloys were the reference fuel for the US Integral Fast Reactor (IFR) program. Metallic fuel has advantages such as simple fabrication procedures, good neutron economy, high thermal conductivity, excellent compatibility with a Na coolant and inherent passive safety. U-Zr-Pu alloy fuels have been used for SFR (sodium-cooled fast reactor) related to the closed fuel cycle for managing minor actinides and reducing a high radioactivity levels since the 1980s. Fabrication technology of metallic fuel for SFR has been in development in Korea as a national nuclear R and D program since 2007. For the final goal of SFR fuel rod fabrication with good performance, recently, three preliminary fuel rods were fabricated. In this paper, the preliminary fuel rods were fabricated, and then the inspection for QC(quality control) of the fuel rods was performed

  7. Temperature sensitivity study of eddy current and digital gauge probes for nuclear fuel rod oxide measurement

    Science.gov (United States)

    Beck, Faith R.; Lind, R. Paul; Smith, James A.

    2018-04-01

    Novel fuels are part of the nationwide effort to reduce the enrichment of Uranium for energy production. Performance of such fuels is determined by irradiating their surfaces. To test irradiated samples, the instrumentation must operate remotely. The plate checker used in this experiment at Idaho National Lab (INL) performs non-destructive testing on fuel rod and plate geometries with two different types of sensors: eddy current and digital thickness gauges. The sensors measure oxide growth and total sample thickness on research fuels, respectively. Sensor measurement accuracy is crucial because even 10 microns of error is significant when determining the viability of an experimental fuel. One parameter known to affect the eddy current and thickness gauge sensors is temperature. Since both sensor accuracies depend on the ambient temperature of the system, the plate checker has been characterized for these sensitivities. The manufacturer of the digital gauge probes has noted a rather large coefficient of thermal expansion for their linear scale. It should also be noted that the accuracy of the digital gauge probes are specified at 20°C, which is approximately 7°C cooler than the average hot-cell temperature. In this work, the effect of temperature on the eddy current and digital gauge probes is studied, and thickness measurements are given as empirical functions of temperature.

  8. Manipulator for fuel elements and control rods in a nuclear reactor

    International Nuclear Information System (INIS)

    Voss, S.H.K.; Kipp, T.

    1974-01-01

    The manipulator serves for the transport and shuffling of fuel elements or control rods. It has a two-piece grab telescope which can be vertically rotated and on which an interior telescopic tube can be moved within a telescopic jacket if the grab head is raised or lowered. For unlimited rotation of the grab telescope a lifting traverse with supporting rocker and control rocker is used, on which the grab head is mounted by means of a hook suspension. On the grab head, two double pawls are arranged which are operated together and which open and close each other. The pawl device is operated by the control rocker. If the grab head is lowered, the double pawls, swinging outwards, with the aid of carrier bolts lock the rotating pawls of a guide matrix with the interior telescopic tube. The guide matrix has slots and bore holes for guiding the elements as well as centering bars for the heads of the fuel elements or control rods to be gripped. In the lowest position, it rests on the centering collar of the telescopic tube. (DG) [de

  9. Manipulator for fuel elements and control rods in a nuclear reactor

    International Nuclear Information System (INIS)

    Voss, S.H.K.; Kipp, T.

    1977-01-01

    The manipulator serves for the transport and shuffling of fuel elements or control rods. It has a two-piece grab telescope which can be vertically rotated and on which an interior telescopic tube can be moved within a telescopic jacket if the grab head is raised or lowered. For unlimited rotation of the grab telescope a lifting traverse with supporting rocker and control rocker is used, on which the grab head is mounted by means of a hook suspension. On the grab head, two double pawls are arranged which are operated together and which open and close each other. The pawl device is operated by the control rocker. If the grab head is lowered, the double pawls, swinging outward with the aid of carrier bolts lock the rotating pawls of a guide matrix with the interior telescopic tube. The guide matrix has slots and bore holes for guiding the elements as well as centering bars for the heads of the fuel elements or control rods to be gripped. In the lowest position, it rests on the centering collar of the telescopic tube. (DG) [de

  10. Development of nuclear fuel rod inspection technique using ultrasonic resonance phenomenon

    International Nuclear Information System (INIS)

    Choi, Myoung Seon; Joo, Young Sang; Jung, Hyun Kyu; Cheong, Yong Moo.

    1997-02-01

    The scattering of plane acoustic waves normally incident on a multilayered cylindrical shell has been formulated using the global matrix approach. And a simple way to formulate the non-resonant background component in the field scattered by an empty elastic shell has been found. This is to replace the surface admittance for the shell with the zero-frequency limit of the surface admittance for the analogous fluid shell (i.e., the shear wave speed in the elastic shell is set to zero). It has been shown that the background thus obtained is exact and applicable to shells of arbitrary thickness and material makeup, and over all frequencies and mode numbers. This way has been also applied to obtain the expressions of the backgrounds for multilayered shells. The resonant ultrasound spectroscopy system has been constructed to measure the resonance spectrum of a single fuel rod. The leak-defective fuel rod detection system of a laboratory scale has been also constructed. Particularly, all techniques and processes necessary for manufacturing the ultrasonic probe of thin (1.2 mm) strip type have been developed. (author). 38 refs., 34 figs

  11. Application of adjoint sensitivity analysis to nuclear reactor fuel rod performance

    International Nuclear Information System (INIS)

    Wilderman, S.J.; Was, G.S.

    1984-01-01

    Adjoint sensitivity analysis in nuclear fuel behavior modeling is extended to operate on the entire power history for both Zircaloy and stainless steel cladding via the computer codes FCODE-ALPHA/SS and SCODE/SS. The sensitivities of key variables to input parameters are found to be highly non-intuitive and strongly dependent on the fuel-clad gap status and the history of the fuel during the cycle. The sensitivities of five key variables, clad circumferential stress and strain, fission gas release, fuel centerline temperature and fuel-clad gap, to eleven input parameters are studied. The most important input parameters (yielding significances between 1 and 100) are fabricated clad inner and outer radii and fuel radius. The least important significances (less than 0.01) are the time since reactor start-up and fuel-burnup densification rate. Intermediate to these are fabricated fuel porosity, linear heat generation rate, the power history scale factor, clad outer temperature, fill gas pressure and coolant pressure. Stainless steel and Zircaloy have similar sensitivities at start-up but these diverges a burnup proceeds due to the effect of the higher creep rate of Zircaloy which causes the system to be more responsive to changes in input parameters. The value of adjoint sensitivity analysis lies in its capability of uncovering dependencies of fuel variables on input parameters that cannot be determined by a sequential thought process. (orig.)

  12. Multiple fuel rod gripper

    International Nuclear Information System (INIS)

    Shields, E.P.

    1987-01-01

    An apparatus is described for gripping an array of rods comprising: (a) gripping members grippingly engageable with the rods, each of which has a hollow portion terminating in an open end for receiving the end of one of the rods; (b) a closing means for causing the hollow portion of each of the gripping members to apply substantially the same gripping force onto the end of its respective rod, including (i) a locking plate having a plurality of tapered holes registrable with the array of rods, wherein the exterior of each of the gripping members is tapered and nested within one of the tapered holes, (ii) a withdrawing means having a hydraulic plunger operatively connected to each of the gripping members for applying a substantially identical withdrawing force on each of the gripping members, whereby the hollow portion of each of the gripping members applies substantially the same gripping force on its respective rod, and (c) means for detecting whether each of the gripping members has grippingly engaged its respective rod

  13. 37-Active rods fuel element for Atucha 1 nuclear power plant. Effects of this change in design over the neutronic behavior, decay power and radioactive inventory

    International Nuclear Information System (INIS)

    Villar, Javier E.

    1999-01-01

    The influence of the use of 37-rods fuel element on the behavior of the Atucha 1 nuclear power plant homogeneous core with slightly enriched fuel to 0.85 w % were studied through representative parameters such as average discharge burnup, channel powers, reactivity coefficients, kinetic parameters, radioactive inventory and decay power. In general, the values of mentioned parameters are similar to those corresponding to a core with the 36-rods fuel element actually in use, although it must be emphasized a decrease both in linear power and, in minor degree, in the efficiency of shut-off and control rods and a slight increase in the discharge burnup. The fuel management strategy developed for a core with 36-rods elements can be maintained. (author)

  14. LOFT advanced fuel rod instrumentation development

    International Nuclear Information System (INIS)

    Billeter, T.R.; Brown, R.L.; Chan, A.I.Y.; Day, C.K.; Meyers, S.C.; Sheen, E.M.; Stringer, J.L.

    1978-01-01

    Advanced fuel rod instrumentation for the Loss of Fluid Test (LOFT) reactor is being developed by the Hanford Engineering Development Laboratory for the Nuclear Regulatory Commission. This effort calls for development of sensors to measure fuel rod axial motion, fuel centerline temperature (to 2200 0 C), fuel rod plenum gas pressure (to 2500 psig), and plenum gas temperature (to 1500 0 F). A parallel test and evaluation of several modified commercial sensors was undertaken and will result in commercial availability of the final qualified sensors. Necessary test facilities were prepared for the development and evaluation effort. Tests to date indicate a three coil Linear Variable Differential Transformer (LVDT), operated from temperature compensating signal source and processing electronics, will meet the desired requirements

  15. Analysis of high burnup fuel behavior under control rod ejection accident in Korea standard nuclear power plant

    International Nuclear Information System (INIS)

    Lee, Chan Bok; Lee, Chung Chan; Kim, Oh Hwan; Kim, Jong Jin

    1996-07-01

    Test results of high burnup fuel behavior under RIA(reactivity insertion accident) indicated that fuel might fail at the fuel enthalpy lower than that in the current fuel failure criteria was derived by the conservative assumptions and analysis of fuel failure mechanisms, and applied to the analysis of control rod ejection accident in the 1,000 MWe Korea standard PWR. Except that three dimensional core analysis was performed instead of conventional zero dimensional analysis, all the other conservative assumptions were kept. Analysis results showed that less than on percent of the fuel rods in the core has failed which was much less than the conventional fuel failure fraction, 9.8 %, even though a newly derived fuel failure criteria -Fuel failure occurs at the power level lower than that in the current fuel failure criteria. - was applied, since transient fuel rod power level was significantly decreased by analyzing the transient fuel rod power level was significantly decreased by analyzing the transient core three dimensionally. Therefore, it can be said that results of the radiological consequence analysis for the control rod ejection accident in the FSAR where fuel failure fraction was assumed 9.8 % is still bounding. 18 tabs., 48 figs., 39 refs. (Author)

  16. Nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    Hindle, E. D.

    1984-01-01

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

  17. Nuclear reactor fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Hindle, E. D.

    1984-10-16

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

  18. Stress Analysis of Fuel Rod under Axial Coolant Flow

    Energy Technology Data Exchange (ETDEWEB)

    Jin, Hai Lan; Lee, Young Shin; Lee, Hyun Seung [Chungnam National University, Daejeon (Korea, Republic of); Park, Num Kyu; Jeon, Kyung Rok [Kerea Nuclear Fuel., Daejeon (Korea, Republic of)

    2010-05-15

    A pressurized water reactor(PWR) fuel assembly, is a typical bundle structure, which uses light water as a coolant in most commercial nuclear power plants. Fuel rods that have a very slender and long clad are supported by fuel assembly which consists of several spacer grids. A coolant is a fluid which flows through device to prevent its overheating, transferring the heat produced by the device to other devices that use or dissipate it. But at the same time, the coolant flow will bring out the fluid induced vibration(FIV) of fuel rods and even damaged the fuel rod. This study has been conducted to investigate the flow characteristics and nuclear reactor fuel rod stress under effect of coolant. Fluid structure interaction(FSI) analysis on nuclear reactor fuel rod was performed. Fluid analysis of the coolant which flow along the axial direction and structural analysis under effect of flow velocity were carried out under different output flow velocity conditions

  19. Stress Analysis of Fuel Rod under Axial Coolant Flow

    International Nuclear Information System (INIS)

    Jin, Hai Lan; Lee, Young Shin; Lee, Hyun Seung; Park, Num Kyu; Jeon, Kyung Rok

    2010-01-01

    A pressurized water reactor(PWR) fuel assembly, is a typical bundle structure, which uses light water as a coolant in most commercial nuclear power plants. Fuel rods that have a very slender and long clad are supported by fuel assembly which consists of several spacer grids. A coolant is a fluid which flows through device to prevent its overheating, transferring the heat produced by the device to other devices that use or dissipate it. But at the same time, the coolant flow will bring out the fluid induced vibration(FIV) of fuel rods and even damaged the fuel rod. This study has been conducted to investigate the flow characteristics and nuclear reactor fuel rod stress under effect of coolant. Fluid structure interaction(FSI) analysis on nuclear reactor fuel rod was performed. Fluid analysis of the coolant which flow along the axial direction and structural analysis under effect of flow velocity were carried out under different output flow velocity conditions

  20. Fabrication of internally instrumented reactor fuel rods

    International Nuclear Information System (INIS)

    Schmutz, J.D.; Meservey, R.H.

    1975-01-01

    Procedures are outlined for fabricating internally instrumented reactor fuel rods while maintaining the original quality assurance level of the rods. Instrumented fuel rods described contain fuel centerline thermocouples, ultrasonic thermometers, and pressure tubes for internal rod gas pressure measurements. Descriptions of the thermocouples and ultrasonic thermometers are also contained

  1. Development for analysis system of rods enrichment of nuclear fuels; Desarrollo de un sistema de analisis de enriquecimiento de barras de combustible nuclear

    Energy Technology Data Exchange (ETDEWEB)

    Rojas C, E.L

    1998-11-01

    Nuclear industry is strongly regulated all over the world and quality assurance is important in every nuclear installation or process related with it. Nuclear fuel manufacture is not the exception. ININ was committed to manufacture four nuclear fuel bundles for the CFE nucleo electric station at Laguna Verde, Veracruz, under General Electric specifications and fulfilling all the requirements of this industry. One of the quality control requisites in nuclear fuel manufacture deals with the enrichment of the pellets inside the fuel bundle rods. To achieve the quality demanded in this aspect, the system described in this work was developed. With this system, developed at ININ it is possible to detect enrichment spikes since 0.4 % in a column of pellets with a 95 % confidence interval and to identify enrichment differences greater than 0.2 % e between homogeneous segments, also with a 95 % confidence interval. ININ delivered the four nuclear fuel bundles to CFE and these were introduced in the core of the nuclear reactor of Unit 1 in the fifth cycle. Nowadays they are producing energy and have shown a correct mechanical performance and neutronic behavior. (Author)

  2. Experience with a fuel rod enrichment scanner

    International Nuclear Information System (INIS)

    Kubik, R.N.; Pettus, W.G.

    1975-01-01

    This enrichment scanner views all fuel rods produced at B and W's Commercial Nuclear Fuel Plant. The scanner design is derived from the PAPAS System reported by R. A. Forster, H. D. Menlove, and their associates at Los Alamos. The spatial resolution of the system and smoothing of the data are discussed in detail. The cost-effectiveness of multi-detector versus single detector scanners of this general design is also discussed

  3. Development of a simplified statistical methodology for nuclear fuel rod internal pressure calculation

    International Nuclear Information System (INIS)

    Kim, Kyu Tae; Kim, Oh Hwan

    1999-01-01

    A simplified statistical methodology is developed in order to both reduce over-conservatism of deterministic methodologies employed for PWR fuel rod internal pressure (RIP) calculation and simplify the complicated calculation procedure of the widely used statistical methodology which employs the response surface method and Monte Carlo simulation. The simplified statistical methodology employs the system moment method with a deterministic statistical methodology employs the system moment method with a deterministic approach in determining the maximum variance of RIP. The maximum RIP variance is determined with the square sum of each maximum value of a mean RIP value times a RIP sensitivity factor for all input variables considered. This approach makes this simplified statistical methodology much more efficient in the routine reload core design analysis since it eliminates the numerous calculations required for the power history-dependent RIP variance determination. This simplified statistical methodology is shown to be more conservative in generating RIP distribution than the widely used statistical methodology. Comparison of the significances of each input variable to RIP indicates that fission gas release model is the most significant input variable. (author). 11 refs., 6 figs., 2 tabs

  4. Nuclear fuels

    International Nuclear Information System (INIS)

    Beauvy, M.; Berthoud, G.; Defranceschi, M.; Ducros, G.; Guerin, Y.; Limoge, Y.; Madic, Ch.; Santarini, G.; Seiler, J.M.; Sollogoub, P.; Vernaz, E.; Guillet, J.L.; Ballagny, A.; Bechade, J.L.; Bonin, B.; Brachet, J.Ch.; Delpech, M.; Dubois, S.; Ferry, C.; Freyss, M.; Gilbon, D.; Grouiller, J.P.; Iracane, D.; Lansiart, S.; Lemoine, P.; Lenain, R.; Marsault, Ph.; Michel, B.; Noirot, J.; Parrat, D.; Pelletier, M.; Perrais, Ch.; Phelip, M.; Pillon, S.; Poinssot, Ch.; Vallory, J.; Valot, C.; Pradel, Ph.; Bonin, B.; Bouquin, B.; Dozol, M.; Lecomte, M.; Vallee, A.; Bazile, F.; Parisot, J.F.; Finot, P.; Roberts, J.F.

    2009-01-01

    Fuel is one of the essential components in a reactor. It is within that fuel that nuclear reactions take place, i.e. fission of heavy atoms, uranium and plutonium. Fuel is at the core of the reactor, but equally at the core of the nuclear system as a whole. Fuel design and properties influence reactor behavior, performance, and safety. Even though it only accounts for a small part of the cost per kilowatt-hour of power provided by current nuclear power plants, good utilization of fuel is a major economic issue. Major advances have yet to be achieved, to ensure longer in-reactor dwell-time, thus enabling fuel to yield more energy; and improve ruggedness. Aside from economics, and safety, such strategic issues as use of plutonium, conservation of resources, and nuclear waste management have to be addressed, and true technological challenges arise. This Monograph surveys current knowledge regarding in-reactor behavior, operating limits, and avenues for R and D. It also provides illustrations of ongoing research work, setting out a few noteworthy results recently achieved. Content: 1 - Introduction; 2 - Water reactor fuel: What are the features of water reactor fuel? 9 (What is the purpose of a nuclear fuel?, Ceramic fuel, Fuel rods, PWR fuel assemblies, BWR fuel assemblies); Fabrication of water reactor fuels (Fabrication of UO 2 pellets, Fabrication of MOX (mixed uranium-plutonium oxide) pellets, Fabrication of claddings); In-reactor behavior of UO 2 and MOX fuels (Irradiation conditions during nominal operation, Heat generation, and removal, The processes involved at the start of irradiation, Fission gas behavior, Microstructural changes); Water reactor fuel behavior in loss of tightness conditions (Cladding, the first containment barrier, Causes of failure, Consequences of a failure); Microscopic morphology of fuel ceramic and its evolution under irradiation; Migration and localization of fission products in UOX and MOX matrices (The ceramic under irradiation

  5. Nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

    Beauvy, M.; Berthoud, G.; Defranceschi, M.; Ducros, G.; Guerin, Y.; Limoge, Y.; Madic, Ch.; Santarini, G.; Seiler, J.M.; Sollogoub, P.; Vernaz, E.; Guillet, J.L.; Ballagny, A.; Bechade, J.L.; Bonin, B.; Brachet, J.Ch.; Delpech, M.; Dubois, S.; Ferry, C.; Freyss, M.; Gilbon, D.; Grouiller, J.P.; Iracane, D.; Lansiart, S.; Lemoine, P.; Lenain, R.; Marsault, Ph.; Michel, B.; Noirot, J.; Parrat, D.; Pelletier, M.; Perrais, Ch.; Phelip, M.; Pillon, S.; Poinssot, Ch.; Vallory, J.; Valot, C.; Pradel, Ph.; Bonin, B.; Bouquin, B.; Dozol, M.; Lecomte, M.; Vallee, A.; Bazile, F.; Parisot, J.F.; Finot, P.; Roberts, J.F

    2009-07-01

    Fuel is one of the essential components in a reactor. It is within that fuel that nuclear reactions take place, i.e. fission of heavy atoms, uranium and plutonium. Fuel is at the core of the reactor, but equally at the core of the nuclear system as a whole. Fuel design and properties influence reactor behavior, performance, and safety. Even though it only accounts for a small part of the cost per kilowatt-hour of power provided by current nuclear power plants, good utilization of fuel is a major economic issue. Major advances have yet to be achieved, to ensure longer in-reactor dwell-time, thus enabling fuel to yield more energy; and improve ruggedness. Aside from economics, and safety, such strategic issues as use of plutonium, conservation of resources, and nuclear waste management have to be addressed, and true technological challenges arise. This Monograph surveys current knowledge regarding in-reactor behavior, operating limits, and avenues for R and D. It also provides illustrations of ongoing research work, setting out a few noteworthy results recently achieved. Content: 1 - Introduction; 2 - Water reactor fuel: What are the features of water reactor fuel? 9 (What is the purpose of a nuclear fuel?, Ceramic fuel, Fuel rods, PWR fuel assemblies, BWR fuel assemblies); Fabrication of water reactor fuels (Fabrication of UO{sub 2} pellets, Fabrication of MOX (mixed uranium-plutonium oxide) pellets, Fabrication of claddings); In-reactor behavior of UO{sub 2} and MOX fuels (Irradiation conditions during nominal operation, Heat generation, and removal, The processes involved at the start of irradiation, Fission gas behavior, Microstructural changes); Water reactor fuel behavior in loss of tightness conditions (Cladding, the first containment barrier, Causes of failure, Consequences of a failure); Microscopic morphology of fuel ceramic and its evolution under irradiation; Migration and localization of fission products in UOX and MOX matrices (The ceramic under

  6. Inspection system for Zircaloy clad fuel rods

    International Nuclear Information System (INIS)

    Yancey, M.E.; Porter, E.H.; Hansen, H.R.

    1975-10-01

    A description is presented of the design, development, and performance of a remote scanning system for nondestructive examination of fuel rods. Characteristics that are examined include microcracking of fuel rod cladding, fuel-cladding interaction, cladding thickness, fuel rod diameter variation, and fuel rod bowing. Microcracking of both the inner and outer fuel rod surfaces and variations in wall thickness are detected by using a pulsed eddy current technique developed by Argonne National Laboratory (ANL). Fuel rod diameter variation and fuel rod bowing are detected by using two linear variable differential transformers (LVDTs) and a signal conditioning system. The system's mechanical features include variable scanning speeds, a precision indexing system, and a servomechanism to maintain proper probe alignment. Initial results indicate that the system is a very useful mechanism for characterizing irradiated fuel rods

  7. Lumped-parameter fuel rod model for rapid thermal transients

    International Nuclear Information System (INIS)

    Perkins, K.R.; Ramshaw, J.D.

    1975-07-01

    The thermal behavior of fuel rods during simulated accident conditions is extremely sensitive to the heat transfer coefficient which is, in turn, very sensitive to the cladding surface temperature and the fluid conditions. The development of a semianalytical, lumped-parameter fuel rod model which is intended to provide accurate calculations, in a minimum amount of computer time, of the thermal response of fuel rods during a simulated loss-of-coolant accident is described. The results show good agreement with calculations from a comprehensive fuel-rod code (FRAP-T) currently in use at Aerojet Nuclear Company

  8. Preliminary design report for prototypical spent nuclear fuel rod consolidation equipment

    International Nuclear Information System (INIS)

    Judson, B.F.; Maillet, J.; O'Neill, G.L.; Tsitsichvili, J.; Tucoulat, D.

    1986-12-01

    The purpose of the Prototypical Consolidation Demonstration Project (PCDP) is to develop and demonstrate the equipment system that will be used to consolidate the bulk of the spent nuclear fuel generated in the United States prior to its placement in a geological repository. The equipment must thus be capable of operating on a routine production basis over a long period of time with stringent requirements for safety, reliability, productivity and cost-effectiveness. Four phases are planned for the PCDP. Phase 1 is the Preliminary Design of generic consolidation equipment that could be installed at a Monitored Retrievable Storage (MRS) facility or in the Receiving ampersand Handling Facility at a geologic repository site. Phase 2 will be the Final Design and preparation of procurement packages for the equipment in a configuration capable of being installed and tested in a special enclosure within the TAN Hot Shop at DOE's Idaho National Engineering Laboratory. In Phase 3 the equipment will be fabricated and then tested with mock fuel elements in a contractor's facility. Finally, in Phase 4 the equipment will be moved to the TAN facility for demonstration operation with irradiated spent fuel elements. 55 figs., 15 tabs

  9. URANUS - a computer programme for the thermal and mechanical analysis of the fuel rods in a nuclear reactor

    International Nuclear Information System (INIS)

    Lassmann, K.

    1978-01-01

    The URANUS code, a digital computer programme for the thermal and mechanical analysis of integral fuel rods, is described. With this code the fuel rods found in the majority of power reactors can be analyzed. URANUS is built around a quasi two-dimensional analysis of fuel and cladding. The mechanical analysis can accommodate seven components of strain: elastic, time-independent plastic, creep and thermal strains, as well as strains due to swelling, cracking and densification. The heat generation and temperature distribution, cladding/fuel gap closure, pellet cracking and crack healing, fission-gas release, corrosion, O/M-distribution and plutonium redistribution are modelled. Geometric non-linearities (large displacements) are included; steady state or transient loading (pressure, temperature) is possible. In this paper special attention is paid to a theory for determining crack structures. The present status of the URANUS computer programme and a critical comparison with other fuel rod codes as well as sample analyses are given. (Auth.)

  10. The study on the methods for improving the gredibility of NDT equipment for the gap of pellets of nuclear fuel rods

    International Nuclear Information System (INIS)

    Zhang Lei; Liu Ming; Wang Changhong; Ma Jinbo

    2014-01-01

    In order to improve the credibility of the new generation of automatic online non-destructive testing equipment for the gap of the pellets of nuclear fuel rods the researchers have done a lot of work in the development of the device. Such measures as multi-thread synchronization, precise timing, upper and lower computer communication control, antijamming processing are adopted such that the detecting device can accurately detecte the size of the gap between pellets, the position and length of the spring cavity at the front end of the nuclear fuel rods at a detection rate of 8 m/min. The detection credibility for the 0.5 mm gap is over 95%, reaching the international advanced level. At present, the device is put into use in the nuclear fuel element production line. (authors)

  11. SCORE-EVET: a computer code for the multidimensional transient thermal-hydraulic analysis of nuclear fuel rod arrays. [BWR; PWR

    Energy Technology Data Exchange (ETDEWEB)

    Benedetti, R. L.; Lords, L. V.; Kiser, D. M.

    1978-02-01

    The SCORE-EVET code was developed to study multidimensional transient fluid flow in nuclear reactor fuel rod arrays. The conservation equations used were derived by volume averaging the transient compressible three-dimensional local continuum equations in Cartesian coordinates. No assumptions associated with subchannel flow have been incorporated into the derivation of the conservation equations. In addition to the three-dimensional fluid flow equations, the SCORE-EVET code ocntains: (a) a one-dimensional steady state solution scheme to initialize the flow field, (b) steady state and transient fuel rod conduction models, and (c) comprehensive correlation packages to describe fluid-to-fuel rod interfacial energy and momentum exchange. Velocity and pressure boundary conditions can be specified as a function of time and space to model reactor transient conditions such as a hypothesized loss-of-coolant accident (LOCA) or flow blockage.

  12. Encapsulating spent nuclear fuel

    International Nuclear Information System (INIS)

    Fleischer, L.R.; Gunasekaran, M.

    1979-01-01

    A system is described for encapsulating spent nuclear fuel discharged from nuclear reactors in the form of rods or multi-rod assemblies. The rods are completely and contiguously enclosed in concrete in which metallic fibres are incorporated to increase thermal conductivity and polymers to decrease fluid permeability. This technique provides the advantage of acceptable long-term stability for storage over the conventional underwater storage method. Examples are given of suitable concrete compositions. (UK)

  13. Transient three-dimensional thermal-hydraulic analysis of nuclear reactor fuel rod arrays: general equations and numerical scheme

    International Nuclear Information System (INIS)

    Wnek, W.J.; Ramshaw, J.D.; Trapp, J.A.; Hughes, E.D.; Solbrig, C.W.

    1975-11-01

    A mathematical model and a numerical solution scheme for thermal-hydraulic analysis of fuel rod arrays are given. The model alleviates the two major deficiencies associated with existing rod array analysis models, that of a correct transverse momentum equation and the capability of handling reversing and circulatory flows. Possible applications of the model include steady state and transient subchannel calculations as well as analysis of flows in heat exchangers, other engineering equipment, and porous media

  14. Nuclear reactor control rod

    International Nuclear Information System (INIS)

    Cearley, J.E.; Izzo, K.R.

    1987-01-01

    This patent describes a vertically oriented bottom entry control rod from a nuclear reactor: a frame including an elongated central spine of cruciform cross section connected between an upper support member and a lower support member both of cruciform shape having four laterally extending arms. The arms are in alignment with the arms of the lower support member and each aligned upper and lower support members has a sheath extending between; absorber plates of neutron absorber material, different from the material of the frame, one of the absorber plates is positioned within a sheath beneath each of the arms; attachment means suspends the absorber plates from the arms of the upper support member within a sheath; elongated absorber members positioned within a sheath between each of the suspended absorber plates and an arm of the lower support member; and joint means between the upper ends of the absorber members and the lower ends of the suspended absorber plates for minimizing gaps; the sheath means encloses the suspended absorber plates and the absorber members extending between aligned arms of the upper and lower support members and secured

  15. Reactor fuel rod

    International Nuclear Information System (INIS)

    Inui, Mitsuhiro; Mori, Kazuma.

    1990-01-01

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

  16. Nuclear reactor fuel assembly

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  17. Fabrication of the instrumented fuel rods for the 3-Pin Fuel Test Loop at HANARO

    International Nuclear Information System (INIS)

    Sohn, Jae Min; Park, Sung Jae; Shin, Yoon Tag; Lee, Jong Min; Ahn, Sung Ho; Kim, Soo Sung; Kim, Bong Goo; Kim, Young Ki; Lee, Ki Hong; Kim, Kwan Hyun

    2008-09-01

    The 3-Pin Fuel Test Loop(hereinafter referred to as the '3-Pin FTL') facility has been installed at HANARO(High-flux Advanced Neutron Application Reactor) and the 3-Pin FTL is under a test operation. The purpose of this report is to fabricate the instrumented fuel rods for the 3-Pin FTL. The fabrication of these fuel rods was based on experiences and technologies of the instrumented fuel rods for an irradiation fuel capsule. The three instrumented fuel rods of the 3-Pin FTL have been designed. The one fuel rod(180 .deg. ) was designed to measure the centerline temperature of the nuclear fuels and the internal pressure of the fuel rod, and others(60 .deg. and 300 .deg. ) were designed to measure the centerline temperature of the fuel pellets. The claddings were made of the reference material 1 and 2 and new material 1 and 2. And nuclear fuel was used UO 2 (2.0w/o) pellet type with large grain and standard grain. The major procedures of fabrication are followings: (1) the assembling and weld of fuel rods with the pellet mockups and the sensor mockups for the qualification tests, (2) the qualification tests(dimension measurements, tensile tests, metallography examinations and helium leak tests) of weld, (3) the assembling and weld of instrumented fuel rods with the nuclear pellets and the sensors for the irradiation test, and (4) the qualification tests(the helium leak test, the dimensional measurement, electric resistance measurements of sensors) of test fuel rods. Satisfactory results were obtained for all the qualification tests of the instrumented fuel rods for the 3-Pin FTL. Therefore the three instrumented fuel rods for the 3-Pin FTL have been fabricated successfully. These will be installed in the In-Pile Section of 3-Pin FTL. And the irradiation test of these fuel rods is planned from the early next year for about 3 years at HANARO

  18. Grab structure of a lifting structure in particular for use in a nuclear reactor for lifting and lowering of fuel elements and fuel rods

    International Nuclear Information System (INIS)

    Dose, G.

    1979-01-01

    A guide tower projects perpendicularly downward from the carriage of the charging machine. It can be rotated about its perpendicular axis. The tower is used to displace a hollow grab structure with two grabs. They can be opened and closed, the closed position being retained as long as they carry the fuel elements or rods. The power and interlocking equipment is installed one unit above the other in the joint grab housing. The tower with the integrated fuel element grab and the rod grab is rotated about its perpendicular axis for inspection of the fuel elements or rods. (DG) [de

  19. Internal attachment of laser beam welded stainless steel sheathed thermocouples into stainless steel upper end caps in nuclear fuel rods for the LOFT Reactor

    International Nuclear Information System (INIS)

    Welty, R.K.; Reid, R.D.

    1980-01-01

    The Exxon Nuclear Company, Inc., acting as a subcontractor to EG and G Idaho Inc., Idaho National Engineering Laboratory, Idaho Falls, Idaho, conducted a laser beam welding study to attach internal stainless steel thermocouples into stainless steel upper end caps in nuclear fuel rods. The objective of this study was to determine the feasibility of laser welding a single 0.063 inch diameter stainless steel (304) sheathed thermocouple into a stainless steel (316) upper end cap for nuclear fuel rods. A laser beam was selected because of the extremely high energy input in unit volume that can be achieved allowing local fusion of a small area irrespective of the difference in material thickness to be joined. A special weld fixture was designed and fabricated to hold the end cap and the thermocouple with angular and rotational adjustment under the laser beam. A commercial pulsed laser and energy control system was used to make the welds

  20. Cutting method and cutting device for spent fuel rod of nuclear reactor

    International Nuclear Information System (INIS)

    Komatsu, Masahiko; Ose, Toshihiko.

    1996-01-01

    A control rod transferred under water in a vertically suspended state is postured horizontally at such a water depth that radiations can be shielded, and then it is cut to a dropping speed limiting portion and a cross-like main body. The separated cross-like main body portion is further cut in the longitudinal direction and separated into a pair of cut pieces each having an L-shaped cross section. A disk like metal saw is used as a cutting tool. Alternatively, a plasma jet cutter or a melting-type water jet cutter is used as a cutting tool. Then, since the spent control rod to be cut is postured horizontally under water, the water depth for the cutting position can be reduced. As a result, the cutting state using the cutting tool can be observed by naked eyes from the position above the water surface thereby enabling to perform the cutting operation reliably. (N.H.)

  1. Performance assessment of the RANS turbulence models in nuclear fuel rod bundles

    International Nuclear Information System (INIS)

    In, Wang Kee; Chun, Tae Hyun; Oh, Dong Seok; Shin, Chang Hwan

    2005-02-01

    The three experiments for turbulent flow in a rod bundle geometry were simulated in this CFD analysis using various RANS models. The CFD predictions were compared with the experimental and DNS results. The RANS models used here are the nonlinear quadratic/cubic κ-ε models and the second-order closure models (SSG, LRR, RSM-ω). The anisotropic models predicted the secondary flow and showed a significantly improved agreement with the measurements from the standard κ-ε model. In particular, the SSG model resulted in the best performance showing the closest agreement with the experimental results. However, the RANS models could not predict the very high anisotropy observed in a rod bundle with a small pitch-to-diameter ratio

  2. Fuel followed control rod installation at AFRRI

    International Nuclear Information System (INIS)

    Moore, Mark; Owens, Chris; Forsbacka, Matt

    1992-01-01

    Fuel Followed Control Rods (FFCRs) were installed at the Armed Forces Radiobiology Research Institute's 1 MW TRIGA Reactor. The procedures for obtaining, shipping, and installing the FFCRs is described. As part of the FFCR installation, the transient rod drive was relocated. Core performance due to the addition of the fuel followed control rods is discussed. (author)

  3. Detection of failed fuel rods in shrouded BWR fuel assemblies

    International Nuclear Information System (INIS)

    Baero, G.; Boehm, W.; Goor, B.; Donnelly, T.

    1988-01-01

    A manipulator and an ultrasonic testing (UT) technique were developed to identify defective fuel rods in shrouded BWR fuel assemblies. The manipulator drives a UT probe axially through the bottom tie plate into the water channels between the fuel rods. The rotating UT probe locates defective fuel rods by ingressed water which attenuates the UT-signal. (author)

  4. Characterization of hydrogen, nitrogen, oxygen, carbon and sulfur in nuclear fuel (UO2) and cladding nuclear rod materials

    International Nuclear Information System (INIS)

    Crewe, Maria Teresa I.; Lopes, Paula Corain; Moura, Sergio C.; Sampaio, Jessica A.G.; Bustillos, Oscar V.

    2011-01-01

    The importance of Hydrogen, Nitrogen, Oxygen, Carbon and Sulfur gases analysis in nuclear fuels such as UO 2 , U 3 O 8 , U 3 Si 2 and in the fuel cladding such as Zircaloy, is a well known as a quality control in nuclear industry. In UO 2 pellets, the Hydrogen molecule fragilizes the metal lattice causing the material cracking. In Zircaloy material the H2 molecules cause the boiling of the cladding. Other gases like Nitrogen, Oxygen, Carbon and Sulfur affect in the lattice structure change. In this way these chemical compounds have to be measure within specify parameters, these measurement are part of the quality control of the nuclear industry. The analytical procedure has to be well established by a convention of the quality assurance. Therefore, the Oxygen, Carbon, Sulfur and Hydrogen are measured by infrared absorption (IR) and the nitrogen will be measured by thermal conductivity (TC). The gas/metal analyzer made by LECO Co. model TCHEN-600 is Hydrogen, Oxygen and Nitrogen analyzer in a variety of metals, refractory and other inorganic materials, using the principle of fusion by inert gas, infrared and thermo-coupled detector. The Carbon and Sulfur compounds are measure by LECO Co. model CS-400. A sample is first weighed and placed in a high purity graphite crucible and is casted on a stream of helium gas, enough to release the oxygen, nitrogen and hydrogen. During the fusion, the oxygen present in the sample combines with the carbon crucible to form carbon monoxide. Then, the nitrogen present in the sample is analyzed and released as molecular nitrogen and the hydrogen is released as gas. The hydrogen gas is measured by infrared absorption, and the sample gases pass through a trap of copper oxide which converts CO to CO 2 and hydrogen into water. The gases enter the cell where infrared water content is then converted making the measurement of total hydrogen present in the sample. The Hydrogen detection limits for the nuclear fuel is 1 μg/g for the Nitrogen

  5. Gray rod for a nuclear reactor

    International Nuclear Information System (INIS)

    Francis, T.A.; Cerni, Samuel.

    1986-01-01

    The invention relates to an improved gray rod for insertion in a nuclear fuel assembly having an array of fuel rods. The gray rod includes a thin-walled cladding tube a first longitudinal section of which is positioned within, and a second longitudinal section of which is positioned essentially without, the array of fuel rods when the gray rod is inserted in the fuel assembly. The first longitudinal section defines a pellet-receiving space having detained therein a stack of annular pellets with an outer diameter sufficient to lend radial support to the wall of the first longitudinal tube section. The second longitudinal section defines a hollow space devoid of pellets and having means to resist radial collapse under external pressure. This means may be a partially compressed spiral spring which serves the dual purpose of retaining the stack of pellets in the pellet-receiving space and of lending radial support to the wall of the second longitudinal tube section or it may be holes through the wall to allow pressure equalisation. The cladding tube is composed of stainless-steel material having a low neutron-capture cross-section, and the annular pellets preferably being composed of Zircaloy or Zirconia material. (author)

  6. Fuel rod simulator effects in flooding experiments single rod tests

    International Nuclear Information System (INIS)

    Nishida, M.

    1984-09-01

    The influence of a gas filled gap between cladding and pellet on the quenching behavior of a PWR fuel rod during the reflood phase of a LOCA has been investigated. Flooding experiments were conducted with a short length electrically heated single fuel rod simulator surrounded by glass housing. The gap of 0.05 mm width between the Zircaloy cladding and the internal Al 2 O 3 pellets of the rod was filled either wit helium or with argon to vary the radial heat resistance across the gap. This report presents some typical data and an evaluation of the reflood behavior of the fuel rod simulator used. The results show that the quench front propagates faster for increasing heat resistance in the gap between cladding and heat source of the rod. (orig.) [de

  7. Forming Limit Diagrams of Zircaloy-4 and Zirlo Sheets for Stamping of Spacer Grids of Nuclear Fuel Rods

    International Nuclear Information System (INIS)

    Seo, Yun Mi; Hyun, Hong Chul; Lee, Hyung Yil; Kim, Nak Soo

    2011-01-01

    In this work, we investigated the theoretical forming limit models for Zircaloy-4 and Zirlo used for spacer grid of nuclear fuel rods. Tensile and anisotropy tests were performed to obtain stress-strain curves and anisotropic coefficients. The experimental forming limit diagrams (FLD) for two materials were obtained by dome stretching tests following NUMISHEET 96. Theoretical FLD depends on FL models and yield criteria. To obtain the right hand side (RHS) of FLD, we applied the FL models (Swift's diffuse necking, M-K theory, S-R vertex theory) to Zircaloy-4 and Zirlo sheets. Hill's local necking theory was adopted for the left hand side (LHS) of FLD. To consider the anisotropy of sheets, the yield criteria of Hill and Hosford were applied. Comparing the predicted curves with the experimental data, we found that the RHS of FLD for Zircaloy-4 can be described by the Swift model (with the Hill's criterion), while the LHS of the FLD can be explained by Hill model. The FLD for Zirlo can be explained by the S-R model and the Hosford's criterion (a = 8)

  8. Analysis on Propagation Characteristics and Experimental Verification of A1 Circumferential Waves in Nuclear Fuel Rods Coated with Oxide Layers

    International Nuclear Information System (INIS)

    Joo, Young Sang; Jung, Hyun Kyu; Cheong, Yong Moo; Ih, Jeong Guon

    1999-01-01

    The resonance scattering of acoustic waves from the cylindrical shells of nuclear fuel rods coated with oxide layers has been theoretically modeled and numerically analyzed for the propagation characteristics of the circumferential waves. The normal mode solutions of the scattering pressure of the coated shells have been obtained. The pure resonance components have been isolated using the newly proposed inherent background coefficients. The propagation characteristics of resonant circumferential waves for the shells coated with oxide layers are affected by the presence and the thickness of an oxide layer. The characteristics have been experimentally confirmed through the method of isolation and identification of resonances. The change of the phase velocity of the A 1 circumferential wave mode for the coated shell is negligible at the specified partial waves in spite of the presence of the oxide layer and the increase in coating thickness. Utilizing the invariability characteristics of the phase velocity of the A 1 mode, the oxide layer thickness of the coated shells can be estimated. A new nondestructive technique for the relative measurement of the coating thickness of coated shells has been proposed

  9. A Study on the Sliding/Impact Wear of a Nuclear Fuel Rod in Room Temperature Air: (I) Development of a Test Rig and Characteristic Analysis

    International Nuclear Information System (INIS)

    Lee, Young Ho; Lee, Kang Hee; Kim, Hyung Kyu

    2007-01-01

    A new type of a fretting wear tester has been designed and developed in order to simulate the actual vibration behavior of a nuclear fuel rod for springs/dimples in room temperature. When considering the actual contact condition between fuel rod and spring/dimple, if fretting wear progress due to the Flow-Induced Vibration (FIV) under a specific normal load exerted on the fuel rod by the elastic deformation of the spring, the contacting force between the fuel rod and dimple that were located in the opposite side should be decreased. Consequently, the evaluation of developed spacer grids against fretting wear damage should be performed with the results of a cell unit experiments because the contacting force is one of the most important variables that influence to the fretting wear mechanism. Therefore, it is necessary to develop a new type of fretting test rig in order to simulate the actual contact condition. In this paper, the development procedure of a new fretting wear tester and its performance were discussed in detail

  10. Procedure and apparatus for measuring the radial gap between fuel and surrounding cladding in a fuel rod for a nuclear reactor

    International Nuclear Information System (INIS)

    Olshausen, K.D.

    1976-01-01

    A device is described for measuring non-destructively the annular fuel-cladding gap in an irradiated or fresh fuel rod. The principle applied is that a force is applied to an arm which presses the cladding diametrically, thus deforming it until it touches the fuel pellet. By presenting the values of the force applied and the deformation produced on an XY recorder, the width of the gap is obtained. Alternatively the gap width may be obtained digitally. Since the gap is so small that the deformation is within the elastic range, the fuel rod may be reloaded in the reactor for further irradiation. (JIW)

  11. Pressurized water reactor fuel rod design methodology

    International Nuclear Information System (INIS)

    Silva, A.T.; Esteves, A.M.

    1988-08-01

    The fuel performance program FRAPCON-1 and the structural finite element program SAP-IV are applied in a pressurized water reactor fuel rod design methodology. The applied calculation procedure allows to dimension the fuel rod components and characterize its internal pressure. (author) [pt

  12. 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.)

  13. Three Transducer Ultrasonic Examination of Nuclear Fuel Rod Flush Welds At ENUSA

    International Nuclear Information System (INIS)

    Domingo, A.; Jimenez, J.M.

    1998-01-01

    From 1991 ENUSA are using UT microscope examination of flush weld with longitudinal wave (perpendicular to weld) looking for welding defects as porosity, weld thickness and penetration. In 1994 we included a new transducer off set placed, with shear waves to control cracks in welding. In 1997 we incorporated a new shear transducer, 30 degree centigrade angle, in order to control different orientation of cracks or Grain Boundary Separation (GBS) and to improve software capabilities. Then actual UT microscope equipment used to inspect rod welds consist, mainly in three transducer of 50 MHz mounted over a rotatory head. UT system is electronically synchronized and obtains 750 data points per revolution by transducer. A set of seven images of approx 100.000 data points is obtained for each weld. Thickness, pore size, pore depth, two of crack size and two of crack depth are presented and evaluated evaluated by the computer to obtain weld disposition. Resolution of 0,05 mm pore size is achieved by this equipment, thickness and penetration precision should be in the order of 0,005 mm and 0,05 mm respectively. Crack detection depend basically on its orientation, nevertheless position of transducer assures a high capability detection of cracks which should be formed at these welds. (Author)

  14. Critical experiments supporting underwater storage of tightly packed configurations of spent fuel rods

    International Nuclear Information System (INIS)

    Hoovler, G.S.; Baldwin, M.N.

    1981-04-01

    Criticla arrays of 2.5%-enriched UO 2 fuel rods that simulate underwater rod storage of spent power reactor fuel are being constructed. Rod storage is a term used to describe a spent fuel storage concept in which the fuel bundles are disassembled and the rods are packed into specially designed cannisters. Rod storage would substantially increase the amount of fuel that could be stored in available space. These experiments are providing criticality data against which to benchmark nuclear codes used to design tightly packed rod storage racks

  15. Fuel element for nuclear reactors

    International Nuclear Information System (INIS)

    Cadwell, D.J.

    1982-01-01

    The invention concerns a fuel element for nuclear reactors with fuel rods and control rod guide tubes, where the control rod guide tubes are provided with flat projections projecting inwards, in the form of local deformations of the guide tube wall, in order to reduce the radial play between the control rod concerned and the guide tube, and to improve control rod movement. This should ensure that wear on the guide tubes is largely prevented which would be caused by lateral vibration of the control rods in the guide tubes, induced by the flow of coolant. (orig.) [de

  16. Cost targets for at-reactor spent fuel rod consolidation

    International Nuclear Information System (INIS)

    Macnabb, W.V.

    1985-01-01

    The high-level nuclear waste management system in the US currently envisions the disposal of spent fuel rods that have been removed from their assemblies and reconfigured into closely packed arrays. The process of fuel rod removal and packaging, referred to as rod consolidation, can occur either at reactors or at an integrated packaging facility, monitored retrievable storage (MRS). Rod consolidation at reactors results in cost savings down stream of reactors by reducing needs for additional storage, reducing the number of shipments, and reducing (eliminating, in the extreme) the amount of fuel handling and consolidation at the MRS. These savings accrue to the nuclear waste fund. Although private industry is expected to pay for at-reactor activities, including rod consolidation, it is of interest to estimate cost savings to the waste system if all fuel were consolidated at reactors. If there are savings, the US Department of Energy (DOE) may find it advantageous to pay for at-reactor rod consolidation from the nuclear waste fund. This paper assesses and compares the costs of rod consolidation at reactors and at the MRS in order to determine at what levels the former could be cost competitive with the latter

  17. Microcomputer system for controlling fuel rod length

    International Nuclear Information System (INIS)

    Meyer, E.R.; Bouldin, D.W.; Bolfing, B.J.

    1979-01-01

    A system is being developed at the Oak Ridge National Laboratory (ORNL) to automatically measure and control the length of fuel rods for use in a high temperature gas-cooled reactor (HTGR). The system utilizes an LSI-11 microcomputer for monitoring fuel rod length and for adjusting the primary factor affecting length. Preliminary results indicate that the automated system can maintain fuel rod length within the specified limits of 1.940 +- 0.040 in. This system provides quality control documentation and eliminates the dependence of the current fuel rod molding process on manual length control. In addition, the microcomputer system is compatible with planned efforts to extend control to fuel rod fissile and fertile material contents

  18. Nuclear fuel pellet loading machine

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  19. External attachment of titanium sheathed thermocouples to zirconium nuclear fuel rods for the loss-of-fluid-test (LOFT) Reactor

    International Nuclear Information System (INIS)

    Welty, R.K.

    1980-01-01

    A welding process to attach titanium sheathed thermocouples to the outside of the zircaloy clad fuel rods has been developed. A laser beam was selected as the optimum welding process because of the extremely high energy input per unit volume that can be achieved allowing local fusion of a small area irrespective of the difference in material thickness to be joined. Irradiation tests showed no degradation of thermocouples or weld structure. Fast thermal cycle and heater rod blowdown reflood tests were made to subject the weldments to high temperatures, high pressure steam, and fast water quench cycles. From the behavior of these tests, it was concluded that the attachment welds would survive a series of reactor safety tests. 2 refs

  20. Water rod and fuel assembly

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  1. Nuclear fuel rod

    International Nuclear Information System (INIS)

    Kaneuji, Akira; Akitake, Jun-ichi; Kondo, Yoshiaki.

    1997-01-01

    A buckling member is inserted between a pellet at the lowermost portion and a lower end plug. The buckling member comprises a buckling pipe and an upper lid which engage coaxially with each other. The buckling pipe is a stainless hollow cylindrical pipe, and an expanding portion is formed at the intermediate portion. If pellets and a cladding tube are adhered by mechanical interaction during reactor operation, the thermal expansion of pellets at a portion upper than the adhered portion is absorbed by a coil spring. The thermal expansion of the pellets at a portion lower than the adhered portion is absorbed to the buckling member, and if the compression load is increased exceeding a predetermined value, the buckling pipe is bucked at an expandable portion and shrinks to absorb thermal expansion of pellets. (I.N.)

  2. Effects of alloys elements, impurities and microstructural factors in austenitic stainless steel to utilize in fuel rod of nuclear reactors

    International Nuclear Information System (INIS)

    Yoshimoto, A.

    1988-08-01

    Austenitic Stainless Steel is used as cladding material of pressurized water reactor fuel rods because of its good performance. The addition of alloy elements and the control of impurities make this to happen. Fission products do not contribute to corrosion. Dimensional changes are not critical up to 1,0 x 10 22 n/cm 2 (E>0,1 MeV) of neutronic doses. The hydrogen does not cause embrittlement in the reactor operation temperatures, and helium contributes to embrittlement if the material is warmed upon 650 0 C. (author) [pt

  3. 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.)

  4. Numerical prediction of critical heat flux in nuclear fuel rod bundles with advanced three-fluid multidimensional porous media based model

    International Nuclear Information System (INIS)

    Zoran Stosic; Vladimir Stevanovic

    2005-01-01

    Full text of publication follows: The modern design of nuclear fuel rod bundles for Boiling Water Reactors (BWRs) is characterised with increased number of rods in the bundle, introduced part-length fuel rods and a water channel positioned along the bundle asymmetrically in regard to the centre of the bundle cross section. Such design causes significant spatial differences of volumetric heat flux, steam void fraction distribution, mass flux rate and other thermal-hydraulic parameters important for efficient cooling of nuclear fuel rods during normal steady-state and transient conditions. The prediction of the Critical Heat Flux (CHF) under these complex thermal-hydraulic conditions is of the prime importance for the safe and economic BWR operation. An efficient numerical method for the CHF prediction is developed based on the porous medium concept and multi-fluid two-phase flow models. Fuel rod bundle is observed as a porous medium with a two-phase flow through it. Coolant flow from the bundle entrance to the exit is characterised with the subsequent change of one-phase and several two-phase flow patterns. One fluid (one-phase) model is used for the prediction of liquid heating up in the bundle entrance region. Two-fluid modelling approach is applied to the bubbly and churn-turbulent vapour and liquid flows. Three-fluid modelling approach is applied to the annular flow pattern: liquid film on the rods wall, steam flow and droplets entrained in the steam stream. Every fluid stream in applied multi-fluid models is described with the mass, momentum and energy balance equations. Closure laws for the prediction of interfacial transfer processes are stated with the special emphasis on the prediction of the steam-water interface drag force, through the interface drag coefficient, and droplets entrainment and deposition rates for three-fluid annular flow model. The model implies non-equilibrium thermal and flow conditions. A new mechanistic approach for the CHF prediction

  5. LOFT fuel rod pressure measurement

    International Nuclear Information System (INIS)

    Billeter, T.R.

    1979-01-01

    Pressure sensors selected for measuring fuel rod pressure within the LOFT reactor exhibited stable, repeatable operating characteristics during calibrations at temperatures up to 800 0 F and pressures to 2500 psig. All sensors have a nominal sensitivity of .5 millivolts per psi, decreasing monotonically with temperature. Output signal increases linearly with increasing pressure up to 2000 psig. For imposed slow and rapid temperature variations and for pressure applied during these tests, the sensor indicates a pressure at variance with the actual value by up to 15% of reading. However, the imposed temperature rates of change often exceeded the value of -10 0 F/sec. specified for LOFT. The series of tests in an autoclave permit creation of an environment most closely resembling sensor operating conditions within LOFT. For multiple blowdowns and for longtime durations the sensor continued to provide pressure-related output signals. For temperature rates up to -87 0 F/sec, the indicated pressure measurement error remained less than 13% of reading. Adverse effects caused by heating the 1/16 inch O.D. signal cable to 800 0 F contributed only insignificantly to the noted pressure measurement error

  6. Final results of the cadmium and spectral ratios obtained inside of the fuel rod positioned in the central position of the IPEN/MB-01 nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bitelli, Ulysses d' Utra; Mura, Luiz Ernesto C.; Santos, Diogo Feliciano dos, E-mail: ubitelli@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Lambiasi, Beatriz G.N. [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), SP (Brazil)

    2015-07-01

    The spectral ratios are very important to determine some nuclear reactors parameters such as reaction rates, fuel lifetime, etc and some safety operational conditions. This study aims to determine the spectral ratios in 2 (two) spatial positions located inside the core of the Nuclear Reactor IPEN/MB-01. These places are at the central position of the nuclear reactor core in an asymptotic neutron flux region. The experiment consists in inserting different activation foil detectors inside an experimental fuel rod. The experimental rod is assembled at the central position of the reactor core. Activation neutron foil detectors of different elements such as {sup 197}Au, {sup 238}U, {sup 45}Sc, {sup 58}Ni, {sup 24}Mg, {sup 47}Ti and {sup 115m}In were used to cover a large range of neutron spectrum. Saturation activity per target nucleus was obtained by gamma spectrometry using a HPGe system. The experimental cadmium ratios compared with values computed by MCNP-4C code show good agreement. (author)

  7. Nuclear fuel element

    International Nuclear Information System (INIS)

    1974-01-01

    A nuclear fuel element for use in the core of a nuclear reactor is disclosed. A heat conducting fission product retaining metal liner of a refractory metal is incorporated in the fuel element between the cladding and the nuclear fuel to inhibit mechanical interaction between the nuclear fuel and the cladding, to isolate fission products and nuclear fuel impurities from contacting the cladding, and to improve the axial thermal peaking gradient along the length of the fuel rod. The metal liner can be in the form of a tube or hollow cylindrical column, a foil of single or multiple layers in the shape of a hollow cylindrical column, or a coating on the internal surface of the cladding. Preferred refractory metal materials are molybdenum, tungsten, rhenium, niobium and alloys of the foregoing metals

  8. Nuclear fuel element

    International Nuclear Information System (INIS)

    Thompson, J.R.; Rowland, T.C.

    1976-01-01

    A nuclear fuel element for use in the core of a nuclear reactor is disclosed. A heat conducting, fission product retaining metal liner of a refractory metal is incorporated in the fuel element between the cladding and the nuclear fuel to inhibit mechanical interaction between the nuclear fuel and the cladding, to isolate fission products and nuclear fuel impurities from contacting the cladding and to improve the axial thermal peaking gradient along the length of the fuel rod. The metal liner can be in the form of a tube or hollow cylindrical column, a foil of single or multiple layers in the shape of a hollow cylindrical column, or a coating on the internal surface of the cladding. Preferred refractory metal materials are molybdenum, tungsten, rhenium, niobium and alloys of the foregoing metals

  9. Nuclear fuel

    International Nuclear Information System (INIS)

    D Hondt, P.

    1998-01-01

    The research and development programme on nuclear fuel at the Belgian Nuclear Research Centre SCK/CEN is described. The objective of this programme is to enhance the quantitative prediction of the operational limits of nuclear fuel and to assess the behaviour of fuel under incidental and accidental conditions. Progress is described in different domains including the modelling of fission gas release in LWR fuel, thermal conductivity, basic physical phenomena, post-irradiation examination for fuel performance assessment, and conceptual studies of incidental and accidental fuel experiments

  10. Fuel rod behaviour at high burnup WWER fuel cycles

    International Nuclear Information System (INIS)

    Medvedev, A.; Bogatyr, S.; Kouznetsov, V.; Khvostov, G.; Lagovsky; Korystin, L.; Poudov, V.

    2003-01-01

    The modernisation of WWER fuel cycles is carried out on the base of complete modelling and experimental justification of fuel rods up to 70 MWd/kgU. The modelling justification of the reliability of fuel rod and fuel rod with gadolinium is carried out with the use of certified START-3 code. START-3 code has a continuous experimental support. The thermophysical and strength reliability of WWER-440 fuel is justified for fuel rod and pellet burnups 65 MWd/kgU and 74 MWd/U, accordingly. Results of analysis are demonstrated by the example of uranium-gadolinium fuel assemblies of second generation under 5-year cycle with a portion of 6-year assemblies and by the example of successfully completed pilot operation of 5-year cycle fuel assemblies during 6 years at unit 3 of Kolskaja NPP. The thermophysical and strength reliability of WWER-1000 fuel is justified for a fuel rod burnup 66 MWd/kgU by the example of fuel operation under 4-year cycles and 6-year test operation of fuel assemblies at unit 1 of Kalininskaya NPP. By the example of 5-year cycle at Dukovany NPP Unit 2 it was demonstrated that WWER fuel rod of a burnup 58 MWd/kgU ensure reliable operation under load following conditions. The analysis has confirmed sufficient reserves of Russian fuel to implement program of JSC 'TVEL' in order to improve technical and economical parameters of WWER fuel cycles

  11. International symposium on fuel rod simulators: development and application

    Energy Technology Data Exchange (ETDEWEB)

    McCulloch, R.W. (comp.)

    1981-05-01

    Separate abstracts are included for each of the papers presented concerning fuel rod simulator operation and performance; simulator design and evaluation; clad heated fuel rod simulators and fuel rod simulators for cladding investigations; fuel rod simulator components and inspection; and simulator analytical modeling. Ten papers have previously been input to the Energy Data Base.

  12. Investigation of typicality of non-nuclear rod and fuel-clad gap effect during reflood phase, and development of a FEM thermal transient analysis code HETFEM

    International Nuclear Information System (INIS)

    Sudoh, Takashi

    1981-06-01

    The objective of this study are: 1) Evaluate the capability of the electrical heater for simulating the fuel rod during the reflood phase, and 2) To investigate the effect of the clad-fuel gap in the fuel rod on the clad thermal response during the reflood phase. A computer code HETFEM which is the two dimensional transient thermal conductivity analysis code utilized a finite element method is developed for analysing thermal responses of heater and fuel rod. The two kinds of electrical heaters and a fuel rod are calculated with simple boundary conditions. 1) direct heater (former JAERI reflood test heater), 2) indirect heater (FLECHT test heater), 3) fuel rod (15 x 15 type in Westinghouse PWR). The comparison of the clad temperature responses shows the quench time is influenced by the thermal diffusivity and gap conductance. In the conclusion, the ELECHT heater shows atypicality in the clad temperature response and heat releasing rate. But the direct heater responses are similar to those of the fuel rod. For the gap effect on the fuel rod behavior, the lower gap conductance causes sooner quench and less heat releasing rate. This calculation is not considered the precursory cooling which is affected by heat releasing rate at near and below the quench front. Therefore two dimensional calculation with heat transfer related to the local fluid conditions will be needed. (author)

  13. Nondestructive assay of HTGR fuel rods

    International Nuclear Information System (INIS)

    Menlove, H.O.

    1974-01-01

    Performance characteristics of three different radioactive source NDA systems are compared for the assay of HTGR fuel rods and stacks of rods. These systems include the fast neutron Sb-Be assay system, the 252 Cf ''Shuffler,'' and the thermal neutron PAPAS assay system. Studies have been made to determinethe perturbation on the measurements from particle size, kernel Th/U ratio, thorium content, and hydrogen content. In addition to the total 235 U determination, the pellet-to-pellet or rod-to-rod uniformity of HTGR fuel rod stacks has been measured by counting the delayed gamma rays with a NaI through-hole in the PAPAS system. These measurements showed that rod substitutions can be detected easily in a fuel stack, and that detailed information is available on the loading variations in a uniform stack. Using a 1.0 mg 252 Cf source, assay rates of 2 to 4 rods/s are possible, thus facilitating measurement of 100 percent of a plant's throughput. (U.S.)

  14. Development of cutting device for irradiated fuel rod

    International Nuclear Information System (INIS)

    Lee, E. P.; Jun, Y. B.; Hong, K. P.; Min, D. K.; Lee, H. K.; Su, H. S.; Kim, K. S.; Kwon, H. M.; Joo, Y. S.; Yoo, K. S.; Joo, J. S.; Kim, E. K.

    2004-01-01

    Post Irradiation Examination(PIE) on irradiated fuel rods is essential for the evaluation of integrity and irradiation performance of fuel rods of commercial reactor fuel. For PIE, fuel rods should be cut very precisely. The cutting positions selected from NDT data are very important for further destructive examination and analysis. A fuel rod cutting device was developed witch can cut fuel rods longitudinal very precisely and can also cut the fuels into the same length rod cuts repeatedly. It is also easy to remove the fuel cutting powder after cutting works and it can extend the life time of cutting device and lower the contamination level of hot cell

  15. Boosting nuclear fuels

    International Nuclear Information System (INIS)

    Demarthon, F.; Donnars, O.; Dupuy-Maury, F.

    2002-01-01

    This dossier gives a broad overview of the present day status of the nuclear fuel cycle in France: 1 - the revival of nuclear power as a solution to the global warming and to the increase of worldwide energy needs; 2 - the security of uranium supplies thanks to the reuse of weapon grade highly enriched uranium; 3 - the fabrication of nuclear fuels from the mining extraction to the enrichment processes, the fabrication of fuel pellets and the assembly of fuel rods; 4 - the new composition of present day fuels (UO x and chromium-doped pellets); 5 - the consumption of plutonium stocks and the Corail and Apa fuel assemblies for the reduction of plutonium stocks and the preservation of uranium resources. (J.S.)

  16. 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.)

  17. Analysis of Double-encapsulated Fuel Rods

    Energy Technology Data Exchange (ETDEWEB)

    Hales, Jason Dean [Idaho National Laboratory; Medvedev, Pavel G [Idaho National Laboratory; Novascone, Stephen Rhead [Idaho National Laboratory; Perez, Danielle Marie [Idaho National Laboratory; Williamson, Richard L [Idaho National Laboratory

    2014-09-01

    In an LWR fuel rod, the cladding encapsulates the fuel, contains fission products, and transfers heat directly to the water coolant. In some situations, it may be advantageous to separate the cladding from the coolant through use of a secondary cladding or capsule. This may be done to increase confidence that the fuel or fission products will not mix with the coolant, to provide a mechanism for controlling the rod temperature, or to place multiple experimental rodlets within a single housing. With an axisymmetric assumption, it is possible to derive closed-form expressions for the temperature profile in a fuel rod using radially-constant thermal conductivity in the fuel. This is true for both a traditional fuel-cladding rod and a double-encapsulated fuel (fuel, cladding, capsule) configuration. Likewise, it is possible to employ a fuel performance code to analyse both a traditional and a double-encapsulated fuel. In the case of the latter, two sets of gap heat transfer conditions must be imposed. In this work, we review the equations associated with radial heat transfer in a cylindrical system, present analytic and computational results for a postulated power and gas mixture history for IFA-744, and describe the analysis of the AFC-2A, 2B metallic fuel alloy experiments at the Advanced Test Reactor, including the effect of a release of fission products into the cladding-capsule gap. The computational results for these two cases were obtained using BISON, a fuel performance code under development at Idaho National Laboratory.

  18. The results of postirradiation examinations of VVER-1000 and VVER-440 fuel rods

    Science.gov (United States)

    Dubrovin, K. P.; Ivanov, E. G.; Strijov, P. N.; Yakovlev, V. V.

    1991-02-01

    The paper presents the results of postirradiation examination of the fuel rods having different fuel-cladding gaps, pellet densities, pellet inner diameters and so on. The fuel rods were irradiated in the material science reactor (MR) of the Kurchatov Institute of Atomic Energy and at 4 unit of the Novo-Voronezh nuclear powerplant. Some data on fission gas release and rod geometry and compared with computer code predictions.

  19. Nuclear fuels

    International Nuclear Information System (INIS)

    2008-01-01

    The nuclear fuel is one of the key component of a nuclear reactor. Inside it, the fission reactions of heavy atoms, uranium and plutonium, take place. It is located in the core of the reactor, but also in the core of the whole nuclear system. Its design and properties influence the behaviour, the efficiency and the safety of the reactor. Even if it represents a weak share of the generated electricity cost, its proper use represents an important economic stake. Important improvements remain to be made to increase its residence time inside the reactor, to supply more energy, and to improve its robustness. Beyond the economical and safety considerations, strategical questions have to find an answer, like the use of plutonium, the management of resources and the management of nuclear wastes and real technological challenges have to be taken up. This monograph summarizes the existing knowledge about the nuclear fuel, its behaviour inside the reactor, its limits of use, and its R and D tracks. It illustrates also the researches in progress and presents some key results obtained recently. Content: 1 - Introduction; 2 - The fuel of water-cooled reactors: aspect, fabrication, behaviour of UO 2 and MOX fuels inside the reactor, behaviour in loss of tightness situation, microscopic morphology of fuel ceramics and evolution under irradiation - migration and localisation of fission products in UOX and MOX matrices, modeling of fuels behaviour - modeling of defects and fission products in the UO 2 ceramics by ab initio calculations, cladding and assembly materials, pellet-cladding interaction, advanced UO 2 and MOX ceramics, mechanical behaviour of the fuel assembly, fuel during a loss of coolant accident, fuel during a reactivity accident, fuel during a serious accident, fuel management inside reactor cores, fuel cycle materials balance, long-term behaviour of the spent fuel, fuel of boiling water reactors; 3 - the fuel of liquid metal fast reactors: fast neutrons radiation

  20. Nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Nakano, H [Power Reactor and Nuclear Fuel Development Corp., Tokyo (Japan)

    1976-10-01

    It is expected that nuclear power generation will reach 49 million kW in 1985 and 129 million kW in 1995, and the nuclear fuel having to be supplied and processed will increase in proportion to these values. The technical problems concerning nuclear fuel are presented on the basis of the balance between the benefit for human beings and the burden on the human beings. Recently, especially the downstream of nuclear fuel attracts public attention. Enriched uranium as the raw material for light water reactor fuel is almost monopolized by the U.S., and the technical information has not been published for fear of the diversion to nuclear weapons. In this paper, the present situations of uranium enrichment, fuel fabrication, transportation, reprocessing and waste disposal and the future problems are described according to the path of nuclear fuel cycle. The demand and supply of enriched uranium in Japan will be balanced up to about 1988, but afterwards, the supply must rely upon the early establishment of the domestic technology by centrifugal separation method. No problem remains in the fabrication of light water reactor fuel, but for the fabrication of mixed oxide fuel, the mechanization of the production facility and labor saving are necessary. The solution of the capital risk for the construction of the second reprocessing plant is the main problem. Japan must develop waste disposal techniques with all-out efforts.

  1. A comparison of thermal algorithms of fuel rod performance code systems

    International Nuclear Information System (INIS)

    Park, C. J.; Park, J. H.; Kang, K. H.; Ryu, H. J.; Moon, J. S.; Jeong, I. H.; Lee, C. Y.; Song, K. C.

    2003-11-01

    The goal of the fuel rod performance is to identify the robustness of a fuel rod with cladding material. Computer simulation of the fuel rod performance becomes one of important parts to designed and evaluate new nuclear fuels and claddings. To construct a computing code system for the fuel rod performance, several algorithms of the existing fuel rod performance code systems are compared and are summarized as a preliminary work. Among several code systems, FRAPCON, and FEMAXI for LWR, ELESTRES for CANDU reactor, and LIFE for fast reactor are reviewed. Thermal algorithms of the above codes are investigated including methodologies and subroutines. This work will be utilized to construct a computing code system for dry process fuel rod performance

  2. A comparison of thermal algorithms of fuel rod performance code systems

    Energy Technology Data Exchange (ETDEWEB)

    Park, C. J.; Park, J. H.; Kang, K. H.; Ryu, H. J.; Moon, J. S.; Jeong, I. H.; Lee, C. Y.; Song, K. C

    2003-11-01

    The goal of the fuel rod performance is to identify the robustness of a fuel rod with cladding material. Computer simulation of the fuel rod performance becomes one of important parts to designed and evaluate new nuclear fuels and claddings. To construct a computing code system for the fuel rod performance, several algorithms of the existing fuel rod performance code systems are compared and are summarized as a preliminary work. Among several code systems, FRAPCON, and FEMAXI for LWR, ELESTRES for CANDU reactor, and LIFE for fast reactor are reviewed. Thermal algorithms of the above codes are investigated including methodologies and subroutines. This work will be utilized to construct a computing code system for dry process fuel rod performance.

  3. Investigation of control rod worth and nuclear end of life of BWR control rods

    International Nuclear Information System (INIS)

    Magnusson, Per

    2008-01-01

    This work has investigated the Control Rod Worth (CRW) and Nuclear End of Life (NEOL) values for BWR control rods. A study of how different parameters affect NEOL was performed with the transport code PHOENIX4. It was found that NEOL, expressed in terms of 10 B depletion, can be generalized beyond the conditions for which the rod is depleted, such as different power densities and void fractions, the corresponding variation in the NEOL will be about 0.2-0.4% 10 B. It was also found that NEOL results for different fuel types and different fuel enrichments have a variation of about 2-3% in 10 B depletion. A comparative study on NHOL and CRW was made between PHOENIX4 and the stochastic Monte Carlo code MCNP. It was found that there is a significant difference, both due to differences in the codes and to limitations in the geometrical modeling in PHOENIX4. Since MCNP is considered more physically correct, a methodology was developed to calculate the nuclear end of life of BWR control rods with MCNP. The advantages of the methodology are that it does not require other codes to perform the depletion of the absorber material, it can describe control rods of any design and it can deplete the control rod absorber material without burning the fuel. The disadvantage of the method is that is it time-consuming

  4. Control rod drive of nuclear reactor

    International Nuclear Information System (INIS)

    Zhuchkov, I.I.; Gorjunov, V.S.; Zaitsev, B.I.

    1980-01-01

    This invention relates to nuclear reactors and, more particularly, to a drive of a control rod of a nuclear reactor and allows power control, excess reactivity compensation, and emergency shut-down of a reactor. (author)

  5. Nuclear fuel

    International Nuclear Information System (INIS)

    Azevedo, J.B.L. de.

    1980-01-01

    All stages of nuclear fuel cycle are analysed with respect to the present situation and future perspectives of supply and demand of services; the prices and the unitary cost estimation of these stages for the international fuel market are also mentioned. From the world resources and projections of uranium consumption, medium-and long term analyses are made of fuel availability for several strategies of use of different reactor types. Finally, the cost of nuclear fuel in the generation of electric energy is calculated to be used in the energetic planning of the electric sector. (M.A.) [pt

  6. The nuclear fuel cycle

    International Nuclear Information System (INIS)

    Patarin, L.

    2002-01-01

    This book treats of the different aspects of the industrial operations linked with the nuclear fuel, before and after its use in nuclear reactors. The basis science of this nuclear fuel cycle is chemistry. Thus a recall of the elementary notions of chemistry is given in order to understand the phenomena involved in the ore processing, in the isotope enrichment, in the fabrication of fuel pellets and rods (front-end of the cycle), in the extraction of recyclable materials (residual uranium and plutonium), and in the processing and conditioning of wastes (back-end of the fuel cycle). Nuclear reactors produce about 80% of the French electric power and the Cogema group makes 40% of its turnover at the export. Thus this book contains also some economic and geopolitical data in order to clearly position the stakes. The last part, devoted to the management of wastes, presents the solutions already operational and also the research studies in progress. (J.S.)

  7. Nuclear fuel assembly

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  8. COBRA - 3C/KFKI: a digital computer program for steady and transient thermal-hydraulic analysis of rod bundle nuclear fuel elements

    International Nuclear Information System (INIS)

    Vigassy, J.; Kovacs, L.M.

    1977-11-01

    COBRA-3C/KFKI is a digital computer program for the CDC-3300 computer in FORTRAN language. The program is a revised version of the original COBRA-3C code. The code calculates steady-state and transient flow and enthalpy transport in rod-bundle nuclear fuel elements in both boiling and nonboiling conditions. The mathematical model is formulated by dividing the bundle flow area into flow subchannels that are assumed to contain one-dimensional flow and are coupled to each other by turbulent and diversion crossflow mixing. The program neglects sonic velocity propagation but allows for a temporal and spatial acceleration of the diversion crossflow in the transverse momentum equation. A semiexplicit finite-difference scheme is used to perform a boundary-value solution where the boundary conditions are the inlet enthalpy, inlet flow rate and exit pressure. (D.P.)

  9. Generation of heat on fuel rod in cosine pattern by using induction heating

    International Nuclear Information System (INIS)

    Keettikkal, Felix; Sajeesh, Divya; Rao, Poornima; Hande, Shashank; Dakave, Ganesh; Kute, Tushar; Mahajan, Akshay; Kulkarni, R.D.

    2017-01-01

    Fuel rods are used in a nuclear reactor for fission process. When these rods are cooled by water during the heat transfer, the temperature stress causes undesirable defects in the fuel rod. Studying these defects occurring in the fuel rod in the nuclear cluster during nuclear reaction is a difficult task because fission reaction makes it difficult to analyse the changes in the rod. Hence there is a need to use a replica of the rod with similar thermal stress to study and analyse the rod for the defects. Normally the heat generated on the fuel rod follows a cosine pattern which is an inherent characteristic inside a nuclear reactor. In view of this, in this paper induction heating method is used on a rod to create an exact replica of the cosine pattern of heat by varying the pitch of the coil. First, a MATLAB simulation is done using simulink. Then a prototype of the model has been developed comprising of carbon steel pipe, with length and outside diameter of 1 meter and 48.2 mm, respectively. Instead of using water as coolant, rod is simulated in air. Therefore, the heat generated is lost by normal convection and radiation. Non-nuclear testing can be a valuable tool in the development or in some kind of experiment using nuclear reactor. Induction heating becomes an alternative to classical heating technologies because of its advantages such as efficiency, quickness, safety, clean heating and accurate power control. (author)

  10. Fuel Rod Flow-Induced Vibration Overview

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kang Hee; Kang, Heung Seok; Kim, Hyung Kyu [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-10-15

    To ensure fuel design safety and structural integrity requires the response prediction of fuel rod to reactor coolant flow excitation. However, there are many obstacles in predicting the response as described. Even if the response can be predicted, the design criteria on wear failure, including correlation with the vibration, may be difficult to establish because of a variety of related parameters, such as material, surface condition and environmental factors. Thus, a prototype test for each new fuel assembly design, i.e. a long-term endurance test, is performed for design validation with respect to flow-induced vibration (FIV) and wear. There are still needs of theoretical prediction methods for the response and anticipated failure. This paper revisits the general aspect on the response prediction, mathematical description, analysis procedure and wear correlation aspect of fuel rod's FIV

  11. Fuel rod failure detection method and system

    International Nuclear Information System (INIS)

    Assmann, H.; Janson, W.; Stehle, H.; Wahode, P.

    1975-01-01

    The inventor claims a method for the detection of a defective fuel rod cladding tube or of inleaked water in the cladding tube of a fuel rod in the fuel assembly of a pressurized-water reactor. The fuel assembly is not disassembled but examined as a whole. In the examination, the cladding tube is heated near one of its two end plugs, e.g. with an attached high-frequency inductor. The water contained in the cladding tube evaporates, and steam bubbles or a condensate are detected by the ultrasonic impulse-echo method. It is also possible to measure the delay of the temperature rise at the end plug or to determine the cooling energy required to keep the end plug temperature stable and thus to detect water ingression. (DG/AK) [de

  12. Power ramp testing method for PWR fuel rod at research reactor

    International Nuclear Information System (INIS)

    Zhou Yidong; Zhang Peisheng; Zhang Aimin; Gao Yongguang; Wang Huarong

    2003-01-01

    A tentative power ramp test for short PWR fuel rod has been conducted at the Heavy Water Research Reactor (HWRR) in China Institute of Atomic Energy (CIAE). The test fuel rod was cooled by the circulating water in the test loop. The power ramp was realized by moving solid neutron-absorbing screen around the fuel rod. The linear power of the fuel rod increased from 220 W/cm to 340 W/cm with a power ramp rate of 20 W/cm/min. The power of the fuel rod was monitored by both in-core thermal and nuclear measurement sensors in the test rig. This test provides experiences for further developing the power ramp test methods for PWR fuel rods at research reactor. (author)

  13. Stress Analysis of Single Spacer Grid Support considering Fuel Rod

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, Y. G.; Jung, D. H.; Kim, J. H. [Chungnam National University, Daejeon (Korea, Republic of); Park, J. K.; Jeon, K. L. [Korea Nuclear Fuel, Daejeon (Korea, Republic of)

    2010-10-15

    Pressurized water reactor (PWR) nuclear fuel assembly is mainly composed of a top-end piece, a bottom-end piece, lots of fuel rods, and several spacer grids. Among them, the main function of spacer grid is protecting fuel rods from Fluid Induced Vibration (FIV). The cross section of spacer grid assembled by laser welding in upper and lower point. When the fuel rod inserted in spacer gird, spring and dimple and around of welded area got a stresses. The main hypothesis of this analysis is the boundary area of HAZ and base metal can get a lot of damage than other area by FIV. So, design factors of spacer grid mainly considered to preventing the fatigue failure in HAZ and spring and dimple of spacer grid. From previous researching, the environment in reactor verified. Pressure and temperature of light water observed 15MPa and 320 .deg. C, and vibration of the fuel rod observed within 0 {approx} 50Hz. In this study, mechanical properties of zirconium alloy that extracted from the test and the spacer grid model which used in the PWR were applied in stress analyzing. General-purpose finite element analysis program was used ANSYS Workbench 12.0.1 version. 3-D CAD program CATIA was used to create spacer grid model

  14. Process development and fabrication for sphere-pac fuel rods

    International Nuclear Information System (INIS)

    Welty, R.K.; Campbell, M.H.

    1981-06-01

    Uranium fuel rods containing sphere-pac fuel have been fabricated for in-reactor tests and demonstrations. A process for the development, qualification, and fabrication of acceptable sphere-pac fuel rods is described. Special equipment to control fuel contamination with moisture or air and the equipment layout needed for rod fabrication is described and tests for assuring the uniformity of the fuel column are discussed. Fuel retainers required for sphere-pac fuel column stability and instrumentation to measure fuel column smear density are described. Results of sphere-pac fuel rod fabrication campaigns are reviewed and recommended improvements for high throughput production are noted

  15. Apparatus and method for applying an end plug to a fuel rod tube end

    International Nuclear Information System (INIS)

    Rieben, S.L.; Wylie, M.E.

    1987-01-01

    An apparatus is described for applying an end plug to a hollow end of a nuclear fuel rod tube, comprising: support means mounted for reciprocal movement between remote and adjacent positions relative to a nuclear fuel rod tube end to which an end plug is to be applied; guide means supported on the support means for movement; and drive means coupled to the support means and being actuatable for movement between retracted and extended positions for reciprocally moving the support means between its respective remote and adjacent positions. A method for applying an end plug to a hollow end of a nuclear fuel rod tube is also described

  16. Fuel rod behaviour during transients

    International Nuclear Information System (INIS)

    Hughes, H.; Haste, T.J.; Cameron, R.F.; Sinclair, J.E.

    1982-04-01

    The fuel pin performance code SLEUTH, the transient codes FRAP-T5 and TRAFIC and the clad deformation code CANSWEL-2 are described. It is shown how the codes treat gas release, pin cooling, cladding deformation and interaction, gap conductance etc. The materials properties used are indicated. (author)

  17. Crud deposition modeling on BWR fuel rods

    International Nuclear Information System (INIS)

    Kucuk, Aylin; Cheng, Bo; Potts, Gerald A.; Shiralkar, Bharat; Morgan, Dave; Epperson, Kenny; Gose, Garry

    2014-01-01

    Deposition of boiling water reactor (BWR) system corrosion products (crud) on operating fuel rods has resulted in performance-limiting conditions in a number of plants. The operational impact of performance-limiting conditions involving crud deposition can be detrimental to a BWR operator, resulting in unplanned or increased frequency of fuel inspections, fuel failure and associated radiological consequences, operational restrictions including core power derate and/or forced shutdowns to remove failed fuel, premature discharge of individual bundles or entire reloads, and/or undesirable core design restrictions. To facilitate improved management of crud-related fuel performance risks, EPRI has developed the CORAL (Crud DepOsition Risk Assessment ModeL) tool. This paper presents a summary of the CORAL elements and benchmarking results. Applications of CORAL as a tool for fuel performance risk assessment are also discussed. (author)

  18. Fuel and control rod failure behavior during degraded core accidents

    International Nuclear Information System (INIS)

    Chung, K.S.

    1984-01-01

    As a part of the pretest and posttest analyses of Light Water Reactor Source Term Experiments (STEP) which are conducted in the Transient Reactor Test (TREAT) facility, this paper investigates the thermodynamic and material behaviors of nuclear fuel pins and control rods during severe core degradation accidents. A series of four STEP tests are being performed to simulate the characteristics of the power reactor accidents and investigate the behavior of fission product release during these accidents. To determine the release rate of the fission products from the fuel pins and the control rod materials, information concerning the timing of the clad failure and the thermodynamic conditions of the fuel pins and control rods are needed to be evaluated. Because the phase change involves a large latent heat and volume expansion, and the phase change is a direct cause of the clad failure, the understanding of the phase change phenomena, particularly information regarding how much of the fuel pin and control rod materials are melted are very important. A simple energy balance model is developed to calculate the temperature profile and melt front in various heat transfer media considering the effects of natural convection phenomena on the melting and freezing front behavior

  19. Linear variable differential transformer and its uses for in-core fuel rod behavior measurements

    International Nuclear Information System (INIS)

    Wolf, J.R.

    1979-01-01

    The linear variable differential transformer (LVDT) is an electromechanical transducer which produces an ac voltage proportional to the displacement of a movable ferromagnetic core. When the core is connected to the cladding of a nuclear fuel rod, it is capable of producing extremely accurate measurements of fuel rod elongation caused by thermal expansion. The LVDT is used in the Thermal Fuels Behavior Program at the U.S. Idaho National Engineering Laboratory (INEL) for measurements of nuclear fuel rod elongation and as an indication of critical heat flux and the occurrence of departure from nucleate boiling. These types of measurements provide important information about the behavior of nuclear fuel rods under normal and abnormal operating conditions. The objective of the paper is to provide a complete account of recent advances made in LVDT design and experimental data from in-core nuclear reactor tests which use the LVDT

  20. Fuel rod design by statistical methods for MOX fuel

    International Nuclear Information System (INIS)

    Heins, L.; Landskron, H.

    2000-01-01

    Statistical methods in fuel rod design have received more and more attention during the last years. One of different possible ways to use statistical methods in fuel rod design can be described as follows: Monte Carlo calculations are performed using the fuel rod code CARO. For each run with CARO, the set of input data is modified: parameters describing the design of the fuel rod (geometrical data, density etc.) and modeling parameters are randomly selected according to their individual distributions. Power histories are varied systematically in a way that each power history of the relevant core management calculation is represented in the Monte Carlo calculations with equal frequency. The frequency distributions of the results as rod internal pressure and cladding strain which are generated by the Monte Carlo calculation are evaluated and compared with the design criteria. Up to now, this methodology has been applied to licensing calculations for PWRs and BWRs, UO 2 and MOX fuel, in 3 countries. Especially for the insertion of MOX fuel resulting in power histories with relatively high linear heat generation rates at higher burnup, the statistical methodology is an appropriate approach to demonstrate the compliance of licensing requirements. (author)

  1. Relation of fuel rod service parameters and design requirements to produced fuel rod and their components

    International Nuclear Information System (INIS)

    Bibilashvili, Yu.K.

    1999-01-01

    Based on the presented material it is possible to state that there is a very close link between the fuel operational parameters and the requirements for its design and production process. The required performance and life-time of a fuel rod can be only assured by the correctly selected design and process solutions. The economical aspect of this problem is significantly depend on the commercial feasibility of a particular selected solution with the provision of an automated and comparative by inexpensive production of a fuel rod and its components. The operational conditions are also important for the life time of the fuel rods. If there are no special measures for the mitigation of the certain operation conditions the leakage of fuel elements can take place before the planned time. (authors)

  2. Elliptical cross section fuel rod study II

    International Nuclear Information System (INIS)

    Taboada, H.; Marajofsky, A.

    1996-01-01

    In this paper it is continued the behavior analysis and comparison between cylindrical fuel rods of circular and elliptical cross sections. Taking into account the accepted models in the literature, the fission gas swelling and release were studied. An analytical comparison between both kinds of rod reveals a sensible gas release reduction in the elliptical case, a 50% swelling reduction due to intragranular bubble coalescence mechanism and an important swelling increase due to migration bubble mechanism. From the safety operation point of view, for the same linear power, an elliptical cross section rod is favored by lower central temperatures, lower gas release rates, greater gas store in ceramic matrix and lower stored energy rates. (author). 6 refs., 8 figs., 1 tab

  3. Zircaloy sheathed thermocouples for PWR fuel rod temperature measurements

    International Nuclear Information System (INIS)

    Anderson, J.V.; Wesley, R.D.; Wilkins, S.C.

    1979-01-01

    Small diameter zircaloy sheathed thermocouples have been developed by EG and G Idaho, Inc., at the Idaho National Engineering Laboratory. Surface mounted thermocouples were developed to measure the temperature of zircaloy clad fuel rods used in the Thermal Fuels Behavior Program (TFBP), and embedded thermocouples were developed for use by the Loss-of-Fluid Test (LOFT) Program for support tests using zircaloy clad electrically heated nuclear fuel rod simulators. The first objective of this developmental effort was to produce zircaloy sheathed thermocouples to replace titanium sheathed thermocouples and thereby eliminate the long-term corrosion of the titanium-to-zircaloy attachment weld. The second objective was to reduce the sheath diameter to obtain faster thermal response and minimize cladding temperature disturbance due to thermocouple attachment

  4. Simulation of the fuel rod thermal hydraulic performance during the blow down phase in a PWR

    International Nuclear Information System (INIS)

    Gadelha, J.A.M.

    1982-10-01

    A digital computer code to predict the fuel rod thermalhydraulic performance during a postulated loss-of-coolant accident (LOCA) in the primary circuit of a PWR nuclear power plant is developed. The fuel rod corresponds to that in an average channel in the core. Only the blowdown phase is considered during the accident. The conservation equations of mass, momentum, and energy, and the heat conduction equation are solved to determine the fuel rod conditions during the accident. Finite differences are applied as a numerical method in the solution of the equations modelling the rod and coolant conditions. (Author) [pt

  5. On-line fuel and control rod integrity management in BWRs

    International Nuclear Information System (INIS)

    Larsson, Irina; Sihver, Lembit

    2011-01-01

    Surveillance of fuel and control rod integrity in a BWR core is essential to maintain a safe and reliable operation of a nuclear power plant. An accurate and prompt way to monitor fuel integrity in a reactor core during reactor operation is by using on-line measurements of the gamma emitting noble gas activities in the off-gas system. The integrity of control rods can be efficiently followed by on-line measurements of the helium (He) concentration in the off-gases. This method also gives information about fuel rod failures since He is used as a fill gas in the fuel rods. To survey fuel and control rod integrity during reactor operation, a system consisting of combined gamma and He on-line measurements in the off-gases should be used. Such a system can detect and follow the behavior of fuel and control rod failures. In addition, it can separate fuel failures from control rod failures since fuel rods contain both He and gamma emitting noble gases, while control rods only contain He. Moreover, the system is able to distinguish primary fuel failures from degradation of already existing ones. In this paper we present a combined system for on-line measurements of He and gamma emitting noble gases in the reactor off-gas system and measuring experiences from different BWRs. (author)

  6. Nuclear fuel element end fitting

    International Nuclear Information System (INIS)

    Jabsen, F.S.

    1979-01-01

    A typical embodiment of the invention has an array of sockets that are welded to the intersections of the plates that form the upper and lower end fittings of a nuclear reactor fuel element. The sockets, which are generally cylindrical in shape, are oriented in directions that enable the longitudinal axes of the sockets to align with the longitudinal axes of the fuel rods that are received in the respective sockets. Detents impressed in the surfaces of the sockets engage mating grooves that are formed in the ends of the fuel rods to provide for the structural integrity of the fuel element

  7. Model investigation of fuel rod behaviour

    International Nuclear Information System (INIS)

    Girgis, M.M.; Wiesenack, W.; Stegemann, D.

    1985-06-01

    Thermal and mechanical behaviour of fuel rods can be explained but unsatisfactorily by models based of an axial symmetry concept. Recently developed models include, with respect to their thermal components, a simple method for the computation of the temperature distribution within the fuel, and they also take into account the influence of excentrically placed pellets for the computation of heat transfer in the cold gap. Additionally, a finite-element model is used to evaluate the effects of cracking and fragmentation on the thermal behaviour of pellets. The reaction of fuel and fuel cladding to external and internal loadings and the axial interaction between fuel and cladding are described in the mechanical portion of the model. A special case of axial coupling is the so-called random stacking interaction caused by fuel pellets placed excentrically at the cladding and sliding radially and axially. In the comparison of measurement results, both thermal and mechanical behaviour of different rods from the OECD Halden Reactor Project are subject to investigations. (RF) [de

  8. Fuel rod behaviour during transients

    International Nuclear Information System (INIS)

    Bilsby, C.F.; Haste, T.J.; Garlick, A.; Cameron, R.F.

    1982-04-01

    The clad deformation code CANSWELL-2 is described. This is used, either as a stand-alone code or within MABEL-2, to predict and analyse the results of LOCA simulations in the Halden and NRU reactors and in the KfK and PROPAT rigs. Experimental evidence on fuel behaviour in RIA, PCM and ATWS events is presented with inclusion of certain FRAP-T5 results. Published calculations from the accident codes FRAP-T4 and FRAP-T5 are compared with experimental results in simulated loss of coolant tests in the Power Burst Facility. The limitations of this code in its treatment of RIA, PCM and ATWS events are considered. (U.K.)

  9. Structural analysis and modeling of water reactor fuel rod behavior

    International Nuclear Information System (INIS)

    Roshan Zamir, M.

    2000-01-01

    An important aspect of the design and analysis of nuclear reactor is the ability to predict the behavior of fuel elements in the adverse environment of a reactor system under normal and emergency operating conditions. To achieve these objectives and in order to provide a suitable computer code based on fundamental material properties for design and study of the thermal-mechanical behavior of water reactor fuel rods during their irradiation life and also to demonstrate the fuel rod design and modeling for students, The KIANA-1 computer program has been developed by the writer at Amir-Kabir university of technology with support of Atomic Energy Organization of Iran. KIANA-1 is an integral one-dimensional computer program for the thermal and mechanical analysis in order to predict fuel rods performance and also parameter study of Zircaloy-clad UO 2 fuel rod during steady state conditions. The code has been designed for the following main objectives: To give a solution for the steady state heat conduction equation for fuel as a heat source and clad by using finite difference, control volume and semi-analytical methods in order to predict the temperature profile in the fuel and cladding. To predict the inner gas pressures due to the filling gases and released gaseous fission products. To predict the fission gas production and release by using a simple diffusion model based on the Booth models and an empirical model. To calculate the fuel-clad gap conductance for cracked fuel with partial contact zones to a closed gap with strong contact. To predict the distribution of stress in three principal directions in the fuel and sheet by assuming one-dimensional plane strain and asymmetric idealization. To calculate the strain distribution in three principal directions and the corresponding deformation in the fuel and cladding. For this purpose the permanent strain such as creep or plasticity as well as the thermoelastic deformation and also the swelling, densification, cracking

  10. Optimization of in-core fuel management and control rod strategy in equilibrium fuel cycle

    International Nuclear Information System (INIS)

    Sekimizu, Koichi

    1975-01-01

    An in-core fuel management problem is formulated for the equilibrium fuel cycle in an N-region nuclear reactor model. The formulation shows that the infinite multiplication factor k infinity requisite for newly charged fuel can be separated into two terms - one corresponding to the average k infinity at the end of the cycle and the other representing the direct contribution of the shuffling scheme and control rod programming. This formulation is applied to a three-region cylindrical reactor to obtain simultaneous optimization of shuffling and control rod programming. It is demonstrated that this formulation aids greatly in gaining a better understanding of the effects of changes in the shuffling scheme and control rod programming on equilibrium fuel cycle performance. (auth.)

  11. Experimental design for HTGR fuel rods

    International Nuclear Information System (INIS)

    Bayne, C.K.

    1975-01-01

    Fuel rods for the high temperature gas cooled reactor are composed of pyrolytic carbon coated fuel particles bounded by a carbonaceous matrix. Because of differential shrinkage between coated particles and the carbonaceous matrix, breakage of the pyrolytic coating has been observed with certain combinations of coated particles and matrix compositions. The pyrolytic coating is intended to be the primary containment for fission products. Therefore, an experiment is desired to determine the breakage characteristics of different strength coated particles combined with different matrix compositions during irradiation

  12. Fuel assembly and burnable poison rod

    International Nuclear Information System (INIS)

    Hirukawa, Koji.

    1993-01-01

    In a fuel assembly having burnable poison rods arranged therein, the burnable poison comprises an elongate small outer tube and an inner tube coaxially disposed within the outer tube. Upper and lower end tubes each sealed at one end are connected to both of the upper and lower ends in the inner and the outer tubes respectively. A coolant inlet hole is disposed to the lower end tube, while a coolant leakage hole is disposed to the upper end tube. Burnable poison members are filled in an annular space. Further, the burnable poison-filling region is disposed excepting portions for 1/20 - 1/12 of the effective fuel length at each of the upper and the lower ends of the fuel rod. Then, the concentration of the burnable poisons in a region above a boundary defined at a position 1/3 - 1/2, from beneath, of the effective fuel length is made smaller than that in the lower region. This enables to suppress excess reactions of fuels to reduce the mass of the burnable neutron. Excellent reactivity control performance at the initial stage of the burning can be attained. (T.M.)

  13. Strategy for Fuel Rod Receipt, Characterization, Sample Allocation for the Demonstration Sister Rods

    Energy Technology Data Exchange (ETDEWEB)

    Marschman, Steven C. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Warmann, Stephan A. [Portage, Inc., Idaho Falls, ID (United States); Rusch, Chris [NAC International, Inc., Norcross, GA (United States)

    2014-03-01

    The U.S. Department of Energy Office of Nuclear Energy (DOE-NE), Office of Fuel Cycle Technology, has established the Used Fuel Disposition Campaign (UFDC) to conduct the research and development activities related to storage, transportation, and disposal of used nuclear fuel and high-level radioactive waste. The mission of the UFDC is to identify alternatives and conduct scientific research and technology development to enable storage, transportation and disposal of used nuclear fuel (UNF) and wastes generated by existing and future nuclear fuel cycles. The UFDC Storage and Transportation staffs are responsible for addressing issues regarding the extended or long-term storage of UNF and its subsequent transportation. The near-term objectives of the Storage and Transportation task are to use a science-based approach to develop the technical bases to support the continued safe and secure storage of UNF for extended periods, subsequent retrieval, and transportation. While low burnup fuel [that characterized as having a burnup of less than 45 gigawatt days per metric tonne uranium (GWD/MTU)] has been stored for nearly three decades, the storage of high burnup used fuels is more recent. The DOE has funded a demonstration project to confirm the behavior of used high burnup fuel under prototypic conditions. The Electric Power Research Institute (EPRI) is leading a project team to develop and implement the Test Plan to collect this data from a UNF dry storage system containing high burnup fuel. The Draft Test Plan for the demonstration outlines the data to be collected; the high burnup fuel to be included; the technical data gaps the data will address; and the storage system design, procedures, and licensing necessary to implement the Test Plan. To provide data that is most relevant to high burnup fuel in dry storage, the design of the test storage system must closely mimic real conditions high burnup SNF experiences during all stages of dry storage: loading, cask drying

  14. Ultrasonics aids the identification of failed fuel rods

    International Nuclear Information System (INIS)

    Anon.

    1985-01-01

    Over a number of years Brown Boveri Reaktor of West Germany has developed and commercialized an ultrasonic failed fuel rod detection system. Sipping has up to now been the standard technique for failed fuel detection, but sipping can only indicate whether or not an assembly contains defective rods; the BBR system can tell which rod is defective. (author)

  15. Quality management of nuclear fuel

    International Nuclear Information System (INIS)

    2006-01-01

    The Guide presents the quality management requirements to be complied with in the procurement, design, manufacture, transport, receipt, storage, handling and operation of nuclear fuel. The Guide also applies to control rods and shield elements to be placed in the reactor. The Guide is mainly aimed for the licensee responsible for the procurement and operation of fuel, for the fuel designer and manufacturer and for other organisations, whose activities affect fuel quality and the safety of fuel transport, storage and operation. General requirements for nuclear fuel are presented in Section 114 of the Finnish Nuclear Energy Decree and in Section 15 of the Government Decision (395/1991). Regulatory control of the safety of fuel is described in Guides YVL6.1, YVL6.2 and YVL6.3. An overview of the regulatory control of nuclear power plants carried out by STUK (Radiation and Nuclear Safety Authority, Finland) is clarified in Guide YVL1.1

  16. Study of pellet clad interaction defects in Dresden-3 fuel rods

    International Nuclear Information System (INIS)

    Pasupathi, V.; Perrin, J.S.

    1979-01-01

    During Cycle-3 operation of Dresden-3, fuel rod failures occurred following a transient power increase. Ten fuel rods from five of the leaking fuel assemblies were examined at Battelle's Columbus Laboratory and General Electric-Vallecitos Nuclear Center. Examinations consisted of nondestructive and destructive methods including metallography and scanning electron microscopy (SEM). Results showed the cause of fuel rod failure to be pellet clad interaction involving stress corrosion cracking. Results of SEM studies of the cladding crack surfaces and deposits on clad inner surfaces were in agreement with those reported by other investigators

  17. 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)

  18. Licensing of spent fuel dry storage and consolidated rod storage

    International Nuclear Information System (INIS)

    Bailey, W.J.

    1990-02-01

    The results of this study, performed by Pacific Northwest Laboratory (PNL) and sponsored by the US Department of Energy (DOE), respond to the nuclear industry's recommendation that a report be prepared that collects and describes the licensing issues (and their resolutions) that confront a new applicant requesting approval from the US Nuclear Regulatory Commission (NRC) for dry storage of spent fuel or for large-scale storage of consolidated spent fuel rods in pools. The issues are identified in comments, questions, and requests from the NRC during its review of applicants' submittals. Included in the report are discussions of (1) the 18 topical reports on cask and module designs for dry storage fuel that have been submitted to the NRC, (2) the three license applications for dry storage of spent fuel at independent spent fuel storage installations (ISFSIs) that have been submitted to the NRC, and (3) the three applications (one of which was later withdrawn) for large-scale storage of consolidated fuel rods in existing spent fuel storage pools at reactors that were submitted tot he NRC. For each of the applications submitted, examples of some of the issues (and suggestions for their resolutions) are described. The issues and their resolutions are also covered in detail in an example in each of the three subject areas: (1) the application for the CASTOR V/21 dry spent fuel storage cask, (2) the application for the ISFSI for dry storage of spent fuel at Surry, and (3) the application for full-scale wet storage of consolidated spent fuel at Millstone-2. The conclusions in the report include examples of major issues that applicants have encountered. Recommendations for future applicants to follow are listed. 401 refs., 26 tabs

  19. LOFT fuel rod surface temperature measurement testing

    International Nuclear Information System (INIS)

    Eaton, A.M.; Tolman, E.L.; Solbrig, C.W.

    1978-01-01

    Testing of the LOFT fuel rod cladding surface thermocouples has been performed to evaluate how accurately the LOFT thermocouples measure the cladding surface temperature during a loss-of-coolant accident (LOCA) sequence and what effect, if any, the thermocouple would have on core performance. Extensive testing has been done to characterize the thermocouple design. Thermal cycling and corrosion testing of the thermocouple weld design have provided an expected lifetime of 6000 hours when exposed to reactor coolant conditions of 620 K and 15.9 MPa and to sixteen thermal cycles with an initial temperature of 480 K and peak temperatures ranging from 870 to 1200K. Departure from nucleate boiling (DNB) tests have indicated a DNB penalty (5 to 28% lower) during steady state operation and negligible effects during LOCA blowdown caused by the LOFT fuel rod surface thermocouple arrangement. Experience with the thermocouple design in Power Burst Facility (PBF) and LOFT nonnuclear blowdown testing has been quite satisfactory. Tests discussed here were conducted using both stainless steel and zircaloy-clad electrically heated rod in the LOFT Test Support Facility (LTSF) blowdown simulation loop

  20. The Defect Inspection on the Irradiated Fuel Rod by Eddy Current Test

    International Nuclear Information System (INIS)

    Koo, D. S.; Park, Y. K.; Kim, E. K.

    1996-01-01

    The eddy current test(ECT) probe of differential encircling coil type was designed and fabricated, and the optimum condition of ECT was derived for the examination of the irradiated fuel rod. The correlation between ECT test frequency and phase and amplitude was derived by performing the test of the standard rig that includes inner notches, outer notches and through-holes. The defect of through-hole was predicted by ECT at the G33-N2 fuel rod irradiated in the Kori-1 nuclear power reactor. The metallographic examination on the G33-N2 fuel rod was Performed at the defect location predicted by ECT. The result of metallographic examination for the G33-N2 fuel rod was in good agreement with that of ECT. This proves that the evaluation for integrity of irradiated fuel rod by ECT is reliable

  1. Optical coherence tomography for nondestructive evaluation of fuel rod degradation

    International Nuclear Information System (INIS)

    Renshaw, Jeremy B.; Jenkins, Thomas P.; Buckner, Benjamin D.; Friend, Brian

    2015-01-01

    Nuclear power plants regularly inspect fuel rods to ensure safe and reliable operation. Excessive corrosion can cause fuel failures which can have significant repercussions for the plant, including impacts on plant operation, worker exposure to radiation, and the plant's INPO rating. While plants typically inspect for fuel rod corrosion using eddy current techniques, these techniques have known issues with reliability in the presence of tenacious, ferromagnetic crud layers that can deposit during operation, and the nondestructive evaluation (NDE) inspection results can often be in error by a factor of 2 or 3. For this reason, alternative measurement techniques, such as Optical Coherence Tomography (OCT), have been evaluated that are not sensitive to the ferromagnetic nature of the crud. This paper demonstrates that OCT has significant potential to characterize the thickness of crud layers that can deposit on the surfaces of fuel rods during operation. Physical trials have been performed on simulated crud samples, and the resulting data show an apparent correlation between the crud layer thickness and the OCT signal

  2. Optical coherence tomography for nondestructive evaluation of fuel rod degradation

    Energy Technology Data Exchange (ETDEWEB)

    Renshaw, Jeremy B., E-mail: jrenshaw@epri.com [Electric Power Research Institute, 1300 West WT Harris Blvd., Charlotte, NC 28262 (United States); Jenkins, Thomas P., E-mail: tjenkins@metrolaserinc.com; Buckner, Benjamin D., E-mail: tjenkins@metrolaserinc.com [MetroLaser, Inc., 22941 Mill Creek Drive, Laguna Hills, CA 92653 (United States); Friend, Brian [AREVA, Inc., 3315 Old Forest Road, Lynchburg, VA 24501 (United States)

    2015-03-31

    Nuclear power plants regularly inspect fuel rods to ensure safe and reliable operation. Excessive corrosion can cause fuel failures which can have significant repercussions for the plant, including impacts on plant operation, worker exposure to radiation, and the plant's INPO rating. While plants typically inspect for fuel rod corrosion using eddy current techniques, these techniques have known issues with reliability in the presence of tenacious, ferromagnetic crud layers that can deposit during operation, and the nondestructive evaluation (NDE) inspection results can often be in error by a factor of 2 or 3. For this reason, alternative measurement techniques, such as Optical Coherence Tomography (OCT), have been evaluated that are not sensitive to the ferromagnetic nature of the crud. This paper demonstrates that OCT has significant potential to characterize the thickness of crud layers that can deposit on the surfaces of fuel rods during operation. Physical trials have been performed on simulated crud samples, and the resulting data show an apparent correlation between the crud layer thickness and the OCT signal.

  3. 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

  4. Conditioning of nuclear reactor fuel

    International Nuclear Information System (INIS)

    1975-01-01

    A method of conditioning the fuel of a nuclear reactor core to minimize failure of the fuel cladding comprising increasing the fuel rod power to a desired maximum power level at a rate below a critical rate which would cause cladding damage is given. Such conditioning allows subsequent freedom of power changes below and up to said maximum power level with minimized danger of cladding damage. (Auth.)

  5. Results from In-pile experiments on LWR fuel rod behavior under LOCA conditions with unirradiated rods

    International Nuclear Information System (INIS)

    Sepold, L.; Karb, E.H.; Pruessmann, M.

    1981-06-01

    This report summarizes the results of the FR2-in-pile tests at KfK (Kernforschungszentrum Karlsruhe) with unirradiated test rods. The in-pile tests with the objective of investigating the influence of a nuclear environment on the mechanisms of fuel rod failure were being performed with irradiated and unirradiated single rods of a PWR design in the DK loop of the FR2 reactor. The main parameter of the test program was the burnup, ranging from 2.500 to 35.000 MWd/t. The program with unirradiated specimens comprised the series A and B with a total of 14 tests. (orig.) [de

  6. An Evaluation on the Fluid Elastic Instability of the Fuel Rod for OPR1000 Plants

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyeong Koo; Jeon, Sang Yoon; Lee, Kyu Seok; Kim, Jeong Ha; Lee, Sang Jong [Reactor Core Technology Department, Korea Nuclear Fuel, 493, Deogjin, Yuseong, Daejeon, 305-353 (Korea, Republic of)

    2009-06-15

    The fuel assembly for a typical PWR (Pressurized Water Reactor) plant suffers severe operating conditions during its lifetime such as high temperature, high pressure and massive coolant passing through the fuel assembly with high speed. Moreover, recently nuclear fuel is requested not only to operate under more severe operation conditions for example high burnup, longer cycle and power up-rate, but also to maintain its integrity in spite of the operation severity. Lots of vendors, therefore, have poured their endeavor to develop an advanced fuel in order to meet these requirements. However, the fuel failures are still reported from time to time. In general, fuel failure mechanisms known as significant causes of PWR fuel failure are grid to rod fretting, corrosion of the cladding, pellet cladding interaction and debris induced fretting. Especially, since the fuel assembly is very tall and flexible structure and the flow velocity of reactor coolant is pretty high, flow induced vibration (FIV) of fuel rod is an inevitable phenomenon in PWR fuel and the energy vibrating fuel rod continually provided by coolant flow can become a root cause of the fuel failure like grid to rod fretting. Moreover, the cross flow of the coolant is highly susceptible to cause the fluid elastic instability (FEI) which produces extraordinarily big amplitudes of the fuel rod suddenly and is eventually ended up fuel failure within very short-term. The FIV problem, therefore, has to be evaluated carefully to avoid unexpected fuel failure. At present, the susceptibility to vibration damage of the fuel rod for OPR1000 plants has been estimated by the comparison of natural frequencies of every fuel rod span with recognized external excitation frequencies like coolant pump blade passing frequencies, vortex shedding frequencies and lower support structure vibration frequencies. That is, in order to prevent fuel failure due to the external excitation, the natural frequencies of unsupported lengths of

  7. PWR fuel rod corrosion in Japan

    International Nuclear Information System (INIS)

    Inoue, S.; Mori, K.; Murata, K.; Kobasyashi, S.

    1997-01-01

    Many particular appearance were observed on the fuel rod surfaces during fuel inspection at reactor outage in 1991. The appearances looked like small black circular nodules. The size was approximately 1 mm. This kind of appearances were found on fuel rods of which burnup exceeded approximately 30 GWd/t and at the second or third spans of the fuel assembly from the top. In order to clarify the cause, PIE was performed. The black nodules were confirmed to be oxide film spalling by visual inspection. Maximum oxide film thickness was 70 μm and spalling was observed where oxide thickness exceeded 40 t0 50 μm. Oxide film thickness was greater than expected. Many small pores were found in the oxide film when the oxide film had become thicker. Many circumferential cracks were also found in the film. It was speculated that these cracks caused the spalling of the oxide film. Hydride precipitates were mainly oriented circumferentially. Dense hydrides were observed near the outer rim of the cladding. No concentrated hydrides were observed near the spalling area. Maximum hydrogen content was 315 ppm. It was confirmed that the results of tensile test showed no significant effects by corrosion. The mechanism of accelerated corrosion was studied in detail. Water chemistry during irradiation was examined. Lithium content was maintained below 2.2 ppm. pH value was kept between 6.9 and 7.2. There was no anomalies in water chemistry during reactor operation. Cladding fabrication record clarified that heat treatment parameter was smaller than the optimum value. In Japan, heat treatment of the cladding was already optimized by improved fabrication process. Also chemical composition optimization of the cladding, such as low Tin and high Silicon content, was adopted for high burnup fuel. These remedies has already reduced fuel cladding corrosion and we believe we have solved this problem. (author). 6 figs, 1 tab

  8. PWR fuel rod corrosion in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Inoue, S [Kansai Electric Power Co., Inc., Osaka (Japan); Mori, K; Murata, K; Kobasyashi, S [Nuclear Fuel Industries, Ltd, Osaka (Japan)

    1997-02-01

    Many particular appearance were observed on the fuel rod surfaces during fuel inspection at reactor outage in 1991. The appearances looked like small black circular nodules. The size was approximately 1 mm. This kind of appearances were found on fuel rods of which burnup exceeded approximately 30 GWd/t and at the second or third spans of the fuel assembly from the top. In order to clarify the cause, PIE was performed. The black nodules were confirmed to be oxide film spalling by visual inspection. Maximum oxide film thickness was 70 {mu}m and spalling was observed where oxide thickness exceeded 40 t0 50 {mu}m. Oxide film thickness was greater than expected. Many small pores were found in the oxide film when the oxide film had become thicker. Many circumferential cracks were also found in the film. It was speculated that these cracks caused the spalling of the oxide film. Hydride precipitates were mainly oriented circumferentially. Dense hydrides were observed near the outer rim of the cladding. No concentrated hydrides were observed near the spalling area. Maximum hydrogen content was 315 ppm. It was confirmed that the results of tensile test showed no significant effects by corrosion. The mechanism of accelerated corrosion was studied in detail. Water chemistry during irradiation was examined. Lithium content was maintained below 2.2 ppm. pH value was kept between 6.9 and 7.2. There was no anomalies in water chemistry during reactor operation. Cladding fabrication record clarified that heat treatment parameter was smaller than the optimum value. In Japan, heat treatment of the cladding was already optimized by improved fabrication process. Also chemical composition optimization of the cladding, such as low Tin and high Silicon content, was adopted for high burnup fuel. These remedies has already reduced fuel cladding corrosion and we believe we have solved this problem. (author). 6 figs, 1 tab.

  9. Fuel assembly for a nuclear reactor

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  10. 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)

  11. Dispersion fuel for nuclear research facilities

    International Nuclear Information System (INIS)

    Kushtym, A.V.; Belash, M.M.; Zigunov, V.V.; Slabospitska, O.O.; Zuyok, V.A.

    2017-01-01

    Designs and process flow sheets for production of nuclear fuel rod elements and assemblies TVS-XD with dispersion composition UO_2+Al are presented. The results of fuel rod thermal calculation applied to Kharkiv subcritical assembly and Kyiv research reactor VVR-M, comparative characteristics of these fuel elements, the results of metallographic analyses and corrosion tests of fuel pellets are given in this paper

  12. Rod consolidation of RG and E's [Rochester Gas and Electric Corporation] spent PWR [pressurized water reactor] fuel

    International Nuclear Information System (INIS)

    Bailey, W.J.

    1987-05-01

    The rod consolidation demonstration involved pulling the fuel rods from five fuel assemblies from Unit 1 of RG and E's R.E. Ginna Nuclear Power Plant. Slow and careful rod pulling efforts were used for the first and second fuel assemblies. Rod pulling then proceeded smoothly and rapidly after some minor modifications were made to the UST and D consolidation equipment. The compaction ratios attained ranged from 1.85 to 2.00 (rods with collapsed cladding were replaced by dummy rods in one fuel assembly to demonstrate the 2:1 compaction ratio capability). This demonstration involved 895 PWR fuel rods, among which there were some known defective rods (over 50 had collapsed cladding); no rods were broken or dropped during the demonstration. However, one of the rods with collapsed cladding unexplainably broke during handling operations (i.e., reconfiguration in the failed fuel canister), subsequent to the rod consolidation demonstration. The broken rod created no facility problems; the pieces were encapsulated for subsequent storage. Another broken rod was found during postdemonstration cutting operations on the nonfuel-bearing structural components from the five assemblies; evidence indicates it was broken prior to any rod consolidation operations. During the demonstration, burnish-type lines or scratches were visible on the rods that were pulled; however, experience indicates that such lines are generally produced when rods are pulled (or pushed) through the spacer grids. Rods with collapsed cladding would not enter the funnel (the transition device between the fuel assembly and the canister that aids in obtaining high compaction ratios). Reforming of the flattened areas of the cladding on those rods was attempted to make the rod cross sections more nearly circular; some of the reformed rods passed through the funnel and into the canister

  13. HCPWR type fuel elements design with respect to its control rods

    International Nuclear Information System (INIS)

    Abbate, M.J.; Sbaffoni, M.M.; Patino, N.E.; Torasso, O.

    1992-01-01

    The high conversion reactors (HCPWR) can improve the nuclear fuel utilization. One of its present problems is the optimization of the control rods' worth and its relationship with the void coefficient. This investigation means, starting from one reference's design, to analyze several possibilities on the number and distribution of the control rods. As main result, one design including 24 control rods in an optimized distribution, is recommended. (author)

  14. Fuel rod puncturing and fission gas monitoring system examination techniques

    International Nuclear Information System (INIS)

    Song, Woong Sup

    1999-02-01

    Fission gas products accumulated in irradiated fuel rod is 1-2 cm 3 in CANDU and 40-50 cm 3 in PWR fuel rod. Fuel rod puncturing and fission gas monitoring system can be used for both CANDU and PWR fuel rod. This system comprises puncturing device located at in cell part and monitoring device located at out cell part. The system has computerized 9 modes and can calculate both void volume and mass volume only single puncturing. This report describes techniques and procedure for operating fuel rod puncturing and gas monitoring system which can be play an important role in successful operation of the devices. Results obtained from the analysis can give more influence over design for fuel rods. (Author). 6 refs., 9 figs

  15. Methodology of fuel rod design for pressurized light water reactors

    International Nuclear Information System (INIS)

    Teixeira e Silva, A.; Esteves, A.M.

    1988-01-01

    The fuel performance program FRAPCON-1 and the structural finite element program SAP-IV are applied in a pressurized water reactor fuel rod design methodology. The applied calculation procedure allows to dimension the fuel rod components and characterize its internal pressure. (author) [pt

  16. Potential impacts of crud deposits on fuel rod behaviour on high powered PWR fuel rods

    International Nuclear Information System (INIS)

    Wilson, W.; Comstock, R.J.

    1999-01-01

    Fuel assemblies operating with significant sub-cooled boiling are subject to deposition of surface deposits commonly referred to as crud. This crud can potentially cause concentration of chemical species within the deposits which can be detrimental to cladding performance in PWRs. In addition, these deposits on the surface of the cladding can result in power anomalies and erroneous reporting of fuel rod oxide thickness which can substantially hamper corrosion and core performance modeling efforts. Data is presented which illustrates the importance of accounting for the presence of crud on fuel cladding surfaces. Several methods used to correct for this phenomenon when collecting and analyzing zirconium alloy field oxide thickness measurements are described. Various observations related to crud characteristics and its impact on fuel rod performance are also addressed. (author)

  17. Control rod for a nuclear reactor

    International Nuclear Information System (INIS)

    Roman, W.G.; Sutton, H.G. Jr.

    1976-01-01

    A control rod assembly for a nuclear reactor is disclosed having a remotely disengageable coupling between the control rod and the control rod drive shaft. The coupling is actuated by first lowering then raising the drive shaft. The described motion causes axial repositioning of a pin in a grooved rotatable cylinder, each being attached to different parts of the drive shaft which are axially movable relative to each other. In one embodiment, the relative axial motion of the parts of the drive shaft is used either to couple or to uncouple the connection by forcing resilent members attached to the drive shaft into or out of shouldered engagement, respectively, with an indentation formed in the control rod

  18. Control rod for a nuclear reactor

    Science.gov (United States)

    Roman, Walter G.; Sutton, Jr., Harry G.

    1979-01-01

    A control rod assembly for a nuclear reactor is disclosed having a remotely disengageable coupling between the control rod and the control rod drive shaft. The coupling is actuated by first lowering then raising the drive shaft. The described motion causes axial repositioning of a pin in a grooved rotatable cylinder, each being attached to different parts of the drive shaft which are axially movable relative to each other. In one embodiment, the relative axial motion of the parts of the drive shaft is used either to couple or to uncouple the connection by forcing resilient members attached to the drive shaft into or out of shouldered engagement, respectively, with an indentation formed in the control rod.

  19. Control rod for a nuclear reactor

    International Nuclear Information System (INIS)

    Roman, W.G.; Sutton, H.G. Jr.

    1979-01-01

    A control rod assembly for a nuclear reactor is disclosed having a remotely disengageable coupling between the control rod and the control rod drive shaft. The coupling is actuated by first lowering then raising the drive shaft. The described motion causes axial repositioning of a pin in a grooved rotatable cylinder, each being attached to different parts of the drive shaft which are axially movable relative to each other. In one embodiment, the relative axial motion of the parts of the drive shaft is used either to couple or to uncouple the connection by forcing resilient members attached to the drive shaft into or out of shouldered engagement, respectively, with an indentation formed in the control rod

  20. Determination of the filling gas composition of the Embalse nuclear power plant's fuel rods by gas chromatography

    International Nuclear Information System (INIS)

    Serdeiro, N.H.; Correia, R.J.

    1990-01-01

    The present method is appropriate to evaluate the argon content in its mixture with hellium in Embalse nuclear power plant tubes, without active material. The maximum relative error within the range from 20% to 80% of argon, is of 1%. The detection limit for the possible nitrogen presence is in the order of magnitude of 5ppm. (Author) [es

  1. Fuel Rod Vibration Measurement Method using a Flap and its Verification

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, Joo Young; Park, Nam Gyu; Suh, Jung Min; Jeon, Kyeong Lak [KEPCO NF Co., Daejeon (Korea, Republic of)

    2011-10-15

    Flow-induced vibration is a critical factor for the mechanical integrity of a fuel rod. This vibration can cause leaked fuel through the mechanism, such as grid to rod fretting. To minimize the failures caused by flow-induced vibration, a robust design is needed which takes into account vibrational characteristics. That is, the spacer grid design should be developed to avoid any excessive vibration. On the one hand, if fuel rod vibration can be measured, an estimation of the excitation forces, which are a critical cause of rod failure, should be possible. Therefore, by applying an external force, flow-induced vibration can be roughly estimated when the fuel rod vibration model is used. KEPCO Nuclear Fuel developed the test loop to research flow-induced vibration as shown in Fig.1. The investigation flow-induced vibration (INFINIT) - the test facility - can measure the grid strap vibration and pressure drop of a 5x5 small scale fuel bundle. Basically, using a Laser Doppler Vibrometer (LDV), the vibration of a structure immersed in high speed fluid can be measured. Grid strap vibration is easily measured using an LDV. However, it is quite difficult to measure fuel rod vibration because of the round surface shape of the rods. In addition, measuring current method using the LDV, it was only possible to directly measure fuel rod vibration at the first row of the bundle as the rods behind the first row are obscured. To solve this problem, a thin flap, as shown in Fig. 2(a) can be used as a reflecting target, gaining access to rods within the bundle. The flap is attached to the fuel rod, as in Fig. 2(b). As a result, most of the inner rod vibration can be measured. Before using a flap to measure fuel rod vibration, a verification process was needed to show whether the LDV signal from the flap vibration provided equivalent and reliable signals. Therefore, impact testing was carried out on the fuel rod using a flap. The LDV signals were then compared with accelerometer

  2. Quality assurance of nuclear fuel

    International Nuclear Information System (INIS)

    1994-01-01

    The guide presents the quality assurance requirements to be completed with in the procurement, design, manufacture, transport, handling and operation of the nuclear fuel. The guide also applies to the procurement of the control rods and the shield elements to be placed in the reactor. The guide is mainly aimed for the licensee responsible for the procurement and operation of fuel, for the fuel designer and manufacturer and for other organizations whose activities affect fuel quality, the safety of fuel transport, storage and operation. (2 refs.)

  3. Coupler for nuclear reactor absorber rods

    International Nuclear Information System (INIS)

    Kerz, K.

    1984-01-01

    A coupler is described for absorber rods being suspended during operation of nuclear reactors which includes plurality of actuating elements being movable for individually and jointly releasing the coupler, the movement of each of the actuating elements for releasing the coupler being independently controllable

  4. Full-length fuel rod behavior under severe accident conditions

    International Nuclear Information System (INIS)

    Lombardo, N.J.; Lanning, D.D.; Panisko, F.E.

    1992-12-01

    This document presents an assessment of the severe accident phenomena observed from four Full-Length High-Temperature (FLHT) tests that were performed by the Pacific Northwest Laboratory (PNL) in the National Research Universal (NRU) reactor at Chalk River, Ontario, Canada. These tests were conducted for the US Nuclear Regulatory Commission (NRC) as part of the Severe Accident Research Program. The objectives of the test were to simulate conditions and provide information on the behavior of full-length fuel rods during hypothetical, small-break, loss-of-coolant severe accidents, in commercial light water reactors

  5. Conceptual design report of the SMART fuel rod

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dae Ho; Lee, Chan Bock; Bang, Je Gun; Jung, Yeon Ho [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-03-01

    The SMART fuel rod is based on 17 x 17 KOFA(Korea Fuel Assembly) fuel rod of the 950MWe pressurize water reactor. The fuel stack length of the KOFA is 3658mm, otherwise SMART fuel rod stack length is 2000mm. The fuel rod contains UO{sub 2} pellets with the enrichment of 4.95%. All the fuel in core will be replaced every 35 months. The average LHGR of the fuel rod is 120 W/cm, commercial PWR is 178 W/cm, SMART LHGR is lower about 31% than commercial PWR. The core inlet and outlet temperature of coolant are respectively 270 deg C and 310 deg C, commercial PWR are respectively 291.6 deg C and 326.8 deg C, SMART inlet and outlet temperature is lower averaged 19.2 deg C than commercial PWR. The coolant use mixed soluble ammonia in high purity water and boron is not in. The general performance of the fuel rod UO{sub 2} pellet has been already verified through the sufficient burnup (60,000 MWd/MTU-rod avg.) experience as the rods of same design in commercial PWR's. But cladding corrosion is required the further verification. (author). 13 refs., 3 figs., 8 tabs.

  6. Researches of WWER fuel rods behaviour under RIA accident conditions

    International Nuclear Information System (INIS)

    Nechaeva, O.; Medvedev, A.; Novikov, V.; Salatov, A.

    2003-01-01

    Unirradiated fuel rod and refabricated fuel rod tests in the BIGR as well as acceptance criteria proving absence of fragmentation and the settlement modeling of refabricated fuel rods thermomechanical behavior in the BIGR-tests using RAPTA-5 code are discussed in this paper. The behaviour of WWER type simulators with E110 and E635 cladding was researched at the BIGR reactor under power pulse conditions simulating reactivity initiated accident. The results of the tests in four variants of experimental conditions are submitted. The behaviour of 12 WWER type refabricated fuel rods was researched in the BIGR reactor under power pulse conditions simulating reactivity initiated accident: burnup 48 and 60 MWd/kgU, pulse width 3 ms, peak fuel enthalpy 115-190 cal/g. The program of future tests in the research reactor MIR with high burnup fuel rod (up to 70 MWd/kgU) under conditions simulating design RIA in WWER-1000 is presented

  7. PWR control rod ejection analysis with the numerical nuclear reactor

    International Nuclear Information System (INIS)

    Hursin, M.; Kochunas, B.; Downar, T. J.

    2008-01-01

    During the past several years, a comprehensive high fidelity reactor LWR core modeling capability has been developed and is referred to as the Numerical Nuclear Reactor (NNR). The NNR achieves high fidelity by integrating whole core neutron transport solution and ultra fine mesh computational fluid dynamics/heat transfer solution. The work described in this paper is a preliminary demonstration of the ability of NNR to provide a detailed intra pin power distribution during a control rod ejection accident. The motivation of the work is to quantify the impact on the fuel performance calculation of a more physically accurate representation of the power distribution within the fuel rod during the transient. The paper addresses first, the validation of the transient capability of the neutronic module of the NNR code system, DeCART. For this purpose, a 'mini core' problem consisting of a 3x3 array of typical PWR fuel assemblies is considered. The initial state of the 'mini core' is hot zero power with a control rod partially inserted into the central assembly which is fresh fuel and is adjacent to once and twice burned fuel representative of a realistic PWR arrangement. The thermal hydraulic feedbacks are provided by a simplified fluids and heat conduction solver consistent for both PARCS and DeCART. The control rod is ejected from the central assembly and the transient calculation is performed with DeCART and compared with the results of the U.S. NRC core simulation code PARCS. Because the pin power reconstruction in PARCS is based on steady state intra assembly pin power distributions which do not account for thermal feedback during the transient and which do not take into account neutron leakage from neighboring assemblies during the transient, there are some small differences in the PARCS and DeCART pin power prediction. Intra pin power density information obtained with DeCART represents new information not available with previous generation of methods. The paper then

  8. Development of joining techniques for fabrication of fuel rod simulators

    International Nuclear Information System (INIS)

    Moorhead, A.J.; McCulloch, R.W.; Reed, R.W.; Woodhouse, J.J.

    1980-10-01

    Much of the safety-related thermal-hydraulic tests on nuclear reactors are conducted not in the reactor itself, but in mockup segments of a core that uses resistance-heated fuel rod simulators (FRS) in place of the radioactive fuel rods. Laser welding and furnace brazing techniques are described for joining subassemblies for FRS that have survived up to 1000 h steady-state operation at 700 to 1100 0 C cladding temperatures and over 5000 thermal transients, ranging from 10 to 100 0 C/s. A pulsed-laser welding procedure that includes use of small-diameter filler wire is used to join one end of a resistance heating element of Pt-8 W, Fe-22 Cr-5.5 Al-0.5 Co, or 80 Ni-20 Cr (wt %) to a tubular conductor of an appropriate intermediate material. The other end of the heating element is laser welded to an end plug, which in turn is welded to a central conductor rod

  9. Mechanical stress analysis for a fuel rod under normal operating conditions

    International Nuclear Information System (INIS)

    Pino, Eddy S.; Giovedi, Claudia; Serra, Andre da Silva; Abe, Alfredo Y.

    2013-01-01

    Nuclear reactor fuel elements consist mainly in a system of a nuclear fuel encapsulated by a cladding material subject to high fluxes of energetic neutrons, high operating temperatures, pressure systems, thermal gradients, heat fluxes and with chemical compatibility with the reactor coolant. The design of a nuclear reactor requires, among a set of activities, the evaluation of the structural integrity of the fuel rod submitted to different loads acting on the fuel rod and the specific properties (dimensions and mechanical and thermal properties) of the cladding material and coolant, including thermal and pressure gradients produced inside the rod due to the fuel burnup. In this work were evaluated the structural mechanical stresses of a fuel rod using stainless steel as cladding material and UO 2 with a low degree of enrichment as fuel pellet on a PWR (pressurized water reactor) under normal operating conditions. In this sense, tangential, radial and axial stress on internal and external cladding surfaces considering the orientations of 0 deg, 90 deg and 180 deg were considered. The obtained values were compared with the limit values for stress to the studied material. From the obtained results, it was possible to conclude that, under the expected normal reactor operation conditions, the integrity of the fuel rod can be maintained. (author)

  10. Fuel and nuclear fuel cycle

    International Nuclear Information System (INIS)

    Prunier, C.

    1998-01-01

    The nuclear fuel is studied in detail, the best choice and why in relation with the type of reactor, the properties of the fuel cans, the choice of fuel materials. An important part is granted to the fuel assembly of PWR type reactor and the performances of nuclear fuels are tackled. The different subjects for research and development are discussed and this article ends with the particular situation of mixed oxide fuels ( materials, behavior, efficiency). (N.C.)

  11. Status and development of RBMK fuel rods and reactor materials

    International Nuclear Information System (INIS)

    Bibilashvili, Yu.K.; Reshetnikov, F.G.; Ioltukhovsky, A.G.

    1998-01-01

    The paper presents current status and development of RBMK fuel rods and reactor materials. With regard to fuel rod cladding the following issues have been discussed: corrosion, tensile properties, welding technology and testing of an alternative cladding alloy with a composition of Zr-Nb-Sn-Fe. Erbium doped fuel has been suggested for safety improvement. Also analysis of fuel reliability is presented in the paper. (author)

  12. Models for fuel rod behaviour at high burnup

    Energy Technology Data Exchange (ETDEWEB)

    Jernkvist, Lars O.; Massih, Ali R. [Quantum Technologies AB, Uppsala Science Park, Uppsala (Sweden)

    2004-12-01

    This report deals with release of fission product gases and irradiation-induced restructuring in uranium dioxide nuclear fuel. Waterside corrosion of zirconium alloy clad tubes to light water reactor fuel rods is also discussed. Computational models, suitable for implementation in the FRAPCON-3.2 computer code, are proposed for these potentially life-limiting phenomena. Hence, an integrated model for the calculation or thermal fission gas release by intragranular diffusion, gas trapping in grain boundaries, irradiation-induced re-solution, grain boundary saturation, and grain boundary sweeping in UO{sub 2} fuel, under time varying temperature loads, is formulated. After a brief review of the status of thermal fission gas release modelling, we delineate the governing equations for the aforementioned processes. Grain growth kinetic modelling is briefly reviewed and pertinent data on grain growth of high burnup fuel obtained during power ramps in the Third Risoe Fission Gas Release Project are evaluated. Sample computations are performed, which clearly show the connection between fission gas release and gram growth as a function of time at different isotherms. Models are also proposed for the restructuring of uranium dioxide fuel at high burnup, the so-called rim formation, and its effect on fuel porosity build-up, fuel thermal conductivity and fission gas release. These models are assessed by use of recent experimental data from the High Burnup Rim Project, as well as from post irradiation examinations of high-burnup fuel, irradiated in power reactors. Moreover, models for clad oxide growth and hydrogen pickup in PWRs, applicable to Zircaloy-4, ZIRLO or M5 cladding, are formulated, based on recent in-reactor corrosion data for high-burnup fuel rods. Our evaluation of these data indicates that the oxidation rate of ZIRLO-type materials is about 20% lower than for standard Zircaloy-4 cladding under typical PWR conditions. Likewise, the oxidation rate of M5 seems to be

  13. Nuclear fuel pellet charging device

    International Nuclear Information System (INIS)

    Komuro, Kojiro.

    1990-01-01

    The present invention concerns a nuclear fuel pellet loading device, in which nuclear fuel pellets are successively charged from an open end of a fuel can while rotating the can. That is, a fuel can sealed at one end with an end plug and opened at the other end is rotated around its pipe axis as the center on a rotationally diriving table. During rotation of the fuel can, nuclear fuel pellets are successively charged by means of a feed rod of a feeding device to the inside of the fuel can. The fuel can is rotated while being supported horizontally and the fuel pellets are charged from the open end thereof. Alternatively, the fuel can is rotated while being supported obliquely and the fuel pellets are charged gravitationally into the fuel can. In this way, the damages to the barrier of the fuel can can be reduce. Further, since the fuel pellets can be charged gravitationally by rotating the fuel can while being supported obliquely, the damages to the barrier can be reduced remarkably. (I.S.)

  14. Corrosion behaviour of zircaloy 4 fuel rod cladding in EDF power plants

    Energy Technology Data Exchange (ETDEWEB)

    Romary, H; Deydier, D [EDF, Direction de l` Equipment SEPTEN, Villeurbanne (France)

    1997-02-01

    Since the beginning of the French nuclear program, a surveillance of fuel has been carried out in order to evaluate the fuel behaviour under irradiation. Until now, nuclear fuels provided by suppliers have met EDF requirements concerning fuel behaviour and reliability. But, the need to minimize the costs and to increase the flexibility of the power plants led EDF to the definition of new targets: optimization of the core management and fuel cycle economy. The fuel behaviour experience shows that some of these new requirements cannot be fully fulfilled by the present standard fuel due to some technological limits. Particularly, burnup enhancement is limited by the oxidation and the hydriding of the Zircaloy 4 fuel rod cladding. Also, fuel suppliers and EDF need to have a better knowledge of the Zy-4 cladding behaviour in order to define the existing margins and the limiting factors. For this reason, in-reactor fuel characterization programs have been set up by fuel suppliers and EDF for a few years. This paper presents the main results and conclusions of EDF experience on Zy-4 in-reactor corrosion behaviour. Data obtained from oxide layer or zirconia thickness measurements show that corrosion performance of Zy-4 fuel rod cladding, as irradiated until now in EDF reactors, is satisfactory but not sufficient to meet the future needs. The fuel suppliers propose in order to improve the corrosion resistance of fuel rod cladding, low tin Zy-4 cladding and then optimized Zy-4 cladding. Irradiation of these claddings are ongoing. The available corrosion data show the better in-reactor corrosion resistance of optimized Zy-4 fuel rod cladding compared to the standard Zy-4 cladding. The scheduled fuel surveillance program will confirm if the optimized Zy-4 fuel rod cladding will meet the requirements for the future high burnup and high flexibility fuel. (author). 10 refs, 19 figs, 4 tabs.

  15. Pressure equalization systems in pressurized water reactor fuel rods

    International Nuclear Information System (INIS)

    Steven, J.; Wunderlich, F.

    1979-01-01

    For the development of a pressure reduction system in PWR fuel rods the capability of charcoal to adsorb Helium, Xenon and Krypton at temperatures of about 300 0 C was investigated. The influence of the adsorption on fuel rod internal pressure and in creep strain on the tube was evaluated in a design study. (orig.) [de

  16. Evaluation of fuel rods behavior - under irradiation test

    International Nuclear Information System (INIS)

    Lameiras, F.S.; Terra, J.L.; Pinto, L.C.M.; Dias, M.S.; Pinheiro, R.B.

    1981-04-01

    By the accompanying of the irradiation of instrumented test fuel rods simulating the operational conditions in reactors, plus the results of post - irradiation exams, tests, evaluation and calibration of analitic modelling of such fuel rods is done. (E.G.) [pt

  17. Assembly mechanism for nuclear fuel bundles

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  18. Nuclear Fuel elements

    International Nuclear Information System (INIS)

    Hirakawa, Hiromasa.

    1979-01-01

    Purpose: To reduce the stress gradient resulted in the fuel can in fuel rods adapted to control the axial power distribution by the combination of fuel pellets having different linear power densities. Constitution: In a fuel rod comprising a first fuel pellet of a relatively low linear power density and a second fuel pellet of a relatively high linear power density, the second fuel pellet is cut at its both end faces by an amount corresponding to the heat expansion of the pellet due to the difference in the linear power density to the adjacent first fuel pellet. Thus, the second fuel pellet takes a smaller space than the first fuel pellet in the fuel can. This can reduce the stress produced in the portion of the fuel can corresponding to the boundary between the adjacent fuel pellets. (Kawakami, Y.)

  19. Experience feedback of operational events of the control rod assembly and its drive mechanism in nuclear power plants

    International Nuclear Information System (INIS)

    Zhou Hong; Xiao Zhi; Tao Shusheng; Zheng Lixin; Chen Zhaolin

    2013-01-01

    Seventeen operational events of the control rod assembly and its drive mechanism are collected from 1992 to 2012 important nuclear operational events and feedback in referred nuclear power plants. After investigated and classified, several important issues, such as the impact of control rod swell and fuel assembly distortion, control rod drive mechanism leakage, and the control system reliability of control rod, are emphatically analyzed. Some suggestions of experience feedback are proposed. (authors)

  20. Design of active-neutron fuel rod scanner

    International Nuclear Information System (INIS)

    Griffith, G.W.; Menlove, H.O.

    1996-01-01

    An active-neutron fuel rod scanner has been designed for the assay of fissile materials in mixed oxide fuel rods. A 252 Cf source is located at the center of the scanner very near the through hole for the fuel rods. Spontaneous fission neutrons from the californium are moderated and induce fissions within the passing fuel rod. The rod continues past a combined gamma-ray and neutron shield where delayed gamma rays above 1 MeV are detected. We used the Monte Carlo code MCNP to design the scanner and review optimum materials and geometries. An inhomogeneous beryllium, graphite, and polyethylene moderator has been designed that uses source neutrons much more efficiently than assay systems using polyethylene moderators. Layers of borated polyethylene and tungsten are used to shield the detectors. Large NaI(Tl) detectors were selected to measure the delayed gamma rays. The enrichment zones of a thermal reactor fuel pin could be measured to within 1% counting statistics for practical rod speeds. Applications of the rod scanner include accountability of fissile material for safeguards applications, quality control of the fissile content in a fuel rod, and the verification of reactivity potential for mixed oxide fuels. (orig.)

  1. A Path Forward to Advanced Nuclear Fuels: Spectroscopic Calorimetry of Nuclear Fuel Materials

    International Nuclear Information System (INIS)

    Tobin, J.G.

    2009-01-01

    The goal is to relieve the shortage of thermodynamic and kinetic information concerning the stability of nuclear fuel alloys. Past studies of the ternary nuclear fuel UPuZr have demonstrated constituent redistribution when irradiated or with thermal treatment. Thermodynamic data is key to predicting the possibilities of effects such as constituent redistribution within the fuel rods and interaction with cladding materials

  2. Internal hydriding in irradiated defected Zircaloy fuel rods: A review (LWBR Development Program)

    International Nuclear Information System (INIS)

    Clayton, J.C.

    1987-10-01

    Although not a problem in recent commercial power reactors, including the Shippingport Light Water Breeder Reactor, internal hydriding of Zircaloy cladding was a persistent cause of gross cladding failures during the 1960s. It occurred in the fuel rods of water-cooled nuclear power reactors that had a small cladding defect. This report summarizes the experimental findings, causes, mechanisms, and methods of minimizing internal hydriding in defected Zircaloy-clad fuel rods. Irradiation test data on the different types of defected fuel rods, intentionally fabricated defected and in-pile operationally defected rods, are compared. Significant factors affecting internal hydriding in defected Zircaloy-clad fuel rods (defect hole size, internal and external sources of hydrogen, Zircaloy cladding surface properties, nickel alloy contamination of Zircaloy, the effect of heat flux and fluence) are discussed. Pertinent in-pile and out-of-pile test results from Bettis and other laboratories are used as a data base in constructing a qualitative model which explains hydrogen generation and distribution in Zircaloy cladding of defected water-cooled reactor fuel rods. Techniques for minimizing internal hydride failures in Zircaloy-clad fuel rods are evaluated

  3. Influence of pellet-clad-gap-size on LWR fuel rod performance

    International Nuclear Information System (INIS)

    Brzoska, B.; Fuchs, H.P.; Garzarolli, F.; Manzel, R.

    1979-01-01

    The as-fabricated pellet-clad-gap size varies due to fabricational tolerances of the cladding inner diameter and the pellet outer diameter. The consequences of these variations on the fuel rod behaviour are analyzed using the KWU fuel rod code CARO. The code predictions are compared with experimental results of special pathfinder test fuel rods irradiated in the OBRIGHEIM nuclear power plant. These test fuel rods include gap sizer in the range of 140 μm to 270 μm, prepressurization between 13 bar to 36 bar and Helium and Argon fill gases irradiated up to a local burnup of 35 MWd/kg(U). Post irradiation examination were performed at different burnups. CARC calculations have been performed with special emphasis in cladding creep down, fission gas release and pellet clad gap closure. (orig.)

  4. 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)

  5. Steady State and Transient Fuel Rod Performance Analyses by Pad and Transuranus Codes

    International Nuclear Information System (INIS)

    Slyeptsov, O.; Slyeptsov, S.; Kulish, G.; Ostapov, A.; Chernov, I.

    2013-01-01

    The report performed under IAEA research contract No.15370/L2 describes the analysis results of WWER and PWR fuel rod performance at steady state operation and transients by means of PAD and TRANSURANUS codes. The code TRANSURANUS v1m1j09 developed by Institute for of Transuranium Elements (ITU) was used based on the Licensing Agreement N31302. The code PAD 4.0 developed by Westinghouse Electric Company was utilized in the frame of the Ukraine Nuclear Fuel Qualification Project for safety substantiation for the use of Westinghouse fuel assemblies in the mixed core of WWER-1000 reactor. The experimental data for the Russian fuel rod behavior obtained during the steady-state operation in the WWER-440 core of reactor Kola-3 and during the power transients in the core of MIR research reactor were taken from the IFPE database of the OECD/NEA and utilized for assessing the codes themselves during simulation of such properties as fuel burnup, fuel centerline temperature (FCT), fuel swelling, cladding strain, fission gas release (FGR) and rod internal pressure (RIP) in the rod burnup range of (41 - 60) GWD/MTU. The experimental data of fuel behavior at steady-state operation during seven reactor cycles presented by AREVA for the standard PWR fuel rod design were used to examine the code FGR model in the fuel burnup range of (37 - 81) GWD/MTU. (author)

  6. Sensitivity Analysis of Gap Conductance for Heat Split in an Annular Fuel Rod

    International Nuclear Information System (INIS)

    Chun, Kun Ho; Chun, Tae Hyun; In, Wang Kee; Song, Keun Woo

    2006-01-01

    To increase of the core power density in the current PWR cores, an annular fuel rod was proposed by MIT. This annular fuel rod has two coolant channels and two cladding-pellet gaps unlike the current solid fuel rod. It's important to predict the heat split reasonably because it affects coolant enthalpy rise in each channel and Departure from Nuclear Boiling Ratio (DNBR) in each channel. Conversely, coolant conditions affect fuel temperature and heat split. In particular if the heat rate leans to either inner or outer channel, it is out of a thermal equilibrium. To control a thermal imbalance, placing another gap in the pellet is introduced. The heat flow distribution between internal and external channels as well as fuel and cladding temperature profiles is calculated with and without the fuel gap between the inner and outer pellets

  7. Measuring element for determining the internal pressure in fuel rods

    International Nuclear Information System (INIS)

    Deckers, H.; Drexler, H.; Reiser, H.

    1983-01-01

    A pressure cell is situated inside the fuel rod, which contains a magnetic core or a core influenced by magnetism, whose position relative to an outer front surface of an end stopper of the fuel rod can vary. The fuel rod contains a pressure cell directly above the lower end stopper or connected to it. This can consist of closed bellows, where if the internal pressure in the fuel rod rises, a ferrite core moves axially. When the pressure drops, this returns to the initial position, which is precisely defined by a stop. To detect a rod defect, the position of the soft iron core relative to the lower edge of the end stopper is scanned by a special measuring device. (orig./HP) [de

  8. Control rod for the operation of nuclear reactor

    International Nuclear Information System (INIS)

    Ishida, Hiromi

    1987-01-01

    Purpose: To conduct spectrum shift operation without complicating the reactor core structures, reducing the probability of failures. Constitution: An operation control rod which is driven while passed vertically in the reactor core comprises a strong absorption portion, moderation portion and weak moderation portion defined orderly from above to below and the length for each of the portions is greater than the effective reactor core height. If the operation control rod is lifted to the maximum limit in the upward direction of the reactor core, the weak moderation portion is corresponded over the effective length of the reactor core. Since the weak moderation portion is filled with zirconium and moderators are not present in the operation control rod, water draining gap is formed, neutron spectral shift is formed, excess reactivity is suppressed, absorption of neutrons to fuel fertile material is increased and the formation of nuclear fission material is increased. From the middle to the final stage of the cycle, the control rod is lowered, by which the moderator/fuel effective volume ratio is increased to increase the reactivity. (Kamimura, M.)

  9. Fuel assembly for a nuclear reactor

    International Nuclear Information System (INIS)

    Gjertsen, R.K.

    1982-01-01

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

  10. Behavior of defective LWR-type fuel rods irradiated under postulated accident conditions

    International Nuclear Information System (INIS)

    Hobbins, R.R.; Croucher, D.W.; Seiffert, S.L.; Cook, B.A.; Kerwin, D.K.; Mehner, A.S.; Ploger, S.A.

    1979-05-01

    The irradiation experiments reported here have been conducted by the Thermal Fuels Behavior Program of EG and G Idaho, Inc., for the United States Nuclear Regulatory Commission in the Power Burst Facility (PBF) at the Idaho National Engineering Laboratory. Five of the rods were irradiated in PCM tests and one in a LOC test. During these tests, the six rods lost cladding integrity prior to or during the transient phase of the test due to either manufacturing defects or intentional rod design and operation. Of the five defective rods tested under PCM conditions, one (Rod IE-008, Test IE-1) had a hydride rupture below the region of the rod, which was in film boiling during the transient; two (Rod A-0021, Test PCM-3 and Rod IE-019, Test IE-5) contained defects (a pin hole and a small axial crack, respectively) within the film boiling zone; and two (Rod 201-1, Test PCM-1 and Rod 205-8, Test PCM-5) failed by cladding embrittlement within the film boiling zone. Rod 312-3 was waterlogged before being subjected to LOC conditions in Test LLR-3

  11. 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

  12. COMETHE III-M for transient fuel rod behaviour prediction

    International Nuclear Information System (INIS)

    Billaux, M.; Vliet, J. van

    1983-01-01

    The COMETHE III-M version is being developed in order to provide fuel rod behaviour prediction capability both in steady-state and in transient situations. It also allows to estimate the fuel rod enthalpy evolution versus time or burnup which may be important in core-related safety studies. This paper describes the transient heat transfer models, including transient heat conduction inside the fuel rod, and a subchannel model providing transient flow as well as enthalpy calculation capability. Transient fission gas release is also modelled on basis of the change rate of oxide temperature. The models are illustrated by a few calculation examples. (author)

  13. Gamma-ray spectroscopy on irradiated fuel rods

    International Nuclear Information System (INIS)

    Terremoto, Luis Antonio Albiac

    2009-01-01

    The recording of gamma-ray spectra along an irradiated fuel rod allows the fission products to be qualitatively and quantitatively examined. Among all nondestructive examinations performed on irradiated fuel rods by gamma-ray spectroscopy, the most comprehensive one is the average burnup measurement, which is quantitative. Moreover, burnup measurements by means of gamma-ray spectroscopy are less time-consuming and waste-generating than burnup measurements by radiochemical, destructive methods. This work presents the theoretical foundations and experimental techniques necessary to measure, using nondestructive gamma-ray spectroscopy, the average burnup of irradiated fuel rods in a laboratory equipped with hot cells. (author)

  14. Method of manufacturing nuclear fuel pellet

    International Nuclear Information System (INIS)

    Oguma, Masaomi; Masuda, Hiroshi; Hirai, Mutsumi; Tanabe, Isami; Yuda, Ryoichi.

    1989-01-01

    In a method of manufacturing nuclear fuel pellets by compression molding an oxide powder of nuclear fuel material followed by sintering, a metal nuclear material is mixed with an oxide powder of the nuclear fuel material. As the metal nuclear fuel material, whisker or wire-like fine wire or granules of metal uranium can be used effectively. As a result, a fuel pellet in which the metal nuclear fuel is disposed in a network-like manner can be obtained. The pellet shows a great effect of preventing thermal stress destruction of pellets upon increase of fuel rod power as compared with conventional pellets. Further, the metal nuclear fuel material acts as an oxygen getter to suppress the increase of O/M ratio of the pellets. Further, it is possible to reduce the swelling of pellet at high burn-up degree. (T.M.)

  15. Cracking and relocation of UO2 fuel during nuclear operation

    International Nuclear Information System (INIS)

    Appelhans, A.D.; Dagbjartsson, S.J.

    1981-01-01

    Cracking and relocation of light water reactor (LWR) fuel pellets affect the axial gas flow path within nuclear reactor fuel rods and the thermal performance of the fuel. As part of the Nuclear Regulatory Commission's Water Reactor Safety Research Fuel Behavior Program, the Thermal Fuels Behavior Program of EG and G Idaho, Inc., is conducting fuel rod behavior studies in the Heavy Boiling Water Reactor in Halden, Norway. The Instrumental Fuel Assembly-430 (IFA-430) operated in that facility is a multipurpose assembly designed to provide information on fuel cracking and relocation, the long-term thermal response of LWR fuel rods subjected to various internal pressures and gas compositions, and the release of fission gases. This report presents the results of an analysis of fuel cracking and relocation phenomena as deduced from fuel rod axial gas flow and fuel temperature data from the first 6.5 GWd/tUO 2 burnup of the IFA-430

  16. SEFLEX - fuel rod simulator effects in flooding experiments. Pt. 2

    International Nuclear Information System (INIS)

    Ihle, P.; Rust, K.

    1986-03-01

    This report presents typical data and a limited heat transfer analysis from unblocked bundle reflood tests of an experimental thermal-hydraulic program. Full-length bundles of 5 x 5 fuel rod simulators having a gas-filled gap between the Zy cladding and the alumina pellets were tested in the test rig designed for the earlier Flooding Experiments with Blocked Arrays (FEBA-program). The 5 x 5 FEBA rod bundle tests were performed with gapless heater rods. These rods have a close thermal contact between the stainless steel cladding and the electric insulation material. A comparison of the SEFLEX data with the reference data of FEBA obtained under identical initial and reflood conditions shows the influence of different fuel rod simulators on the thermal-hydraulic behavior during forced feed bottom reflooding of unblocked and blocked arrays. Compared to bundles of gapless rods, bundles of rods with Zy claddings and a gas filled gap between claddings and pellets, which more closely represent the features that exist in an actual fuel rod geometry, produced higher quench front velocities, enhanced removal of stored heat in the rods, reduced peak cladding temperatures, increased grid spacer effects and absolutely unproblematic coolability of 90 percent blockages with bypass. The data offer the opportunity for further validation of computer codes to make realistic predictions of safety margins during a LOCA in a PWR. (orig./HP) [de

  17. SEFLEX fuel rod simulator effects in flooding experiments. Pt. 3

    International Nuclear Information System (INIS)

    Ihle, P.; Rust, K.

    1986-03-01

    This report presents typical data and a limited heat transfer analysis from blocked bundle reflood tests of an experimental thermal-hydraulic program. Full-length bundles of 5x5 fuel rod simulators having a gas-filled gap between the Zy cladding and the alumina pellets were tested in the test rig designed for the earlier Flooding Experiments with Blocked Arrays (FEBA-program). The 5x5 FEBA rod bundle tests were performed with gapless heater rods. These rods have a close thermal contact between the stainless steel cladding and the electric insulation material. A comparison of the SEFLEX data with the reference data of FEBA obtained under identical initial and reflood conditions shows the influence of different fuel rod simulators on the thermal-hydraulic behavior during forced feed bottom reflooding of unblocked and blocked arrays. Compared to bundles of gapless rods, bundles of rods with Zy claddings and a gas filled gap between claddings and pellets, which more closely represent the features that exist in an actual fuel rod geometry, produced higher quench front velocities, enhanced removal of stored heat in the rods, reduced peak cladding temperatures, increased grid spacer effects and absolutely unproblematic coolability of 90 percent blockages with bypass. The data offer the opportunity for further validation of computer codes to make realistic predictions of safety margins during a LOCA in a PWR. (orig./HP) [de

  18. The M5 Fuel Rod Cladding

    International Nuclear Information System (INIS)

    Mardon, J.P.; Charquet, D.; Senevat, J.

    1998-01-01

    The large-scale program for the development and irradiation of new Zr alloys started by FRAMATOME and its industrial partners CEZUS and ZIRCOTUBE more than 10 years ago is now enabling FRAGEMA to offer the ternary M5 (ZrNbO) as the cladding material for PWR advanced fuel rods. Compared with the former product (low-tin-Zircaloy-4), this alloy exhibits impressive gains under irradiation at extended burnup (55 GWd/t) relatively to corrosion (factor 3 to 4), hydriding (factor 5 to 6), growth and creep (factor 2 to 3). In this paper, we shall successively address: - the industrial development and manufacturing experience - the corrosion, hydriding, creep and growth performances obtained over a wide range of PWR normal irradiation conditions (France and other countries) up to burnups of 55 GWd/t - The interpretation of these results by means of analytical experiments conducted in test reactors (free growth, creep) and microstructural observations on the irradiated material - and the behaviour under accident (LOCA) and severe environment and irradiation (Li, boiling) conditions. (Author)

  19. 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.

  20. 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

  1. 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.)

  2. The measurements of critical mass with uranium fuel elements and thorium rods

    International Nuclear Information System (INIS)

    Yao Zhiquan; Chen Zhicheng; Yao Zewu; Ji Huaxiang; Bao Borong; Zhang Jiahua

    1991-01-01

    The critical experiments with uranium elements and Thorium rods have been performed in zero power reactor at Shanghai Institute of Nuclear Research. The critical masses have been measured in various U/Th ratios. The fuels are 3% 235 U-enriched uranium. The Thorium rods are made from power of ThF 4 . Ratios of calculated values to experimental values are nearly constant at 0.995

  3. Does rim microstructure formation degrade the fuel rod performance?

    International Nuclear Information System (INIS)

    Baron, D.; Spino, J.

    2002-01-01

    High burnup extension of LWR fuel is progressing to reduce the total process flow and eventually the costs of the nuclear fuel cycle. A particular fuel restructuring at high burnups, commonly observed at the periphery of LWR fuel pellets (rim structure), but also in FBR fuels to some extent and in the Plutonium rich clusters of the MOX Fuels, was considered a priori as a limitation for burnup extension. Since more than ten years this rim effect have been deeply investigated. Its causes and consequences are however not yet totally elucidated. The three steps actually identified of this phenomenon are first a progressive disappearing of the intra-granular Xenon, the outset of numerous 0.5 to 1 m pores and finally a grain subdivision around the pores. Penalty of the porosity increase on the thermal conductivity is obvious. One expect the fission gases to remain trapped in the rim porosity up to a 75 MWd/kgUO 2 local burnup. Above this threshold, 15 to 20 % of the fission gases seem to be quickly released. Microindentation tests conducted at ITU have shown the rim structure to resist fracture extension under punching. It is still open whether this implies certain ductility and viscosity of the material, or if it corresponds to stress relaxation by microcracking. Whatever the case be, it is suggested that the rim material would be able to decrease the interaction stresses and to equalise the cladding strains during a power ramp. Moreover, in the RIA tests, it was concluded so far that the grain de-cohesion caused by gas expansion at the grain boundaries was responsible for the cladding strain and failure. However, not the rim zone was affected by grain de-cohesion but the region adjacent to it. Therefore, in front of the question whether the rim structure degrades the fuel rod behaviour, we continue to argue on its benefit for fuel burnup extension. (author)

  4. Nuclear fuel

    International Nuclear Information System (INIS)

    Quinauk, J.P.

    1990-01-01

    Since 1985, Fragema has been marketing and selling the Advanced Fuel Assemby AFA whose main features are its zircaloy grids and removable top and bottom nozzles. It is this product, which exists for several different fuel assembly arrays and heights, that will be employed in the reactors at Daya Bay. Fragema employs gadolinium as the consumable poison to enable highperformance fuel management. More recently, the company has supplied fuel assemblies of the mixed-oxide(MOX) and enriched reprocessed uranium type. The reliability level of the fuel sold by Fragema is one of the highest in the world, thanks in particular to the excellence of the quality assurance and quality control programs that have been implemented at all stages of its design and manufacture

  5. Failed fuel rod detection system and computerized manipulator during outages

    International Nuclear Information System (INIS)

    Boehm, H.H.; Foerch, H.

    1984-01-01

    During regular outages spent fuel assemblies need to be replaced and relocated within the core. Defective fuel rods in particular fuel assemblies have to be removed from further service and before delivery of such faulty fuel assemblies to a reprocessing plant. The system which Brown Boveri Reaktor GmbH and Krautkraemer have developed in the Federal Republic of Germany is capable of directly locating the defective rods in a proper fuel assembly. Inspection times are comparable to those of standard sipping methods, with the advantages of immediately available results and direct identification of the defective fuel rods. During the repair of fuel assemblies this system allows withdrawal of individual defective rods. With the sipping method all the fuel rods of a defective fuel assembly need to be removed and inspected by eddy current testing. During steam generator inspection and repair personnel are exposed to ample radiation. A remotely controlled, computerized manipulator was used to significantly reduce the radiation dose by automating steps in the procedures; at the same time inspection and repair times were reduced. The main features of the manipulator are a rigid component construction of the leg and two arms, and a resolver control for horizontal and vertical motion that enables rapid and accurate access to a desired tube (author)

  6. Failed fuel rod detection method by ultrasonic wave

    International Nuclear Information System (INIS)

    Takamatsu, Masatoshi; Muraoka, Shoichi; Ono, Yukio; Yasojima, Yujiro.

    1990-01-01

    Ultrasonic wave signals sent from an ultrasonic receiving element are supplied to an evaluation circuit by way of a gate. A table for gate opening and closing timings at the detecting position in each of the fuel rods in a fuel assembly is stored in a memory. A fuel rod is placed between an ultrasonic transmitting element and the receiving element to determine the positions of the transmitting element and the receiving element by positional sensors. The opening and closing timings at the positions corresponding to the result of the detection are read out from the table, and the gates are opened and closed by the timing. This can introduce the ultrasonic wave signals transmitted through a control rod always to the evaluation circuit passing through the gate. Accordingly, the state of failure of the fuel rod can be detected accurately. (I.N.)

  7. Ultrasonic inspection for testing the PWR fuel rod endplug welds

    International Nuclear Information System (INIS)

    Pillet, C.; Destribats, M.T.; Papezyk, F.

    1976-01-01

    A method of ultrasonic testing with local immersion and transversal waves was developed. It is possible to detect defects as the lacks of fusion and penetration and porosity in the PWR fuel rod endplug welds [fr

  8. Analysis of irradiation temperature in fuel rods of OGL-1 fuel assembly

    International Nuclear Information System (INIS)

    Fukuda, Kousaku; Kobayashi, Fumiaki; Minato, Kazuo; Ikawa, Katsuichi; Iwamoto, Kazumi

    1984-10-01

    Irradiation temperature in the fuel rods of 5th OGL-1 fuel assembly was analysed by the system composed by STPDSP2 and TRUMP codes. As the measured input-data, following parameters were allowed for; circumferential heating distribution around the fuel rod, which was measured in the JMTR critical assembly, axial heating distribution through the fuel rod, ratio of peak heatings of three fuel rods, and pre- and post-irradiation outer radii of the fuel compacts and inner radii of the graphite sleeves, which had been measured in PIE of the 5th OGL-1 fuel assembly. In computation the axial distributions of helium coolant temperature through the fuel rod and the heating value of each fuel rod were, firstly, calculated as input data for TRUMP. The TRUMP calculation yielded the temperatures which were fitted in those measured by all of the thermo-couples installed in the fuel rods, by adjusting only the value of the surface heat transfer coefficient, and consequently, the temperatures in all portions of the fuel rod were obtained. The apparent heat transfer coefficient changed to 60% of the initial values in the middle period of irradiation. For this reduction it was deduced that shoot had covered the surface of the fuel rod during irradiation, which was confirmed in PIE. Beside it, several things were found in this analysis. (author)

  9. Fuel rod behavior of a PWR during load following

    International Nuclear Information System (INIS)

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

    1982-01-01

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

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

    International Nuclear Information System (INIS)

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

    1981-01-01

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

  11. Study of correcting the effect of daughter age on determining 235U enrichment of fuel rods

    International Nuclear Information System (INIS)

    Deng Jingshan; Zhou Chengfang; Luo Minxuan; Liu Yun

    1997-01-01

    Gamma-ray passive technique is a very effective method to assay and determine 235 U enrichment of nuclear power plant fuel rods. There is a weakness in this passive method, i.e. only after the uranium isotope daughters of UO 2 pellets have reached to equilibrium with uranium parent, then the 235 U enrichment can be determined. This weakness greatly restricts the application of the method. A new two-peak and two-window technique is developed that can overcome the interference of uranium daughter decay in determining 235 U enrichment of nuclear fuel rods, and the results are very satisfactory. The new technique will play an important role in the gamma-ray passive technique for determining 235 U enrichment of fuel rods. This new technique also makes the gamma-ray passive method perfectly. (11 figs., 6 tabs.)

  12. Rodded shutdown system for a nuclear reactor

    International Nuclear Information System (INIS)

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

    1978-01-01

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

  13. Analysis of Radial Plutonium Isotope Distribution in Irradiated Test MOX Fuel Rods

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Jae Yong; Lee, Byung Ho; Koo, Yang Hyun; Kim, Han Soo

    2009-01-15

    After Rod 3 and 6 (KAERI MOX) were irradiated in the Halden reactor, their post-irradiation examinations are being carried out now. In this report, PLUTON code was implemented to analyze Rod 3 and 6 (KAERI MOX). In the both rods, the ratio of a maximum burnup to an average burnup in the radial distribution was 1.3 and the contents of {sup 239}Pu tended to increase as the radial position approached the periphery of the fuel pellet. The detailed radial distribution of {sup 239}Pu and {sup 240}Pu, however, were somewhat different. To find the reason for this difference, the contents of Pu isotopes were investigated as the burnup increased. The content of {sup 239}Pu decreased with the burnup. The content of {sup 240}Pu increased with the burnup by the 20 GWd/tM but decreased over the 20 GWd/tM. The local burnup of Rod 3 is higher than that of Rod 6 due to the hole penetrated through the fuel rod. The content of {sup 239}Pu decreased more rapidly than that of {sup 240}Pu in the Rod 6 with the increased burnup. It resulted in a radial distribution of {sup 239}Pu and {sup 240}Pu similar to Rod 3. The ratio of Xe to Kr is a parameter to find where the fissions occur in the nuclear fuel. In both Rod 3 and 6, it was 18.3 in the whole fuel rod cross section, which showed that the fissions occurred in the plutonium.

  14. Results of automatic system implementation for the friction control rods execution in Cofrentes nuclear power plant

    International Nuclear Information System (INIS)

    Curiel, M.; Palomo, M. J.; Urrea, M.; Arnaldos, A.

    2010-10-01

    The purpose of this presentation is to show the obtained results in Cofrentes nuclear power plant (Spain) of control rods Pcc/24 friction test procedure. In order to perform this, a control rod friction test system has been developed. Principally, this system consists on software and data acquisition hardware that obtains and analyzes the control rod pressure variation on which the test is being made. The Pcc/24 procedure objective is to detect an excessive friction in the control rod movement that could cause a control rod drive movement slower than usual. This test is necessary every time that an anomalous alteration is produced in the reactor core that could affect to a fuel rod, and it is executed before the time measure of control rods rapid scram test of the affected rods. This test has to be carried out to all the reactor control rods and takes valuable time during plant refuelling. So, by means of an automatic system to perform the test, we obtain an important time saving during refuelling. On the other hand, the on-line monitoring of the control rod insertion and changes in differential pressure, permits a control rod operation fast and safe validation. Moreover, an automatic individual report of every rod is generated by the system and a final global result report of the entire test developed in refuelling is generated. The mentioned reports can be attached directly to the procedure documents obtaining an office data processing important saving time. (Author)

  15. Results of automatic system implementation for the friction control rods execution in Cofrentes nuclear power plant

    International Nuclear Information System (INIS)

    Palomo, M.; Urrea, M.; Arnaldos, A.

    2011-01-01

    The purpose of this presentation is to show the obtained results in Cofrentes Nuclear Power Plant (Spain) of Control Rods PCC/24 Friction Test Procedure. In order to perform this, a Control Rod Friction Test System has been developed. Principally, this system consists on software and data acquisition hardware that obtains and analyzes the control rod pressure variation on which the test is being made. The PCC/24 Procedure objective is to detect an excessive friction in the control rod movement that could cause a CRD (Control Rod Drive) movement slower than usual. This test is necessary every time that an anomalous alteration is produced in the reactor core that could affect to a fuel rod, and it is executed before the time measure of control rods rapid scram test of the affected rods. This test has to be carried out to all the reactor control rods and takes valuable time during plant refuelling. So, by means of an automatic system to perform the test, we obtain an important time saving during refuelling. On the other hand, the on-line monitoring of the control rod insertion and changes in differential pressure, permits a control rod operation fast and safe validation. Moreover, an automatic individual report of every rod is generated by the system and a final global result report of the entire test developed in refuelling is generated. The mentioned reports can be attached directly to the procedure documents obtaining an office data processing important saving time.(author)

  16. Results of automatic system implementation for the friction control rods execution in Cofrentes nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Palomo, M., E-mail: mpalomo@iqn.upv.es [Universidad Politecnica de Valencia (UPV) (Spain); Urrea, M., E-mail: matias.urrea@iberdrola.es [Iberdrola Generacion S.A. Valencia (Spain). C.N. Cofrentes; Curiel, M., E-mail: m.curiel@lainsa.com [Logistica y Acondicionamientos Industriales (LAINSA), Valencia (Spain); Arnaldos, A., E-mail: a.arnaldos@titaniast.com [TITANIA Servicios Teconologicos, Valencia (Spain)

    2011-07-01

    The purpose of this presentation is to show the obtained results in Cofrentes Nuclear Power Plant (Spain) of Control Rods PCC/24 Friction Test Procedure. In order to perform this, a Control Rod Friction Test System has been developed. Principally, this system consists on software and data acquisition hardware that obtains and analyzes the control rod pressure variation on which the test is being made. The PCC/24 Procedure objective is to detect an excessive friction in the control rod movement that could cause a CRD (Control Rod Drive) movement slower than usual. This test is necessary every time that an anomalous alteration is produced in the reactor core that could affect to a fuel rod, and it is executed before the time measure of control rods rapid scram test of the affected rods. This test has to be carried out to all the reactor control rods and takes valuable time during plant refuelling. So, by means of an automatic system to perform the test, we obtain an important time saving during refuelling. On the other hand, the on-line monitoring of the control rod insertion and changes in differential pressure, permits a control rod operation fast and safe validation. Moreover, an automatic individual report of every rod is generated by the system and a final global result report of the entire test developed in refuelling is generated. The mentioned reports can be attached directly to the procedure documents obtaining an office data processing important saving time.(author)

  17. Results of automatic system implementation for the friction control rods execution in Cofrentes nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Curiel, M. [Logistica y Acondicionamientos Industriales SAU, Sorolla Center, local 10, Av. de las Cortes Valencianas, 46015 Valencia (Spain); Palomo, M. J. [ISIRYM, Universidad Politecnica de Valencia, Camino de Vera s/n, Valencia (Spain); Urrea, M. [Iberdrola Generacion S. A., Central Nuclear Cofrentes, Carretera Almansa Requena s/n, 04662 Cofrentes, Valencia (Spain); Arnaldos, A., E-mail: m.curiel@lainsa.co [TITANIA Servicios Tecnologicos SL, Sorolla Center, local 10, Av. de las Cortes Valencianas No. 58, 46015 Valencia (Spain)

    2010-10-15

    The purpose of this presentation is to show the obtained results in Cofrentes nuclear power plant (Spain) of control rods Pcc/24 friction test procedure. In order to perform this, a control rod friction test system has been developed. Principally, this system consists on software and data acquisition hardware that obtains and analyzes the control rod pressure variation on which the test is being made. The Pcc/24 procedure objective is to detect an excessive friction in the control rod movement that could cause a control rod drive movement slower than usual. This test is necessary every time that an anomalous alteration is produced in the reactor core that could affect to a fuel rod, and it is executed before the time measure of control rods rapid scram test of the affected rods. This test has to be carried out to all the reactor control rods and takes valuable time during plant refuelling. So, by means of an automatic system to perform the test, we obtain an important time saving during refuelling. On the other hand, the on-line monitoring of the control rod insertion and changes in differential pressure, permits a control rod operation fast and safe validation. Moreover, an automatic individual report of every rod is generated by the system and a final global result report of the entire test developed in refuelling is generated. The mentioned reports can be attached directly to the procedure documents obtaining an office data processing important saving time. (Author)

  18. Support grid for fuel elements in a nuclear reactor

    International Nuclear Information System (INIS)

    Finch, L.M.

    1977-01-01

    A support grid is provided for holding nuclear fuel rods in a rectangular array. Intersecting sheet metal strips are interconnected using opposing slots in the strips to form a rectangular cellular grid structure for engaging the sides of a multiplicity of fuel rods. Spring and dimple supports for engaging fuel and guide rods extending through each cell in the support grid are formed in the metal strips with the springs thus formed being characterized by nonlinear spring rates

  19. Neutronic evaluation of annular fuel rods to assemblies 13 x 13, 14 x 14 and 15 x 15

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Raphael H.M.; Ramos, Mario C.; Velasquez, Carlos E.; Silva, Clarysson A.M. da; Pereira, Cláubia; Costa, Antonella L., E-mail: rapha.galo@hotmail.com, E-mail: marc5663@gmail.com, E-mail: carlosvelcab@hotmail.com, E-mail: clarysson@nuclear.ufmg.br, E-mail: claubia@nuclear.ufmg.br, E-mail: antonella@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear

    2017-07-01

    Research and development in nuclear reactor field has been proposed a new concept of fuel rod such as annular shape. The design of the annular fuel rods allows the coolant flow through the inner and outer side of it. Such project was proposed as an alternative to the traditional fuel rods used in LWR reactors. This new geometry allows an increase in power density in the reactor core with greater heat transfer from the fuel to the coolant which reduces the temperature in central region of the rod, in which a better configuration and dimension of fuel elements are aimed due to improvement of cooling in possible replacement of PWR traditional rods for annular rods. The aim of this work is to evaluate the neutronic parameters of fuel element with annular fuel rods where three configurations were studied: 13 x 13, 14 x 14 and 15 x 15. The goal is compare the neutronic between the advanced and the standard fuel assembly 16 x 16. In these studies, the external dimension and the moderator to fuel volume ratio (V{sub M}/V{sub F}) of standard 16 x 16 is the same in all annular fuels assemblies. The MCNPX 2.6.0 (Monte Carlo N-Particle eXtended – version 2.6.0) code was used in all simulations. After all procedures, the annular fuel assemblies 13 have obtained greater neutronics parameters and were selected to more neutronics simulations. (author)

  20. Neutronic evaluation of annular fuel rods to assemblies 13 x 13, 14 x 14 and 15 x 15

    International Nuclear Information System (INIS)

    Silva, Raphael H.M.; Ramos, Mario C.; Velasquez, Carlos E.; Silva, Clarysson A.M. da; Pereira, Cláubia; Costa, Antonella L.

    2017-01-01

    Research and development in nuclear reactor field has been proposed a new concept of fuel rod such as annular shape. The design of the annular fuel rods allows the coolant flow through the inner and outer side of it. Such project was proposed as an alternative to the traditional fuel rods used in LWR reactors. This new geometry allows an increase in power density in the reactor core with greater heat transfer from the fuel to the coolant which reduces the temperature in central region of the rod, in which a better configuration and dimension of fuel elements are aimed due to improvement of cooling in possible replacement of PWR traditional rods for annular rods. The aim of this work is to evaluate the neutronic parameters of fuel element with annular fuel rods where three configurations were studied: 13 x 13, 14 x 14 and 15 x 15. The goal is compare the neutronic between the advanced and the standard fuel assembly 16 x 16. In these studies, the external dimension and the moderator to fuel volume ratio (V M /V F ) of standard 16 x 16 is the same in all annular fuels assemblies. The MCNPX 2.6.0 (Monte Carlo N-Particle eXtended – version 2.6.0) code was used in all simulations. After all procedures, the annular fuel assemblies 13 have obtained greater neutronics parameters and were selected to more neutronics simulations. (author)

  1. Development of a Computer Code for the Estimation of Fuel Rod Failure

    Energy Technology Data Exchange (ETDEWEB)

    Rhee, I.H.; Ahn, H.J. [Korea Electric Power Research Institute, Daejeon (Korea, Republic of)

    1997-12-31

    Much research has already been performed to obtain the information on the degree of failed fuel rods from the primary coolant activities of operating PWRs in the last few decades. The computer codes that are currently in use for domestic nuclear power plants, such as CADE code and ABB-CE codes developed by Westinghouse and ABB-CE, respectively, still give significant overall errors in estimating the failed fuel rods. In addition, with the CADE code, it is difficult to predict the degree of fuel rod failures during the transient period of nuclear reactor operation, where as the ABB-CE codes are relatively more difficult to use for end-users. In particular, the rapid progresses made recently in the area of the computer hardware and software systems that their computer programs be more versatile and user-friendly. While the MS windows system that is centered on the graphic user interface and multitasking is now in widespread use, the computer codes currently employed at the nuclear power plants, such as CADE and ABB-CE codes, can only be run on the DOS system. Moreover, it is desirable to have a computer code for the fuel rod failure estimation that can directly use the radioactivity data obtained from the on-line monitoring system of the primary coolant activity. The main purpose of this study is, therefore, to develop a Windows computer code that can predict the location, the number of failed fuel rods,and the degree of failures using the radioactivity data obtained from the primary coolant activity for PWRs. Another objective is to combine this computer code with the on-line monitoring system of the primary coolant radioactivity at Kori 3 and 4 operating nuclear power plants and enable their combined use for on-line evaluation of the number and degree of fuel rod failures. (author). 49 refs., 85 figs., 30 tabs.

  2. 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

  3. 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.)

  4. Thermal hydraulic analysis of gas-cooled reactors with annular fuel rods

    International Nuclear Information System (INIS)

    Han, Kyu Hyun; Chang, Soon Heung

    2005-01-01

    More than half of the world's energy is used in industrial processes and for heating applications which have hardly been touched by the nuclear industry. Nuclear power could be brought into a wide range of applications for industrial processes, provided that gas outlet temperatures of gascooled reactors are sufficiently high. The most limiting core design requirement which controls the core outlet temperature is the maximum acceptable fuel compact temperature. An innovative fuel design is required for a significant decrease in the fuel temperature. This study investigated the possibilities of implementing internally and externally cooled annular fuel rods in a gas-cooled reactor

  5. Axial transport of fission gas in LWR fuel rods

    International Nuclear Information System (INIS)

    Kinoshita, M.

    1983-01-01

    With regard to fission gas transportation inside the fuel rod, the following three mechanisms are important: (1) a localized and time dependent fission gas release from UO 2 fuel to pellet/clad gap, (2) the consequent gas pressure difference between the gap and the plenum, and (3) the inter-diffusion of initially filled Helium and released fission gas such as Xenon. Among these three mechanisms, the 2nd mechanism would result in the one dimensional flow through P/C gap in the axial direction, while the 3rd would average the local fission gas concentration difference. In this paper, an attempt was made to develop a computerized model, LINUS (LINear flow and diffusion under Un-Steady condition) describing the above two mechanisms, items (2) and (3). The item (1) is treated as an input. The code was applied to analyse short length experimental fuel rods and long length commercial fuel rods. The calculated time evolution of Xe concentration along the fuel column shows that the dilution rate of Xe in commercial fuel rods is much slower than that in short experimental fuel rods. Some other sensitivity studies, such as the effect of pre-pressurization, are also presented. (author)

  6. In-pile experiments on fuel rod behavior during a LOCA

    International Nuclear Information System (INIS)

    Karb, E.; Pruessmann, M.; Sepold, L.

    1980-05-01

    This report describes the results of the Test Series F, Tests F 1 through F 5, in the in-pile experimental program with single rods in the DK loop of the FR2 reactor at the Kernforschungszentrum Karlsruhe (KfK). The research is part of the Nuclear Safety Project's (PNS) fuel behavior program. The main objective of the FR2-LOCA tests is to provide information about the effects of a nuclear environment on the mechanisms of fuel rod failure in the second heatup phase of a LOCA. The test rods have a heated length of 50 cm, and their radial dimensions are identical with those of a commercial German PWR. The main parameter of the FR2-LOCA test program is the burnup. The F tests were perfomed from Oct. 25, 1977 to Nov. 22, 1977. They were the first tests in this program to use pre-irradiated fuel rods. The nominal burnup of the test rods was 20 000 MWd/t. During the transient test, the test rods were subjected to rod powers between 36 and 41 W/cm and were pressurized with He to hot internal pressures between 46 and 83 bar. The test rods during the heatup phase at pressures of 56, 53, 42, 72 and 60 bar, respectively. The burst temperatures were determined to be 890, 893, 932, 835 and 880 0 C for test F 1 through F 5. The maximum total circumferential elongations amount to 59, 38, 27, 34 and 41%, respectively. The F tests revealed a fragmentation of the fuel after the irradiation (prior to the tests) and a disintegration of the fuel pellet column after the transient tests due to cladding ballooning. The post-test results indicated a significant reduction of the pellet stack length for all five test rods. The burst data of the F tests did not reveal any difference between tests with unirradiated fuel rods and the irradiated fuel rods of this test series. (orig./HP) [de

  7. In-pile experiments on fuel rod behaviour during a LOCA

    International Nuclear Information System (INIS)

    Sepold, E.H.; Karb, E.H.; Pruessmann, M.

    1981-07-01

    This report describes the results of the Test Series G2/3 within the in-pile experimental program for the investigation of LWR fuel rod behavior. The results were obtained with single rods of a PWR design in the DK loop of the FR2 reactor at the Kernforschungszentrum Karlsruhe (KfK). The in-pile tests with the objective of investigating the influence of a nuclear environment on the mechanisms of fuel rod failure were being performed with irradiated and unirradiated rods. The main parameter of the test program ist the burnup, ranging from 2500 to 35000 MWd/t. The results of test series G2/3 (35000 MWd/t) with respect to the burst data, i.e. burst temperature, burst pressure, and burst strain, do not indicate major differences from the in-pile tests with unirradiated test specimens. (orig.) [de

  8. In-pile experiemts on fuel rod behavior during a LOCA

    International Nuclear Information System (INIS)

    Pruessmann, M.; Karb, E.H.; Sepold, L.

    1981-02-01

    This report describes the results of the Test Series G1 within the in-pile experimental program for the investigation of LWR fuel rod behavior. The results were obtained with single rods of a PWR design in the DK loop of the FR2 reactor at the Kernforschungszentrum Karlsruhe (KfK). The in-pile tests with the objective of investigating the influence of a nuclear environment on the mechansims of fuel rod failure were being performed with irradiated and unirradiated rods. The main parameter of the test program is the burnup ranging from 2500 to 35 000 MWd/t. The results of test series G1 (35 000 MWd/t) with respect to the burst data, i.e. burst temperature, burst pressure, and burst strain, do not indicate major differences from the in-pile tests with unirradiated test specimens. (orig.) [de

  9. Development of Fuel ROd Behavior Analysis code (FROBA) and its application to AP1000

    International Nuclear Information System (INIS)

    Yu, Hongxing; Tian, Wenxi; Yang, Zhen; SU, G.H.; Qiu, Suizheng

    2012-01-01

    Highlights: ► A Fuel ROd Behavior Analysis code (FROBA) has been developed. ► The effects irradiation and burnup has been considered in FROBA. ► The comparison with INL’s results shows a good agreement. ► The FROBA code was applied to AP1000. ► Peak fuel temperature, gap width, hoop strain, etc. were obtained. -- Abstract: The reliable prediction of nuclear fuel rod behavior is of great importance for safety evaluation of nuclear reactors. In the present study, a thermo-mechanical coupling code FROBA (Fuel ROd Behavior Analysis) has been independently developed with consideration of irradiation and burnup effects. The thermodynamic, geometrical and mechanical behaviors have been predicted and were compared with the results obtained by Idaho National Laboratory to validate the reliability and accuracy of the FROBA code. The validated code was applied to analyze the fuel behavior of AP1000 at different burnup levels. The thermal results show that the predicted peak fuel temperature experiences three stages in the fuel lifetime. The mechanical results indicate that hoop strain at high power is greater than that at low power, which means that gap closure phenomenon will occur earlier at high power rates. The maximum cladding stress meets the requirement of yield strength limitation in the entire fuel lifetime. All results show that there are enough safety margins for fuel rod behavior of AP1000 at rated operation conditions. The FROBA code is expected to be applied to deal with more complicated fuel rod scenarios after some modifications.

  10. Simulating control rod and fuel assembly motion using moving meshes

    Energy Technology Data Exchange (ETDEWEB)

    Gilbert, D. [Department of Electrical and Computer Engineering, McMaster University, 1280 Main Street West, Hamilton Ontario, L8S 4K1 (Canada)], E-mail: gilbertdw1@gmail.com; Roman, J.E. [Departamento de Sistemas Informaticos y Computacion, Universidad Politecnica de Valencia, Camino de Vera s/n, 46022 Valencia (Spain); Garland, Wm. J. [Department of Engineering Physics, McMaster University, 1280 Main Street West, Hamilton Ontario, L8S 4K1 (Canada); Poehlman, W.F.S. [Department of Computing and Software, McMaster University, 1280 Main Street West, Hamilton Ontario, L8S 4K1 (Canada)

    2008-02-15

    A prerequisite for designing a transient simulation experiment which includes the motion of control and fuel assemblies is the careful verification of a steady state model which computes k{sub eff} versus assembly insertion distance. Previous studies in nuclear engineering have usually approached the problem of the motion of control rods with the use of nonlinear nodal models. Nodal methods employ special approximations for the leading and trailing cells of the moving assemblies to avoid the rod cusping problem which results from the naive volume weighted cell cross-section approximation. A prototype framework called the MOOSE has been developed for modeling moving components in the presence of diffusion phenomena. A linear finite difference model is constructed, solutions for which are computed by SLEPc, a high performance parallel eigenvalue solver. Design techniques for the implementation of a patched non-conformal mesh which links groups of sub-meshes that can move relative to one another are presented. The generation of matrices which represent moving meshes which conserve neutron current at their boundaries, and the performance of the framework when applied to model reactivity insertion experiments is also discussed.

  11. 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

  12. Evaluation of LWR fuel rod behavior under operational transient conditions

    International Nuclear Information System (INIS)

    Nakamura, M.; Hiramoto, K.; Maru, A.

    1984-01-01

    To evaluate the effects of fission gas flow and diffusion in the fuel-cladding gap on fuel rod thermal and mechanical behaviors in light water reactor (LWR) fuel rods under operational transient conditions, computer sub-programs which can calculate the gas flow and diffusion have been developed and integrated into the LWR fuel rod performance code BEAF. This integrated code also calculates transient temperature distribution in the fuel-pellet and cladding. The integrated code was applied to an analysis of Inter Ramp Project data, which showed that by taking into account the gas flow and diffusion effects, the calculated cladding damage indices predicted for the failed rods in the ramp test were consistent with iodine-SCC (Stress Corrosion Cracking) failure conditions which were obtained from out-of-reactor pressurized tube experiments with irradiated Zircaloy claddings. This consistency was not seen if the gas flow and diffusion effects were neglected. Evaluation were also made for the BWR 8x8 RJ fuel rod temperatures under power ramp conditions. (orig.)

  13. The modeling of fuel rod behaviour under RIA conditions in the code DYN3D

    International Nuclear Information System (INIS)

    Rohde, U.

    1998-01-01

    A description of the fuel rod behaviour and heat transfer model used in the code DYN3D for nuclear reactor core dynamic simulations is given. Besides the solution of heat conduction equations in fuel and cladding, the model comprises detailed description of heat transfer in the gas gap by conduction, radiation and fuel-cladding contact. The gas gap behaviour is modeled in a mechanistic way taking into account transient changes of the gas gap parameters based on given conditions for the initial state. Thermal, elastic and plastic deformations of fuel and cladding are taken into account within 1D approximation. Numerical studies concerning the fuel rod behaviour under RIA conditions in power reactors are reported. Fuel rod behaviour at high pressures and flow rates in power reactors is different from the behaviour under atmospheric pressure and stagnant flow conditions in the experiments. The mechanisms of fuel rod failure for fresh and burned fuel reported from the literature can be qualitatively reproduced by the DYN3D model. (author)

  14. Evaluation of the TIG welding mechanical behavior in AISI 316 tubes for fuel rods

    International Nuclear Information System (INIS)

    Bittencourt, M.S.Q.; Carvalho Perdigao, S. de

    1985-10-01

    The effect of service temperature, the mechanical resistance and the creep behaviour of a steel which is intendend to be used as fuel rods in Nuclear Reactors was investigated. The tests were performed in seamless tubes of austenitic stainless steel, AISI 316, 20% cold worked, TIG welded. (Author) [pt

  15. Fuel Rod Consolidation Project: Phase 2, Final report: Volume 1

    International Nuclear Information System (INIS)

    1987-01-01

    This design report describes the NUS final design of the Prototype Spent Nuclear Fuel Rod Consolidation System. This summary presents the approach and the subsequent sections describe, in detail, the final design. Detailed data, drawings, and the design Basis Accident Report are provided in Volumes II thru V. The design as presented, represents one cell of a multicell facility for the dry consolidation of any type of PWR and BWR fuel used in the United States LWR industry that will exceed 1% of the fuel inventory at the year 2000. The system contains the automatically-controlled equipment required to consolidate 750MT (heavy metal)/year, at 75% availability. The equipment is designed as replaceable components using state-of-the-art tchnology. The control system utilizes the most advanced commercially available equipment on the market today. Two state-of-the-art advanced servo manipulators are provided for system maintenance. In general the equipment is designed utilizing fabricated and commercial components. For example, the main drive systems use commercially available roller screws. These rollers screws have 60,000 hours of operation in nuclear power plants and have been used extensively in other applications. The motors selected represent the most advanced designed servo motors on the market today for the precision control of machinery. In areas where precise positioning was not required, less expensive TRW Globe motors were selected. These are small compact motors with a long history of operations in radiation environments. The Robotic Bridge Transporters are modified versions of existing bridge cranes for remote automatic operations. Other equipment such as the welder for fuel canister closure operations is a commercially available product with an operating history applicable to this process. In general, this approach was followed throughout the design of all the equipment and will enable the system to be developed without costly development programs

  16. 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

  17. Characterization of irradiated fuel rods using pulsed eddy current techniques

    International Nuclear Information System (INIS)

    Martin, M.R.; Francis, W.C.

    1975-11-01

    A number of irradiated fuel rods and unfueled zircaloy cladding tubes (''water tubes'') were obtained from the Saxton reactor through arrangements with the Westinghouse Electric Corporation for use in subsequent irradiation effects and fuel behavior programs. A comprehensive nondestructive and corroborative destructive characterization program was undertaken on these fuel rods and tubes by ANC to provide baseline data on their characteristics prior to further testing and for comparison against post-post data. This report deals primarily with one portion of the NDT program performed remotely in the hot cells. The portion of interest in this paper is the pulsed eddy current inspection used in the nondestructive phase of the work. 6 references

  18. Gap conductance in Zircaloy-clad LWR fuel rods

    International Nuclear Information System (INIS)

    Ainscough, J.B.

    1982-04-01

    This report describes the procedures currently used to calculate fuel-cladding gap conductance in light water reactor fuel rods containing pelleted UO 2 in Zircaloy cladding, under both steady-state and transient conditions. The relevant theory is discussed together with some of the approximations usually made in performance modelling codes. The state of the physical property data which are needed for heat transfer calculations is examined and some of the relevant in- and out-of-reactor experimental work on fuel rod conductance is reviewed

  19. Detection of defective fuel rods in water reactors - a review

    International Nuclear Information System (INIS)

    Hartog, J.M.

    1980-01-01

    Consideration of the fundamental processes of fission product release within fuel pellets and at the pellet surface, and its transport in the fuel/cladding interspace and from fuel rod to coolant, indicates what radio-nuclides will be detectable in the coolant from small and large cladding failures. A better understanding of the aggregate fission product transport is required to allow reactor operators to interpret signals from detection systems in terms of quantitative cladding deterioration. This needs experimental investigation in a specially instrumented loop, as well as development of a technique to cause a rod to defect deliberately during steady power operation. (author)

  20. Characteristics of axial splits in failed BWR fuel rods

    International Nuclear Information System (INIS)

    Lysell, G.; Grigoriev, V.

    2000-01-01

    Secondary cladding defects in BWR fuel sometimes have the shape of long axial cracks or ''splits''. Due to the large open UO 2 surfaces exposed to the water, fission product and UO 2 release to the coolant can reach excessive levels leading to forced shut downs to remove the failed fuel rods. A number of such fuel rods have been examined in Studsvik over the last 10 years. The paper describes observations from the PIE of long cracks and discusses the driving force of the cracks. Details such as starting cracks, macroscopic and microscopic fracture surface appearance, cross sections of cracks, hydride precipitates, location and degree of plastic deformation are given. (author)

  1. The thermo-mechanics of the PWR fuel rod

    International Nuclear Information System (INIS)

    Barral, J.C.; Gautier, B.; Chaigne, G.

    1999-01-01

    The fuel rod mechanics is of a great importance in the safety and performance of the reactors. In this domain a meeting has been organized by the SFEN the 18 march 1998 at Paris. With the participation of scientists from CEA, EDF and Framatome, the physics of the fuel rods was presented based on four main aspects. Two first papers dealt with the solicitations of the fuel rod in normal and accidental conditions. The physical phenomena under irradiation were then detailed in the four following talks. Three papers presented the simulation and the codes of the fuel-cladding interactions with the diabolo effect. The last paper was devoted to the experiment feedback and the research programs. (A.L.B.)

  2. The modeling of fuel rod behaviour under RIA conditions in the code DYN3D

    International Nuclear Information System (INIS)

    Rohde, U.

    2001-01-01

    A description of the fuel rod behaviour and heat transfer model used in the code DYN3D for nuclear reactor core dynamic simulations is given. Besides the solution of heat conduction equations in fuel and cladding, the model comprises a detailed description of heat transfer in the gas gap by conduction, radiation and fuel-cladding contact. The gas gap behaviour is modeled in a mechanistic way taking into account transient changes of the gas gap parameters based on given conditions for the initial state. Thermal, elastic and plastic deformations of fuel and cladding are taken into account within 1D approximation. A creeping law for time-dependent estimation of plastic deformations is implemented. Metal-water reaction of the cladding material in the high temperature region is considered. The cladding-coolant heat transfer regime map covers the region from one-phase liquid convection to dispersed flow with superheated steam. Special emphasis is put on taking into account the impact of thermodynamic non-equilibrium conditions on heat transfer. For the validation of the model, experiments on fuel rod behaviour during RIAs carried out in Russian and Japanese pulsed research reactors with shortened probes of fresh fuel rods are calculated. Comparisons between calculated and measured results are shown and discussed. It is shown, that the fuel rod behaviour is significantly influenced by plastic deformation of the cladding, post crisis heat transfer with sub-cooled liquid conditions and heat release from the metal-water reaction. Numerical studies concerning the fuel rod behaviour under RIA conditions in power reactors are reported on. It is demonstrated, that the fuel rod behaviour at high pressures and flow rates in power reactors is different from the behaviour under atmospheric pressure and stagnant flow conditions in the experiments. The mechanisms of fuel rod failure for fresh and burned fuel reported from the literature can be qualitatively reproduced by the DYN3D

  3. Experimental data report for Test TS-2 reactivity initiated accident test in NSRR with pre-irradiated BWR fuel rod

    International Nuclear Information System (INIS)

    Nakamura, Takehiko; Yoshinaga, Makio; Sobajima, Makoto; Fujishiro, Toshio; Kobayashi, Shinsho; Yamahara, Takeshi; Sukegawa, Tomohide; Kikuchi, Teruo

    1993-02-01

    This report presents experimental data for Test TS-2 which was the second test in a series of Reactivity Initiated Accident (RIA) condition test using pre-irradiated BWR fuel rods, performed at the Nuclear Safety Research Reactor (NSRR) in February, 1990. Test fuel rod used in the Test TS-2 was a short sized BWR (7x7) type rod which was fabricated from a commercial rod irradiated at Tsuruga Unit 1 power reactor. The fuel had an initial enrichment of 2.79% and a burnup of 21.3Gwd/tU (bundle average). A pulse irradiation of the test fuel rod was performed under a cooling condition of stagnant water at atmospheric pressure and at ambient temperature which simulated a BWR's cold start-up RIA event. The energy deposition of the fuel rod in this test was evaluated to be 72±5cal/g·fuel (66±5cal/g·fuel in peak fuel enthalpy) and no fuel failure was observed. Descriptions on test conditions, test procedures, transient behavior of the test rod during the pulse irradiation, and, results of pre and post pulse irradiation examinations are described in this report. (author)

  4. Calculation of neutron activation of control rods of a nuclear reactor, using MCNP5

    International Nuclear Information System (INIS)

    Pena V, J.D.

    2016-01-01

    The control rods of a nuclear reactor are activated by neutron irradiation. The generated activity produces a dose around the rod which is irrelevant inside the reactor, but significant when the rod is withdrawn and placed in a storage pool, because this dose is a potential risk to the surrounding personnel. On the other hand, most of the activation occurs in the stainless steel components of the rod. The Monte Carlo model can reliably determine the activation produced in a stainless steel part exposed to a neutron flux in a reactor and the dose measurement around this part. This thesis presents the Monte Carlo models developed for the activation of the control rods of the TRIGA Mark III reactor of Instituto Nacional de Investigaciones Nucleares (ININ) when only standard fuel was available. Therefore, the validations of the Monte Carlo models are reliable. (Author)

  5. Prototypical spent fuel rod consolidation equipment preliminary design report: Volume 1, Report

    International Nuclear Information System (INIS)

    1986-01-01

    This design report describes the NUS Preliminary Design of the Prototype Spent Nuclear Fuel Rod Consolidation Equipment for the Department of Energy. The sections of the report elaborate on each facet of the preliminary design. A concept summary is provided to assist the reader in rapidly understanding the complete design. The NUS Prototype Spent Fuel Rod Consolidation System is an automatically controlled system to consolidate a minimum of 750 MT (heavy metal)/year of US commercial nuclear reactor fuel, at 75% availability. The system is designed with replaceable components utilizing the latest state-of-the-art technology. This approach gives the system the flexibility to be developed without costly development programs, yet accept new technology as it evolves over the next ten years. Capability is also provided in the system design to accommodate a wide variety of fuel conditions and to recover from any situation which may arise

  6. Nondestructive examination of irradiated fuel rods by pulsed eddy current techniques

    International Nuclear Information System (INIS)

    Francis, W.C.; Quapp, W.J.; Martin, M.R.; Gibson, G.W.

    1976-02-01

    A number of fuel rods and unfueled zircaloy cladding tubes which had been irradiated in the Saxton reactor have undergone extensive nondestructive and corroborative destructive examinations by Aerojet Nuclear Company as part of the Water Reactor Safety Research Program, Irradiation Effects Test Series. This report discusses the pulsed eddy current (PEC) nondestructive examinations on the fuel rods and tubing and the metallography results on two fuel rods and one irradiated zircaloy tube. The PEC equipment, designed jointly by Argonne National Laboratory and Aerojet, performed very satisfactorily the functions of diameter, profile, and wall thickness measurements and OD and ID surface defect detection. The destructive examination provided reasonably good confirmation of ''defects'' detected in the nondestructive examination

  7. Fuel rod for use in BWR type reactor

    International Nuclear Information System (INIS)

    Takeuchi, Kiyoshi.

    1989-01-01

    A hollow intermediate end plug is disposed to a plenum portion of a fuel rod and a plenum spring is disposed between the end plug and the upper end of a fuel pellet. Then, a hollow portion is disposed between the intermediate end plug and an upper end plug. Thus, since a only a non exothermic portion is present from the intermediate end plug to the upper end plug, oxidation, corrosion, etc. to the fuel can are not caused so much as in the exothermic portion. Accordingly, the wall thickness of the fuel may be reduced to such a extent as only capable of withstanding the external pressure by coolants and the increasing inner pressure due to the release of FP gases and, accordingly, the wall thickness can be reduced as compared with that of the fuel portion in the fuel can. Further, since the power density per unit length of the fuel rod is reduced for fuels with increased number of fuel rods, it is possible to design so as to reduce the release amount of FP gases thereby decreasing the plenum volume. Further, since the surface area in the coolant phase stream portion is reduced, it can be expected for decreasing the pressure loss of fuels and accompanying effect for improving the channel stability. (T.M.)

  8. Fuel performance computer code simulation of steady-state and transient regimes of the stainless steel fuel rods

    International Nuclear Information System (INIS)

    Gomes, Daniel de Souza

    2014-01-01

    The immediate cause of the accident at the Fukushima Daiichi nuclear plant in March 2011 was the meltdown of the reactor core. During this process, the zirconium cladding of the fuel reacts with water, producing a large amount of hydrogen. This hydrogen, combined with volatile radioactive materials leaked from the containment vessel and entered the building of the reactor, resulting in explosions. In the past, stainless steel was used as the coating in many pressurized water reactors (PWR) under irradiation and their performance was excellent, however, the stainless steel was replaced by a zirconium-based alloy as a coating material mainly due to its lower section shock-absorbing neutrons. Today, the stainless steel finish appears again as a possible solution for security issues related to the explosion and hydrogen production. The objective of this thesis is to discuss the performance under irradiation of fuel rods using stainless steel as a coating material. The results showed that stainless steel rods exhibit lower temperatures and higher fuel pellet width of the gap - coating the coated rods Zircaloy and this gap does not close during the irradiation. The thermal performance of the two fuel rods is very similar, and the penalty of increased absorption of neutrons due to the use of stainless steel can be offset by the combination of a small increase in the enrichment of U- 235 and changes in the size of the spacing between the fuel rods. (author)

  9. Control rod guide tube of nuclear reactor

    International Nuclear Information System (INIS)

    Suda, Yoshitaka; Ito, Kenji; Matsumoto, Kunio.

    1994-01-01

    Zr having a residual tensile stress of 3 to 10kg/mm 2 in a circumferential direction is used for the main ingredient of a control guide tube of a nuclear reactor. For this purpose, an appropriate correction method such as a roll-correction, tension-correction and press-correction method is applied to an existent Zr-base alloy tube with no substantial residual stress. If the residual tensile stress in the circumferential direction is smaller than 3kg/mm 2 , an effect sufficient to suppress irradiation growth is not obtainable, if it exceeds 10kg/mm 2 , dimensional changes, cracks or the like occurs locally since the wall thickness of the control rod guide tube is small and, accordingly, this often results in failed products as the control guide tube. (N.H.)

  10. Status of work on the final repository concept concerning direct disposal of spent fuel rods in fuel rod casks (BSK)

    International Nuclear Information System (INIS)

    Filbert, W.; Wehrmann, J.; Bollingerfehr, W.; Graf, R.; Fopp, S.

    2008-01-01

    The reference concept in Germany on direct final storage of spent fuel rods is the burial of POLLUX containers in the final repository salt dome. The POLLUX container is self-shielded. The final storage concept also includes un-shielded borehole storage of high-level waste and packages of compacted waste. GNS has developed a spent fuel container (BSK-3) for unshielded borehole storage with a mass of 5.2 tons that can carry the fuel rods of three PWR reactors of 9 BWR reactors. The advantages of BSK storage include space saving, faster storage processes, less requirements concerning technical barriers, cost savings for self-shielded casks.

  11. Nuclear fuel assembly

    International Nuclear Information System (INIS)

    Borrman, B.; Nylund, O.

    1984-01-01

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

  12. Prevention of nuclear fuel cladding materials corrosion

    International Nuclear Information System (INIS)

    Yang, K.R.; Yang, J.C.; Lee, I.C.; Kang, H.D.; Cho, S.W.; Whang, C.K.

    1983-01-01

    The only way which could be performed by the operator of nuclear power plant to minimizing the degradation of nuclear fuel cladding material is to control the water quality of primary coolant as specified standard conditions which dose not attack the cladding material. If the water quality of reactor coolant does not meet far from the specification, the failure will occure not only cladding material itself but construction material of primary system which contact with the coolant. The corrosion product of system material are circulate through the whole primary system with the coolant and activated by the neutron near the reactor core. The activated corrosion products and fission products which released from fuel rod to the coolant, so called crud, will repeate deposition and redeposition continuously on the fuel rod and construction material surface. As a result we should consider heat transfer problem. In this study following activities were performed; 1. The crud sample was taken from the spent fuel rod surface of Kori unit one and analized for radioactive element and non radioactive chemical species. 2. The failure mode of nuclear fuel cladding material was estimated by the investigation of releasing type of fission products from the fuel rod to the reactor coolant using the iodine isotopes concentration of reactor coolants. 3. A study was carried out on the sipping test results of spent fuel and a discussion was made on the water quality control records through the past three cycle operation period of Kori unit one plant. (Author)

  13. Nuclear fuel preheating system

    International Nuclear Information System (INIS)

    Andrea, C.

    1975-01-01

    A nuclear reactor new fuel handling system which conveys new fuel from a fuel preparation room into the reactor containment boundary is described. The handling system is provided with a fuel preheating station which is adaptd to heat the new fuel to reactor refueling temperatures in such a way that the fuel is heated from the top down so that fuel element cladding failure due to thermal expansions is avoided. (U.S.)

  14. Nuclear fuel shipping inspection device

    International Nuclear Information System (INIS)

    Takahashi, Toshio; Hada, Koji.

    1988-01-01

    Purpose: To provide an nuclear fuel shipping inspection device having a high detection sensitivity and capable of obtaining highly reliable inspection results. Constitution: The present invention concerns a device for distinguishing a fuel assembly having failed fuel rods in LMFBR type reactors. Coolants in a fuel assembly to be inspected are collected by a sampling pipeway and transferred to a filter device. In the filter device, granular radioactive corrosion products (CP) in the coolants are captured, to reduce the background. The coolants, after being passed through the filter device, are transferred to an FP catching device and gamma-rays of iodine and cesium nuclides are measured in FP radiation measuring device. Subsequently, the coolants transferred to a degasing device to separate rare gas FP in the coolants from the liquid phase. In a case if rare gas fission products are detected by the radiation detector, it means that there is a failed fuel rod in the fuel assembly to be inspected. Since the CP and the soluble FP are separated and extracted for the radioactivity measurement, the reliability can be improved. (Kamimura, M.)

  15. Remote helium leak test of the DUPIC fuel rod

    International Nuclear Information System (INIS)

    Kim, W. K; Kim, S. S.; Lim, S. P.; Lee, J. W.; Yang, M. S.

    1998-01-01

    DUPIC(Direct Use of spent PWR fuel In CANDU reactor) is one of dry reprocessing fuel cycles to reuse irradiated PWR fuel in CANDU power plant. DUPIC fuel is so radioactive that DUPIC fuel is remotely fabricated at hot cell such as IMEF hot cell in which radiation is shielded and remote operation is possible. In this study, Helium leakage has been tested for the simulated DUPIC fuel rod manufactured by Nd:YAG laser end-cap welding at simulated hot cell. The remote inspection technique has been developed to evaluate the soundness of DUPIC fuel fabricated through new processes. Vacuum chamber has been developed to be remotely operated by manipulators at hot cell. As the result of remote test, Helium leakage of DUPIC fuel rod is around background level, CANDU specification has been satisfied. In the result of the study, remote test has been successfully performed at the simulated hot cell, and the soundness of DUPIC fuel rod welded by Nd:YAG laser has been confirmed

  16. Pressure drop ana velocity measurements in KMRR fuel rod bundles

    International Nuclear Information System (INIS)

    Yagn, Sun Kyu; Chung, Heung June; Chung, Chang Whan; Chun, Se Young; Song, Chul Wha; Won, Soon Yeun; Chung, Moon Ki

    1990-01-01

    The detailed hydraulic characteristic measurements in subchannels of longitudinally finned rod bundles using one-component LDV(Laser Doppler Velocimeter) were performed. Time mean axial velocity, turbulent intensity, and turbulent micro scales, such as time auto-correlation, Eulerian integral and micro scale, Kolmogorov length and time scale, and Taylor micro length scale were measured. The signals from LDV are inherently more or less discontinuous. The spectra of signals having such intermittent defects can be obtained by the fast Fourier transformation (FFT) of the auto-correlation function. The turbulent crossflow mixing rate between neighboring subchannels and dominant frequencies were evaluated from the measured data. Pressure drop data were obtained for the typical 36-element and 18-element fuel rod bundles fabricated by the design requirement of KMRR fuel and for other type of fuels assembled with 6-fin rods to investigate the fin effects on the pressure drop characteristics

  17. System and method for consolidating spent fuel rods

    International Nuclear Information System (INIS)

    Baudro, T.O.

    1987-01-01

    A system is described for consolidating spent fuel rods from spent fuel assemblies, comprising: a consolidation container in which the fuel rods may be packed; a frame capable of holding a fuel assembly and the container during consolidation, the frame permitting each of the fuel assembly and the container to be removed; tool means with gripper means for gripping and releasing a rod, the tool means including means for moving the gripper means upwardly and downwardly; a first indexing head having first guide means for guiding the gripper means while the gripper means moves downwardly; a first rail, the first indexing head being slidably mounted on the first rail; a second indexing head having second guide means for guiding the gripper means while the gripper means moves downwardly; a second rail, the second indexing head being slidably mounted on the second rail; and a third rail, the first rail and the second rail being slidably mounted on the third rail; wherein the first indexing head is slidable on the first and third rails to a first position that is above a preselected rod in the fuel assembly; and wherein the second indexing head is slidable on the second and third rails to a second position that is above a preselected location in the container

  18. A system automatic study for the spent fuel rod cutting and simulated fuel pellet extraction device

    International Nuclear Information System (INIS)

    Jeong, J. H.; Yun, J. S.; Hong, D. H.; Kim, Y. H.; Park, K. Y.

    2001-01-01

    A fuel pellet extraction device of the spent fuel rods is described. The device consists of a cutting device of the spent fuel rods and the decladding device of the fuel pellets. The cutting device is to cut a spent fuel rod to n optimal size for fast decladding operation. To design the device, the fuel rod properties are investigated including the dimension and material of fuel rod tubes and pellets. Also, various methods of existing cutting method are investigated. The design concepts accommodate remote operability for the Hot-Cell(radioactive ) area operation. Also, the modularization of the device structure is considered for the easy maintenance. The decladding device is to extract the fuel pellet from the rod cut. To design this device, the existing method is investigated including the chemical and mechanical decladding methods. From the view point of fuel recovery and feasibility of implementation. it is concluded that the chemical decladding method is not appropriate due to the mass production of radioactive liquid wastes, in spite of its high fuel recovery characteristics. Hence, in this paper, the mechanical decladding method is adopted and the device is designed so as to be applicable to various lengths of rod-cuts. As like the cutting device,the concepts of remote operability and maintainability is considered. Both devices are fabricated and the performance is investigated through a series of experiments. From the experimental result, the optimal operational condition of the devices is established

  19. Evolution of fuel rod support under irradiation impact on the mechanical behaviour of fuel assemblies

    International Nuclear Information System (INIS)

    Billerey, Antoine; Waeckel, Nicolas

    2005-01-01

    New fuel management targets imply to increase fuel assembly discharge burnup. Therefore, the prediction of the mechanical behaviour of the irradiated fuel assembly is essential such as excessive fuel assembly distortion induce incomplete Rod Cluster Control Assembly insertion problems (safety issue) or fuel rod vibration induced wear leading to leaking rods (plant operation problems). Within this framework, one of the most important parameter is the knowledge of the fuel rod support in the grid cell because it directly governs the mechanical behaviour of the fuel assembly and consequently allows to predict the behaviour of irradiated structures in terms of (1) axial and lateral deformation (global behaviour of the assembly) and (2) rod vibration induced wear (local behaviour of the rod). Generally, fuel rod support is provided by a spring-dimple system fixed to the grid. During irradiation, the spring force decreases and a gap between the rod and the spring may occur. This phenomenon is due to (1) stress relieving in the spring and in the dimples, (2) grid growth and (3) reduction of the rod diameter. Two models have been developed to predict the behaviour of the rod in the cell. The first model is dedicated to the evaluation of the spring force relaxation during irradiation. The second one can assess the rotation characteristic of the fuel rod in the cell, function of the spring force. The main input parameters are (1) the creep laws of the grid materials, (2) the growth law of the grid, (3) the evolution of rod diameter and (4) the design of the fuel rod support. The aim of this paper is to: (1) evaluate the consequences of grid support design modifications on the rod vibration sensitivity in terms of predicted rod to grid maximum gap during irradiation and time in operation with an open rod to grid gap, (2) evaluate, using a linear or non-linear Finite Element assembly model, the impact of the evolution of grid support under irradiation on the overall mechanical

  20. Computer analysis of elongation of the WWER fuel rod claddings

    International Nuclear Information System (INIS)

    Scheglov, A.; Proselkov, V.

    2008-01-01

    In this paper description of mechanisms influencing changes of the WWER fuel cladding length and axial forces influencing fuel and cladding are presented. It is shown that shortening of the fuel claddings in case of high burnup can be explained by the change of the fuel and cladding reference state caused by reduction of the fuel rod power level - during reactor outages. It is noted that the presented calculated data are to be reviewed and interpreted as the preliminary results; further work is needed for their confirmation. (authors)

  1. Parameter study on the influence of prepressurization on LWR fuel rod behaviour during normal operation and hypothetical LOCA

    International Nuclear Information System (INIS)

    Fuchs, H.P.; Brzoska, B.; Depisch, F.; Sauermann, W.

    1978-01-01

    To analyse the influence of prepressurization on fuel rod behaviour, a parametric study has been performed considering the effects of the as-fabricated fuel rod internal prepressure on the normal operation and postulated LOCA red behaviour of a 1300 MWe1 KWU standard nuclear power plant pressurized water reactor. A reduction of prepressurization in the analysed range results in a negligible worsened normal operation behaviour whereas the LOCA behaviour is improved significantly. (author)

  2. Acoustic sensor for in-pile fuel rod fission gas release measurement

    International Nuclear Information System (INIS)

    Fourmentel, D.; Villard, J. F.; Ferrandis, J. Y.; Augereau, F.; Rosenkrantz, E.; Dierckx, M.

    2009-01-01

    We have developed a specific acoustic sensor to improve the knowledge of fission gas release in Pressurized Water Reactor (PWR) fuel rods when irradiated in materials testing reactors. In order to perform experimental programs related to the study of the fission gas release kinetics, the CEA (French Nuclear Energy Commission) acquired the ability to equip a pre-irradiated PWR fuel rod with three sensors, allowing the simultaneous on-line measurements of the following parameters: - fuel temperature with a centre-line thermocouple type C, - internal pressure with a specific counter-pressure sensor, - fraction of fission gas released in the fuel rod with an innovative acoustic sensor. The third detector is the subject of this paper. This original acoustic sensor has been designed to measure the molar mass and pressure of the gas contained in the fuel rod plenum. For in-pile instrumentation, the fraction of fission gas, such as Krypton and Xenon, in Helium, can be deduced online from this measurement. The principle of this acoustical sensor is the following: a piezoelectric transducer generates acoustic waves in a cavity connected to the fuel rod plenum. The acoustic waves are propagated and reflected in this cavity and then detected by the transducer. The data processing of the signal gives the velocity of the acoustic waves and their amplitude, which can be related respectively to the molar mass and to the pressure of the gas. The piezoelectric material of this sensor has been qualified in nuclear conditions (gamma and neutron radiations). The complete sensor has also been specifically designed to be implemented in materials testing reactors conditions. For this purpose some technical points have been studied in details: - fixing of the piezoelectric sample in a reliable way with a suitable signal transmission, - size of the gas cavity to avoid any perturbation of the acoustic waves, - miniaturization of the sensor because of narrow in-pile experimental devices

  3. Romanian nuclear fuel program

    International Nuclear Information System (INIS)

    Budan, O.

    1999-01-01

    The paper presents and comments the policy adopted in Romania for the production of CANDU-6 nuclear fuel before and after 1990. The CANDU-6 nuclear fuel manufacturing started in Romania in December 1983. Neither AECL nor any Canadian nuclear fuel manufacturer were involved in the Romanian industrial nuclear fuel production before 1990. After January 1990, the new created Romanian Electricity Authority (RENEL) assumed the responsibility for the Romanian Nuclear Power Program. It was RENEL's decision to stop, in June 1990, the nuclear fuel production at the Institute for Nuclear Power Reactors (IRNE) Pitesti. This decision was justified by the Canadian specialists team findings, revealed during a general, but well enough technically founded analysis performed at IRNE in the spring of 1990. All fuel manufactured before June 1990 was quarantined as it was considered of suspect quality. By that time more than 31,000 fuel bundles had already been manufactured. This fuel was stored for subsequent assessment. The paper explains the reasons which provoked this decision. The paper also presents the strategy adopted by RENEL after 1990 regarding the Romanian Nuclear Fuel Program. After a complex program done by Romanian and Canadian partners, in November 1994, AECL issued a temporary certification for the Romanian nuclear fuel plant. During the demonstration manufacturing run, as an essential milestone for the qualification of the Romanian fuel supplier for CANDU-6 reactors, 202 fuel bundles were produced. Of these fuel bundles, 66 were part of the Cernavoda NGS Unit 1 first fuel load (the balance was supplied by Zircatec Precision Industries Inc. ZPI). The industrial nuclear fuel fabrication re-started in Romania in January 1995 under AECL's periodical monitoring. In December 1995, AECL issued a permanent certificate, stating the Romanian nuclear fuel plant as a qualified and authorised CANDU-6 fuel supplier. The re-loading of the Cernavoda NGS Unit 1 started in the middle

  4. Fuel rod failure due to marked diametral expansion and fuel rod collapse occurred in the HBWR power ramp experiment

    International Nuclear Information System (INIS)

    Yanagisawa, Kazuaki

    1985-12-01

    In the power ramp experiment with the BWR type light water loop at the HBWR, the two pre-irradiated fuel rods caused an unexpected pellet-cladding interaction (PCI). One occurred in the fuel rod with small gap of 0.10 mm, which was pre-irradiated up to the burn-up of 14 MWd/kgU. At high power, the diameter of the rod was increased markedly without accompanying significant axial elongation. The other occurred in the rod with a large gap of 0.23 mm, which was pre-irradiated up to the burn-up of 8 MWd/kgU. The diameter of the rod collapsed during a diameter measurement at the maximum power level. The causes of those were investigated in the present study by evaluating in-core data obtained from equipped instruments in the experiment. It was revealed from the investigation that these behaviours were attributed to the local reduction of the coolant flow occurred in the region of a transformer in the ramp rig. The fuel cladding material is seemed to become softened due to temperature increase caused by the local reduction of the coolant flow, and collapsed by the coolant pressure, either locally or wholly depending on the rod diametral gap existed. (author)

  5. Isotopic Details of the Spent Catawba-1 MOX Fuel Rods at ORNL

    Energy Technology Data Exchange (ETDEWEB)

    Ellis, Ronald James [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-04-01

    The United States Department of Energy funded Shaw/AREVA MOX Services LLC to fabricate four MOX Lead Test Assemblies (LTA) from weapons-grade plutonium. A total of four MOX LTAs (including MX03) were irradiated in the Catawba Nuclear Station (Unit 1) Catawba-1 PWR which operated at a total thermal power of 3411 MWt and had a core with 193 total fuel assemblies. The MOX LTAs were irradiated along with Duke Energy s irradiation of eight Westinghouse Next Generation Fuel (NGF) LEU LTAs (ref.1) and the remaining 181 LEU fuel assemblies. The MX03 LTA was irradiated in the Catawba-1 PWR core (refs.2,3) during cycles C-16 and C-17. C-16 began on June 5, 2005, and ended on November 11, 2006, after 499 effective full power days (EFPDs). C-17 started on December 29, 2006, (after a shutdown of 48 days) and continued for 485 EFPDs. The MX03 and three other MOX LTAs (and other fuel assemblies) were discharged at the end of C-17 on May 3, 2008. The design of the MOX LTAs was based on the (Framatome ANP, Inc.) Mark-BW/MOX1 17 17 fuel assembly design (refs. 4,5,6) for use in Westinghouse PWRs, but with MOX fuel rods with three Pu loading ranges: the nominal Pu loadings are 4.94 wt%, 3.30 wt%, and 2.40 wt%, respectively, for high, medium, and low Pu content. The Mark-BW/MOX1 (MOX LTA) fuel assembly design is the same as the Advanced Mark-BW fuel assembly design but with the LEU fuel rods replaced by MOX fuel rods (ref. 5). The fabrication of the fuel pellets and fuel rods for the MOX LTAs was performed at the Cadarache facility in France, with the fabrication of the LTAs performed at the MELOX facility, also in France.

  6. Dealing with Historical Discrepancies: The Recovery of National Research Experiment (NRX) Reactor Fuel Rods at Chalk River Laboratories (CRL) - 13324

    International Nuclear Information System (INIS)

    Vickerd, Meggan

    2013-01-01

    Following the 1952 National Research Experiment (NRX) Reactor accident, fuel rods which had short irradiation histories were 'temporarily' buried in wooden boxes at the 'disposal grounds' during the cleanup effort. The Nuclear Legacy Liabilities Program (NLLP), funded by Natural Resources Canada (NRCan), strategically retrieves legacy waste and restores lands affected by Atomic Energy of Canada Limited (AECL) early operations. Thus under this program the recovery of still buried NRX reactor fuel rods and their relocation to modern fuel storage was identified as a priority. A suspect inventory of NRX fuels was compiled from historical records and various research activities. Site characterization in 2005 verified the physical location of the fuel rods and determined the wooden boxes they were buried in had degraded such that the fuel rods were in direct contact with the soil. The fuel rods were recovered and transferred to a modern fuel storage facility in 2007. Recovered identification tags and measured radiation fields were used to identify the inventory of these fuels. During the retrieval activity, a discrepancy was discovered between the anticipated number of fuel rods and the number found during the retrieval. A total of 32 fuel rods and cans of cut end pieces were recovered from the specified site, which was greater than the anticipated 19 fuel rods and cans. This discovery delayed the completion of the project, increased the associated costs, and required more than anticipated storage space in the modern fuel storage facility. A number of lessons learned were identified following completion of this project, the most significant of which was the potential for discrepancies within the historical records. Historical discrepancies are more likely to be resolved by comprehensive historical record searches and site characterizations. It was also recommended that a complete review of the wastes generated, and the total affected lands as a result of this historic

  7. Nuclear fuel lease accounting

    International Nuclear Information System (INIS)

    Danielson, A.H.

    1986-01-01

    The subject of nuclear fuel lease accounting is a controversial one that has received much attention over the years. This has occurred during a period when increasing numbers of utilities, seeking alternatives to traditional financing methods, have turned to leasing their nuclear fuel inventories. The purpose of this paper is to examine the current accounting treatment of nuclear fuel leases as prescribed by the Financial Accounting Standards Board (FASB) and the Federal Energy Regulatory Commission's (FERC's) Uniform System of Accounts. Cost accounting for leased nuclear fuel during the fuel cycle is also discussed

  8. Nuclear reactor shutdown control rod assembly

    International Nuclear Information System (INIS)

    Bilibin, K.

    1988-01-01

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

  9. SFAK, Unscattered Gamma Self-Absorption from Regular Fuel Rod Assemblies

    International Nuclear Information System (INIS)

    Wand, H.

    1982-01-01

    1 - Description of problem or function: Calculation of the self- absorption of unscattered (gamma-) radiation from fuel assemblies which contain a regular arrangement of identical fuel rods. 2 - Method of solution: The point-kernel is integrated over the radiation sources, i.e. the fuel rods. A uniform mesh of integration points is used for each of the fuel rods. 3 - Restrictions on the complexity of the problem: Number of fuel rods is dynamically allocated

  10. The irradiation performance of austenitic stainless steel clade PWR fuel rods

    International Nuclear Information System (INIS)

    Teixeira e Silva, A.; Esteves, A.M.

    1988-01-01

    The steady state irradiation performance of austenitic stainless steel clad pressurized water reactor fuel rods is modeled with fuel performance codes of the FRAP series. These codes, originally developed to model the thermal-mechanical behavior of zircaloy clad fuel rods, are modified to model stainless steel clad fuel rods. The irradiation thermal-mechanical behavior of type 348 stainless steel and zircaloy fuel rods is compared. (author) [pt

  11. The Nuclear Fuel Cycle

    International Nuclear Information System (INIS)

    2011-08-01

    This brochure describes the nuclear fuel cycle, which is an industrial process involving various activities to produce electricity from uranium in nuclear power reactors. The cycle starts with the mining of uranium and ends with the disposal of nuclear waste. The raw material for today's nuclear fuel is uranium. It must be processed through a series of steps to produce an efficient fuel for generating electricity. Used fuel also needs to be taken care of for reuse and disposal. The nuclear fuel cycle includes the 'front end', i.e. preparation of the fuel, the 'service period' in which fuel is used during reactor operation to generate electricity, and the 'back end', i.e. the safe management of spent nuclear fuel including reprocessing and reuse and disposal. If spent fuel is not reprocessed, the fuel cycle is referred to as an 'open' or 'once-through' fuel cycle; if spent fuel is reprocessed, and partly reused, it is referred to as a 'closed' nuclear fuel cycle.

  12. Nuclear fuel elements

    International Nuclear Information System (INIS)

    Nakai, Keiichi

    1983-01-01

    Purpose: To decrease the tensile stresses resulted in a fuel can as well as prevent decladding of fuel pellets into the bore holes by decreasing the inner pressure within the nuclear fuel element. Constitution: A fuel can is filled with hollow fuel pellets, inserted with a spring for retaining the hollow fuel pellets with an appropriate force and, thereafter, closely sealed at the both ends with end plugs. A cylindrical body is disposed into the bore holes of the hollow fuel pellets. Since initial sealing gases and/or gaseous nuclear fission products can thus be excluded from the bore holes where the temperature is at the highest level, the inner pressure of the nuclear fuel element can be reduced to decrease the tensile strength resulted to the fuel can. Furthermore, decladding of fuel pellets into the bore holes can be prevented. (Moriyama, K.)

  13. Nuclear fuel replacement device

    International Nuclear Information System (INIS)

    Ritz, W.C.; Robey, R.M.; Wett, J.F.

    1984-01-01

    A fuel handling arrangement for a liquid metal cooled nuclear reactor having a single rotating plug eccentric to the fuel core and a fuel handling machine radially movable along a slot in the plug with a transfer station disposed outside the fuel core but covered by the eccentric plug and within range of movement of said fuel handling machine to permit transfer of fuel assemblies between the core and the transfer station. (author)

  14. FARST: A computer code for the evaluation of FBR fuel rod behavior under steady-state/transient conditions

    International Nuclear Information System (INIS)

    Nakamura, M.; Sakagami, M.

    1984-01-01

    FARST, a computer code for the evaluation of fuel rod thermal and mechanical behavior under steady-state/transient conditions has been developed. The code characteristics are summarized as follows: (I) FARST evaluates the fuel rod behavior under the transient conditions. The code analyzes thermal and mechanical phenomena within a fuel rod, taking into account the temperature change in coolant surrounding the fuel rod. (II) Permanent strains such as plastic, creep and swelling strains as well as thermoelastic deformations can be analyzed by using the strain increment method. (III) Axial force and contact pressure which act on the fuel stack and cladding are analyzed based on the stick/slip conditions. (IV) FARST used a pellet swelling model which depends on the contact pressure between pellet and cladding, and an empirical pellet relocation model, designated as 'jump relocation model'. The code was successfully applied to analyses of the fuel rod irradiation data from pulse reactor for nuclear safety research in Cadarache (CABRI) and pulse reactor for nuclear safety research in Japan Atomic Energy Research Institute (NSRR). The code was further applied to stress analysis of a 1000 MW class large FBR plant fuel rod during transient conditions. The steady-state model which was used so far gave the conservative results for cladding stress during overpower transient, but underestimated the results for cladding stress during a rapid temperature decrease of coolant. (orig.)

  15. Development of mechanical analysis module for simulation of SFR fuel rod behavior using finite element method

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Andong; Jeong, Hyedong; Suh, Namduk [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of); Kim, Hyochan; Yang, Yongsik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    Korean SFR developer decided to adapt metal fuel, current study focused on the metal fuel instead of oxide fuel. The SFR metal fuel has been developed by Korea Atomic Energy Research Institute (KAERI) and many efforts focused on designing and manufacturing the metal fuel. Since a nuclear fuel is the first barrier to protect radioactive isotope release, the fuel's integrity must be secured during steady-state operation and accident condition within an acceptable range. Whereas the design and evaluation methodologies, code systems and test procedures of a light water reactor fuel are sufficiently established, those of the SFR fuel needs more technical advances. In the view of regulatory point, there are still many challenging issues which are required to secure the safety of fuel and reactors. For this reason, the Korea Institute of Nuclear Safety (KINS) has launched the new project to develop the regulatory technology for SFR system including a fuel area. The ALFUS code was developed by CRIEPI and employs mechanistic model for fission gas release and swelling of fuel slug. In the code system, a finite element method was introduced to analyze the fuel and cladding's mechanical behaviors. The FEAST code is more advanced code system for SFR which adopted mechanistic FGR and swelling model but still use analytical model to simulate fuel and cladding mechanical behavior. Based on the survey of the previous studies, fuel and cladding mechanical model should be improved. Analysis of mechanical behavior for fuel rod is crucial to evaluate overall rod's integrity. In addition, it is because contact between fuel slug and cladding or an over-pressure of rod internal pressure can cause rod failure during steady-state and other operation condition. The most of reference codes have simplified mechanical analysis model, so called 'analytical mode', because the detailed mechanical analysis requires large amount of calculation time and computing power. Even

  16. Development of mechanical analysis module for simulation of SFR fuel rod behavior using finite element method

    International Nuclear Information System (INIS)

    Shin, Andong; Jeong, Hyedong; Suh, Namduk; Kim, Hyochan; Yang, Yongsik

    2014-01-01

    Korean SFR developer decided to adapt metal fuel, current study focused on the metal fuel instead of oxide fuel. The SFR metal fuel has been developed by Korea Atomic Energy Research Institute (KAERI) and many efforts focused on designing and manufacturing the metal fuel. Since a nuclear fuel is the first barrier to protect radioactive isotope release, the fuel's integrity must be secured during steady-state operation and accident condition within an acceptable range. Whereas the design and evaluation methodologies, code systems and test procedures of a light water reactor fuel are sufficiently established, those of the SFR fuel needs more technical advances. In the view of regulatory point, there are still many challenging issues which are required to secure the safety of fuel and reactors. For this reason, the Korea Institute of Nuclear Safety (KINS) has launched the new project to develop the regulatory technology for SFR system including a fuel area. The ALFUS code was developed by CRIEPI and employs mechanistic model for fission gas release and swelling of fuel slug. In the code system, a finite element method was introduced to analyze the fuel and cladding's mechanical behaviors. The FEAST code is more advanced code system for SFR which adopted mechanistic FGR and swelling model but still use analytical model to simulate fuel and cladding mechanical behavior. Based on the survey of the previous studies, fuel and cladding mechanical model should be improved. Analysis of mechanical behavior for fuel rod is crucial to evaluate overall rod's integrity. In addition, it is because contact between fuel slug and cladding or an over-pressure of rod internal pressure can cause rod failure during steady-state and other operation condition. The most of reference codes have simplified mechanical analysis model, so called 'analytical mode', because the detailed mechanical analysis requires large amount of calculation time and computing power. Even

  17. Mechanical behaviour of PWR fuel rods during intermediate storage

    International Nuclear Information System (INIS)

    Bouffioux, P.; Dalmas, R.; Bernaudat, C.

    2000-01-01

    EDF, which owns the irradiated fuel coming from its NPPs, has initiated studies regarding the mechanical behaviour of a fuel rod and the integrity of its cladding, in the case where the spent fuel is stored for a significant duration. During the phases following in-reactor irradiation (ageing in a water-pool, transport and intermediate storage), many phenomena, which are strongly coupled, may influence the cladding integrity: - residual power and temperature decay; - helium production and release in the free volume of the rod (especially for MOX fuel); - fuel column swelling; - cladding creep-out under the inner gas pressure of the fuel rod; - metallurgical changes due to high temperatures during transportation. In parallel, the quantification of the radiological risk is based on the definition of a cladding integrity criterion. Up to now, this criterion requires that the clad hoop strain due to creep-out does not exceed 1%. A more accurate criterion is being investigated. The study and modelling of all the phenomena mentioned above are included in a R and D programme. This programme also aims at redefining the cladding integrity criterion, which is assumed to be too conservative. The R and D programme will be presented. In order to predict the overall behaviour of the rod during the intermediate storage phases, the AVACYC code has been developed. It includes the models developed in the R and D programme. The input data of the AVACYC code are provided by the results of in-reactor rod behaviour simulations, using the thermal-mechanical CYRANO3 code. Its main results are the evolution vs. time of hoop stresses in the cladding, rod internal pressure and cladding hoop strains. Chained CYRANO-AVACYC calculations have been used to simulate the behaviour of MOX fuel rods irradiated up to 40 GWd/t and stored under air during 100 years, or under water during 50 years. For such fuels, where the residual power remains high, we show that a large part of the cladding strain

  18. Development of techniques for joining fuel rod simulators to test assemblies

    International Nuclear Information System (INIS)

    Moorhead, A.J.; Reed, R.W.

    1980-01-01

    A unique tubular electrode carrier is described for gas tungsten-arc welding small-diameter nuclear fuel rod simulators to the tubesheet of a test assembly. Both the close-packed geometry of the array of simulators and the extension of coaxial electrical conductors from each simulator hindered access to the weld joint. Consequently, a conventional gas tungsten-arc torch could not be used. Two seven-rod assemblies that were mockups of the simulator-to-tubesheet joint area were welded and successfully tested. Modified versions of the electrode carrier for brazing electrical leads to the upper ends of the fuel pin simulators are also described. Satisfactory brazes have been made on both single-rod mockups and an array of 25 simulators by using the modified electrode carrier and a filler metal with a composition of 71.5 Ag-28 Cu-0.5 Ni

  19. Transient fuel rod behavior prediction with RODEX-3/SIERRA

    Energy Technology Data Exchange (ETDEWEB)

    Billaux, M R; Shann, S H; Swam, L.F. Van [Siemens Power Corp., Richland, WA (United States)

    1997-08-01

    This paper discusses some aspects of the fuel performance code SIERRA (SIEmens Rod Response Analysis). SIERRA, the latest version of the code RODEX-3, has been developed to improve the fuel performance prediction capabilities of the code, both at high burnup and during transient reactor conditions. The paper emphasizes the importance of the mechanical models of the cracked pellet and of the cladding, in the prediction of the transient response of the fuel rod to power changes. These models are discussed in detail. Other aspects of the modelling of high burnup effects are also presented, in particular the modelling of the rim effect and the way it affects the fuel temperature. (author). 12 refs, 5 figs.

  20. Reactor core with rod-shaped fuel cells

    International Nuclear Information System (INIS)

    Dworak, A.

    1975-01-01

    Power distribution in a high-temperature gas-cooled reactor is optimized. Especially the axial as well as the radial power distribution is kept constant, the core consisting of several consecutive rod-shaped fuel cells. To this end, the dwell times of the fuel cells are fitted to the given power distribution. Fuel cells with equal dwell times, seen in flow direction, are arranged side by side, and those with the shortest dwell times are placed in areas with the greatest power release. These areas ly on the coolant inlet side. To keep the power distribution constant, fuel cells with neutron poison or absorber rods with absorbing rates decreasing in flow direction can also be inserted. (RW/PB) [de

  1. Transient fuel rod behavior prediction with RODEX-3/SIERRA

    International Nuclear Information System (INIS)

    Billaux, M.R.; Shann, S.H.; Swam, L.F. Van

    1997-01-01

    This paper discusses some aspects of the fuel performance code SIERRA (SIEmens Rod Response Analysis). SIERRA, the latest version of the code RODEX-3, has been developed to improve the fuel performance prediction capabilities of the code, both at high burnup and during transient reactor conditions. The paper emphasizes the importance of the mechanical models of the cracked pellet and of the cladding, in the prediction of the transient response of the fuel rod to power changes. These models are discussed in detail. Other aspects of the modelling of high burnup effects are also presented, in particular the modelling of the rim effect and the way it affects the fuel temperature. (author). 12 refs, 5 figs

  2. Advanced Fuel Pellet Materials and Fuel Rod Design for Water Cooled Reactors. Proceedings of a Technical Committee Meeting

    International Nuclear Information System (INIS)

    2010-10-01

    The economics of current nuclear power plants have improved through increased fuel burnup and longer fuel cycles, i.e. increasing the effective time that fuel remains in the reactor core and the amount of energy it generates. Efficient consumption of fissile material in the fuel element before it is discharged from the reactor means that less fuel is required over the reactor's life cycle, which results in lower amounts of fresh fuel, lower spent fuel storage costs, and less waste for ultimate disposal. Better utilization of fissile nuclear materials, as well as more flexible power manoeuvring, place challenging operational demands on materials used in reactor components, and first of all, on fuel and cladding materials. It entails increased attention to measures ensuring desired in-pile fuel performance parameters that require adequate improvements in fuel material properties and fuel rod designs. These are the main reasons that motivated the IAEA Technical Working Group on Fuel Performance and Technology (TWG-FPT) to recommend the organization of a Technical Committee Meeting on Advanced Fuel Pellet Materials and Fuel Rod Designs for Power Reactors. The proposal was supported by the IAEA TWGs on Advanced Technologies for Light and Heavy Water-Cooled Reactors (TWG-LWR and TWG-HWR), and the meeting was held at the invitation of the Government of Switzerland at the Paul Scherrer Institute in Villigen, from 23 to 26 November 2009. This was the third IAEA meeting on these subjects (the first was held in 1996 in Tokyo, Japan, and the second in 2003 in Brussels, Belgium), which reflects the continuous interest in the above issues among Member States. The purpose of the meeting was to review the current status in the development of fuel pellet materials and to explore recent improvements in fuel rod designs for light and heavy water cooled power reactors. The meeting was attended by 45 specialists representing fuel vendors, nuclear utilities, research and development

  3. Parametric study of fuel rod behaviour during the RIA using the modified FALCON code

    International Nuclear Information System (INIS)

    Khvostov, G.; Zimmermann, M.A.; Ledergerber, G.

    2010-01-01

    Presented in the paper are the results of a parametric study with the use of optimised modules of the FALCON code (FALCON-PSI) that addresses the effects of the selected characteristics of fast thermal transients (e.g., impulse width), fuel rod design (e.g., active fuel attack length) and boundary conditions (e.g., the coolant conditions) on fuel behaviour during a RIA. Specifically, the analysis of the governing processes for the fuel rod behaviour during the RIA events simulated in the experimental facility of the Nuclear Safety Research Reactor (NSRR, Japan) are in the focus of the present study. The results obtained can be useful for a better transfer of the NSRR test results in relation to the corresponding behaviour in LWRs and furthermore might also support the planning of future additional experiments. (authors)

  4. Absorber rod drive for nuclear reactors

    International Nuclear Information System (INIS)

    Acher, H.

    1985-01-01

    The invention concerns a further addition to the invention of DE 33 42 830 A1. The free contact of the hollow piston with the nut due to hydraulic pressure is replaced by a hydraulic or spring attachment. The pressure system required to produce the hydraulic pressure is therefore omitted, and the electrical power required for driving the pump or the mass flow is also omitted. The absorber rod slotted along its longitudinal axis is replaced by an absorber rod, in the longitudinal axis of which a hollow piston is connected together with the absorber rod. This makes the absorber rod more stable, and assembly is simplified. (orig./HP) [de

  5. CONDOR: neutronic code for fuel elements calculation with rods

    International Nuclear Information System (INIS)

    Villarino, E.A.

    1990-01-01

    CONDOR neutronic code is used for the calculation of fuel elements formed by fuel rods. The method employed to obtain the neutronic flux is that of collision probabilities in a multigroup scheme on two-dimensional geometry. This code utilizes new calculation algorithms and normalization of such collision probabilities. Burn-up calculations can be made before the alternative of applying variational methods for response flux calculations or those corresponding to collision normalization. (Author) [es

  6. Pellet clad interaction analysis of AFA 3G fuel rod

    International Nuclear Information System (INIS)

    Liu Tong; Shen Caifen; Jiao Yongjun; Lu Huaquan; Zhou Zhou

    2002-01-01

    The author described Pellet Clad Interaction (PCI) analysis of AFA 3G fuel rod during condition II transients for GNPS 18-months alternating equilibrium cycles. It provided PCI technical limit, analytical methods and computer code used in the analyses of condition II transients and thermal-mechanical. Finally, given main calculation results and the conclusion for GNPS 18-months cycles

  7. Estimation and control in HTGR fuel rod fabrication

    International Nuclear Information System (INIS)

    Downing, D.J.; Bailey, J.M.

    1980-01-01

    A control algorithm has been derived for an HTGR Fuel Rod Fabrication Process utilizing the method of G.E.P. Box and G.M. Jenkins. The estimator is a Kalman filter and is compared with a Least Square estimator and a standard control chart. The effects of system delays are presented. 1 ref

  8. Estimation and control in HTGR fuel rod fabrication

    International Nuclear Information System (INIS)

    Downing, D.J.; Bailey, M.J.

    1980-01-01

    A control algorithm has been derived for a HTGR Fuel Rod Fabrication Process utilizing the method of Box and Jenkins. The estimator is a Kalman filter and is compared with a Least Square estimator and a standard control chart. The effects of system delays are presented

  9. Development of the spent fuel rod cutting device using the blade cutters

    International Nuclear Information System (INIS)

    Jung, Jae Hoo; Yoon, Ji Sup; Hong, Dong Hee; Kim, Young Hwan; Park, Gee Yong; Kim, Do Woo

    2000-11-01

    A spent fuel rod cutting device is to cut a spent nuclear fuel rod to optimal size for consequent decladding operation. In this paper, various properties of fuel rod, such as a dimension and material of zircaloy tubes and fuel pellets, are investigated. Also, commercially available cutting method and tools is investigated in terms of its performance. As a result, the blade cutter is selected for the design. In order to fabricate the durable blade cutter, various materials are analyzed in terms of material properties, cutter shape, and heat treatment method, etc. Also, the durability of this tool is tested by cutting the SUS tubes and zircaloy tubes. In the device design, the remote maintainability is considered so that the modularized design is accomplished. Also, the other factors considered in the design are the round shape sustainability at the cut surface, the amount of debris generation, and the fire risk, etc. Considering these design consideration, the spent fuel rod cutting device is fabricated and tested

  10. Nuclear fuel accounting

    International Nuclear Information System (INIS)

    Aisch, D.E.

    1977-01-01

    After a nuclear power plant has started commercial operation the actual nuclear fuel costs have to be demonstrated in the rate making procedure. For this purpose an accounting system has to be developed which comprises the following features: 1) All costs associated with nuclear fuel shall be correctly recorded; 2) it shall be sufficiently flexible to cover also deviations from proposed core loading patterns; 3) it shall be applicable to different fuel cycle schemes. (orig./RW) [de

  11. The nuclear fuel cycle

    International Nuclear Information System (INIS)

    1998-05-01

    After a short introduction about nuclear power in the world, fission physics and the French nuclear power plants, this brochure describes in a digest way the different steps of the nuclear fuel cycle: uranium prospecting, mining activity, processing of uranium ores and production of uranium concentrates (yellow cake), uranium chemistry (conversion of the yellow cake into uranium hexafluoride), fabrication of nuclear fuels, use of fuels, reprocessing of spent fuels (uranium, plutonium and fission products), recycling of energetic materials, and storage of radioactive wastes. (J.S.)

  12. Experimental data report for Test TS-1 Reactivity Initiated Accident Test in NSRR with pre-irradiated BWR fuel rod

    International Nuclear Information System (INIS)

    Nakamura, Takehiko; Yoshinaga, Makio; Sobajima, Makoto; Fujishiro, Toshio; Horiki, Ohichiro; Yamahara, Takeshi; Ichihashi, Yoshinori; Kikuchi, Teruo

    1992-01-01

    This report presents experimental data for Test TS-1 which was the first in a series of tests, simulating Reactivity Initiated Accident (RIA) conditions using pre-irradiated BWR fuel rods, performed in the Nuclear Safety Research Reactor (NSRR) in October, 1989. Test fuel rod used in the Test TS-1 was a short-sized BWR (7 x 7) type rod which was fabricated from a commercial rod provided from Tsuruga Unit 1 power reactor. The fuel had an initial enrichment of 2.79 % and burnup of 21.3 GWd/t (bundle average). Pulse irradiation was performed at a condition of stagnant water cooling, atmospheric pressure and ambient temperature using a newly developed double container-type capsule. Energy deposition of the rod in this test was evaluated to be about 61 cal/g·fuel (55 cal/g·fuel in peak fuel enthalpy) and no fuel failure was observed. Descriptions on test conditions, test procedures, fuel burnup measurements, transient behavior of the test rod during pulse irradiation and results of post pulse irradiation examinations are contained in this report. (author)

  13. IFPE/TRIBULATION R1, Fuel Rod Behaviour at High Burnup

    International Nuclear Information System (INIS)

    Turnbull, J.A.

    2002-01-01

    Description: The TRIBULATION (Tests Relative to High Burnup Limitations Arising Normally in LWRs) International Programme started in July 1980 and was organized jointly by BelgoNucleaire and the Nuclear Energy Centre at Mol (CEN/SCK) with the co-sponsorship of 14 participating organizations. The objectives of the programme were twofold. It was primarily a demonstration programme aimed at assessing the fuel rod behaviour at high burn-up, when an earlier transient had occurred in the power plant. The second objective was to investigate the behaviour of different fuel rod designs and manufacturers when subjected to a steady state irradiation history to high burn-up. The first objective was met by irradiating fuel rods under steady state conditions in the BR3 reactor and under transient conditions in BR2. The effect of the transient was determined by comparing data from 4 identical rods tested as follows: i) BR3 irradiation followed by PIE; ii) BR3 irradiation followed by BR2 transient then PIE; iii) BR3 irradiation followed by BR2 transient and re-irradiated in BR3 before PIE; iv) BR3 irradiation and continued BR3 irradiation to maximum burn-up before PIE. The Database contains data from 19 cases using rods fabricated by BelgoNucleaire (BN) (11) and Brown Boveri Reactor GmbH (BBR) (8)

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

    International Nuclear Information System (INIS)

    Jabsen, F.S.

    1978-01-01

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

  15. Nuclear reactor with scrammable part length rod

    International Nuclear Information System (INIS)

    Bevilacqua, F.

    1979-01-01

    A new part length rod is provided. It may be used to control xenon induced power oscillations but to contribute to shutdown reactivity when a rapid shutdown of the reactor is required. The part length rod consists of a control rod with three regions. The lower control region is a longer weaker active portion separated from an upper stronger shorter poison section by an intermediate section which is a relative non-absorber of neutrons. The combination of the longer weaker control section with the upper high worth poison section permits the part length rod of this to be scrammed into the core when a reactor shutdown is required but also permits the control rod to be used as a tool to control power distribution in both the axial and radial directions during normal operation

  16. Analysis of transient heat conduction in a PWR fuel rod by an improved lumped parameter approach

    Energy Technology Data Exchange (ETDEWEB)

    Dourado, Eneida Regina G. [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil); Cotta, Renato M. [Coordenacao de Pos-Graduacao e Pesquisa de Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Mecanica; Jian, Su, E-mail: eneidadourado@gmail.com, E-mail: sujian@nuclear.ufrj.br, E-mail: cotta@mecanica.ufrj.br [Coordenacao de Pos-Graduacao e Pesquisa de Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear

    2017-07-01

    This paper aims to analyze transient heat conduction in a nuclear fuel rod by an improved lumped parameter approach. One-dimensional transient heat conduction is considered, with the circumferential symmetry assumed and the axial conduction neglected. The thermal conductivity and specific heat in the fuel pellet are considered temperature dependent, while the thermophysical properties of the cladding are considered constant. Hermite approximation for integration is used to obtain the average temperature and heat flux in the radial direction. Significant improvement over the classical lumped parameter formulation has been achieved. The proposed model can be also used in dynamic analysis of PWR and nuclear power plant simulators. (author)

  17. Analysis of transient heat conduction in a PWR fuel rod by an improved lumped parameter approach

    International Nuclear Information System (INIS)

    Dourado, Eneida Regina G.; Cotta, Renato M.; Jian, Su

    2017-01-01

    This paper aims to analyze transient heat conduction in a nuclear fuel rod by an improved lumped parameter approach. One-dimensional transient heat conduction is considered, with the circumferential symmetry assumed and the axial conduction neglected. The thermal conductivity and specific heat in the fuel pellet are considered temperature dependent, while the thermophysical properties of the cladding are considered constant. Hermite approximation for integration is used to obtain the average temperature and heat flux in the radial direction. Significant improvement over the classical lumped parameter formulation has been achieved. The proposed model can be also used in dynamic analysis of PWR and nuclear power plant simulators. (author)

  18. LOCA scenario tests of irradiated fuel rod specimens

    International Nuclear Information System (INIS)

    Scott, Harold

    2004-01-01

    Full text: The NRC's cladding performance program at Argonne National Laboratory (ANL) is testing fueled high-burnup segments subjected to LOCA integral phenomena. The data are provided to NRC and the nuclear industry for their independent assessment of the adequacy of licensing criteria for LOCA events. The tests are being conducted with high-burnup 30 cm segments from Limerick (9x9 Zry-2) and H.B. Robinson (15x15 Zry-4) reactors. Prior to testing, sibling samples are characterized with respect to fuel morphology, fuel-cladding bond, cladding oxide layer thickness, hydrogen content and high-temperature steam oxidation kinetics. Specimens that survive quench are subjected to four-point bend tests, followed by local diametral compression tests. The retention of post-quench ductility is a more limiting requirement than surviving thermal stresses during quench. Companion tests are conducted with unirradiated cladding to generate baseline data for comparison with the high-burnup fuel results. LOCA integral tests have the following sequential steps: stabilization of temperature, internal pressure and steam flow at 300 C, ramping of temperature (∼5C/s) through ballooning and burst to 1204 C, hold at 1204 C for 1-5 minutes, slow-cooling (∼3C/s) to 800 C, and water quenching at ∼800C. Two high-burnup tests were completed in 2002 with Limerick BWR rod segments: ramp to burst in argon followed by slow cooling; and the LOCA test with 5-minute hold time at 1204 C, followed by slow cooling. With the exception of burst-opening shape, results for burst temperature, burst pressure, burst length, and ballooning strain profile are more similar to, than different from, results for unirradiated Zry-2 cladding exposed to the same time-temperature history. The 3rd Limerick test with quench was performed in December 2003, and a 4th Limerick test was performed in March 2004. Tests on high-burnup Robinson PWR fuel segments are scheduled to begin in June 2004. The presentation points

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

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  20. Evaluation of the thermal-mechanical performance of fuel rods of a BWR during a power ramp using the FUELSIM code

    International Nuclear Information System (INIS)

    Pantoja C, R.

    2010-01-01

    To avoid the risk to environment due to release of radioactive material, because of occurrence of an accident, it is the priority of the design and performance of the diverse systems of safety of a commercial nuclear power plant. The safety of nuclear power plants requires, therefore, monitoring those parameters having some direct or indirect effect on safety. The thermal limits are values set for those parameters considered having most impact on the safe operation of a nuclear power reactor. Some thermal limits monitoring requires the thermal-mechanical analysis of the rods containing the nuclear fuel. The fuel rod thermal-mechanical behavior under irradiation is a complex process in which there exists a great deal of interrelated physical and chemical phenomena, so that the fuel rod performance analysis in the core of a nuclear power reactor is generally accomplished by using computer codes, which integrate several of the phenomena that are expected to occur during the lifetime of the fuel rod in the core. The main application of the thermal-mechanical analysis codes is the prediction of occurrence of conditions and/or phenomena that could lead to the deterioration or even mechanical failure of the fuel rod cladding, as, for example, the pellet-cladding interaction. In the operation of a nuclear power reactor, fuel preconditioning operations refer to the operational procedures employed to reduce the fuel rod failure probability due to fuel-cladding interaction, specially during reactor startup. Preconditioning simulations are therefore necessary to determine in advance limit values for the power that can be generated in a fuel rod, and thus avoiding any rod damage. In this work, a first analysis of the thermal-mechanical performance of typical fuel rods used in nuclear reactors of the type BWR 5/6, as those two nuclear reactors in Laguna Verde, Veracruz, is performed. This study includes two types of fuel rods: one from a fuel assembly design with an array 8 x 8

  1. Development of examination technique for oxide layer thickness measurement of irradiated fuel rods

    International Nuclear Information System (INIS)

    Koo, D. S.; Park, S. W.; Kim, J. H.; Seo, H. S.; Min, D. K.; Kim, E. K.; Chun, Y. B.; Bang, K. S.

    1999-06-01

    Technique for oxide layer thickness measurement of irradiated fuel rods was developed to measure oxide layer thickness and study characteristic of fuel rods. Oxide layer thickness of irradiated fuels were measured, analyzed. Outer oxide layer thickness of 3 cycle-irradiated fuel rods were 20 - 30 μm, inner oxide layer thickness 0 - 10 μm and inner oxide layer thickness on cracked cladding about 30 μm. Oxide layer thickness of 4 cycle-irradiated fuel rods were about 2 times as thick as those of 1 cycle-irradiated fuel rods. Oxide layer on lower region of irradiated fuel rods was thin and oxide layer from lower region to upper region indicated gradual increase in thickness. Oxide layer thickness from 2500 to 3000 mm showed maximum and oxide layer thickness from 3000 to top region of irradiated fuel rods showed decreasing trend. Inner oxide layer thicknesses of 4 cycle-irradiated fuel rod were about 8 μm at 750 - 3500 mm from the bottom end of fuel rod. Outer oxide layer thickness were about 8 μm at 750 - 1000 mm from the bottom end of fuel rod. These indicated gradual increase up to upper region from the bottom end of fuel rod. These indicated gradual increase up to upper region from the bottom end of fuel. Oxide layer thickness technique will apply safety evaluation and study of reactor fuels. (author). 6 refs., 14 figs

  2. Test requirement for PIE of HANARO irradiated fuel rod

    International Nuclear Information System (INIS)

    Lim, I. C.; Cho, Y. G.

    2000-06-01

    Since the first criticality of HANARO reached in Feb. of 1995, the rod type U 3 Si-A1 fuel imported from AECL has been used. From the under-water fuel inspection which has been conducted since 1997, a ballooning-rupture type abnormality was observed in several fuel rods. In order to find the root cause of this abnormality and to find the resolution, the post irradiation examination(PIE) was proposed as the best way. In this document, the information from the under-water inspection as well as the PIE requirements are described. Based on the information in this document, a detail test plan will be developed by the project team who shall conduct the PIE

  3. Process and equipment for locating defective fuel rods of a reactor fuel element

    International Nuclear Information System (INIS)

    Jester, A.; Honig, H.

    1977-01-01

    By this equipment, well-known processes for determining defective fuel rods of a reactor fuel element are improved in such a fashion that defective fuel rods can be located individually, so that it is possible to replace them. The equipment consists of a cylindrical test vessel open above, which accommodates the element to be tested, so that an annular space is left between the latter's external circumference and the wall of the vessel, and so that the fuel rods project above the vessel. A bell in the shape of a frustrum of a cone is inverted over the test vessel, which has an infra-red measuring equipment at a certain distance above the tops of the fuel rods. The fuel element to be tested together with the test vessel and hood are immersed in a basin full of water, which displaces water by means of gas from the hood. The post-shutdown heat increases the temperature in the water space of the test vessel, which is stabilised at 100 0 C. In each defective fuel rod the water which has penetrated the defective fuel rod previously, or does so now, starts to boil. The steam rising in the fuel rod raises the temperature of the defective fuel rod compared to all the sound ones. The subsequent measurement easily determines this. Where one can expect interference with the measurement by appreciable amounts of gamma rays, the measuring equipment is removed from the path of radiation by mirror deflection in a suitably shaped measuring hood. (FW) [de

  4. Fuel rod with axial regions of annular and standard fuel pellets

    International Nuclear Information System (INIS)

    Freeman, T.R.

    1991-01-01

    This patent describes a fuel rod for use in a nuclear reactor fuel assembly. It comprises: an elongated hollow cladding tube; a pair of end plugs connected to and sealing the cladding tube at opposite ends of thereof; and an axial stack of fuel pellets contained in and extending between the end plugs at the opposite ends of the tube, all of the fuel pellets contained in the tube being composed of fissile material being enriched above the level of natural enrichment; the fuel pellets in the stack thereof being provided in an arrangement of axial regions. The arrangement of axial regions including a pair of first axial regions defined respectively at the opposite ends of the pellet stack adjacent to the respective end plugs. The pellets in the first axial regions being identical in number and having annular configurations with an annulus of a first void size. The arrangement of axial regions also including another axial region defined between the first axial regions, some of the pellets in the another axial region having solid configurations

  5. Simulation of vibration modes of the fuel rod damaged due to the grid-to-rod fretting wear

    International Nuclear Information System (INIS)

    Kim, Kyu Tae; Kim, Kyeong Koo; Jang, Young Ki; Lee, Kyou Seok

    1997-01-01

    The flow-induced fuel fretting wear observed in some PWRs mainly proceeds in the grid-to-rod contact positions. The grid-to-rod fretting wear in the PWR fuel assembly depends on grid-to-rod gap size, its axial profile and flow-induced vibration. This paper describes the GRIDFORCE program which generates the axially dependent grid-to-rod gap size as a function of burnup. The axially dependent grid-to-rod gap profiles are employed to predict the fuel rod vibration mode shapes by the ANSYS code. With the help of the Paidousis empirical formula, this paper also calculates the fuel rod vibration amplitudes under various supporting conditions, which indicates that the increase of the number of unsupported mid-grids will increase the fuel rod vibration amplitude. On the other hand, the comparison of the predicted vibration mode shapes and the observed mid-grid fretting wear pattern indicates that the 1st and 6th vibration mode shapes under the supporting inactive condition at the mid-grids can simulate the observed mid-grid fretting wear profile. This paper also proposes design guidelines against the grid-to-rod fretting wear. (author). 3 refs., 8 figs

  6. Nuclear fuel production

    International Nuclear Information System (INIS)

    Randol, A.G.

    1985-01-01

    The production of new fuel for a power plant reactor and its disposition following discharge from the power plant is usually referred to as the ''nuclear fuel cycle.'' The processing of fuel is cyclic in nature since sometime during a power plant's operation old or ''depleted'' fuel must be removed and new fuel inserted. For light water reactors this step typically occurs once every 12-18 months. Since the time required for mining of the raw ore to recovery of reusable fuel materials from discharged materials can span up to 8 years, the management of fuel to assure continuous power plant operation requires simultaneous handling of various aspects of several fuel cycles, for example, material is being mined for fuel to be inserted in a power plant 2 years into the future at the same time fuel is being reprocessed from a discharge 5 years prior. Important aspects of each step in the fuel production process are discussed

  7. Nuclear fuel element

    International Nuclear Information System (INIS)

    Mogard, J.H.

    1977-01-01

    A nuclear fuel element is disclosed for use in power producing nuclear reactors, comprising a plurality of axially aligned ceramic cylindrical fuel bodies of the sintered type, and a cladding tube of metal or metal alloys, wherein said cladding tube on its cylindrical inner surface is provided with a plurality of slightly protruding spacing elements distributed over said inner surface

  8. Control rod position detector for nuclear reactor

    International Nuclear Information System (INIS)

    Kudo, Mitsuru; Fujiwara, Hiroshi.

    1981-01-01

    Purpose: To improve the reliability of a control rod position detector by detecting a reactive code with a combination of control rod position change signals produced from vertical and horizontal axis decoders, generation an error signal and thus simultaneously detecting the operation of more than two lead switches. Constitution: Horizontal and vertical axis position signals responsive to changes in the control rod position are applied from lead switches connected in a predetermined matrix connection corresponding to the notches of the positions of respective position detecting probes, the reactive output from the decoder is detected by a reactive code detecting circuit, which in turn generates a fault signal, and the control rod position code converted in a notch number generating circuit is converted to a predetermined value indicating invalidity. Accordingly, a fault caused by the simultaneous operation of a plurality of failed lead switches can be effectively detected. (Yoshino, Y.)

  9. 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)

  10. Comparison of thermal behavior of different PWR fuel rod simulators for LOCA experiments

    International Nuclear Information System (INIS)

    Casal, V.; Malang, S.; Rust, K.

    1982-10-01

    For experimental investigations of a loss-of-coolant accident (LOCA) of a PWR electrical heater rods are applied as thermal fuel rod simulators. To substitute heater rods from the SEMISCALE program by INTERATOM-KfK heater rods in a current experimental program at the Instituut for Energiteknikk-(OECD-Halden), the thermodynamic behavior of different heater rods during a LOCA were compared. The results show, that SEMISCALE-heater rods can be replaced by those fabricated by INTERATOM. (orig.) [de

  11. Investigation of water films on fuel rods in boiling water reactors using neutron tomography

    International Nuclear Information System (INIS)

    Lanthen, Jonas

    2006-09-01

    In a boiling water reactor, thin films of liquid water around the fuel rods play a very important role in cooling the fuel, and evaporation of the film can lead to fuel damage. If the thickness of the water film could be measured accurately the reactor operation could be both safer and more economical. In this thesis, the possibility to use neutron tomography, to study thin water films on fuel rods in an experimental nuclear fuel set-up, has been investigated. The main tool for this has been a computer simulation software. The simulations have shown that very thin water films, down to around 20 pm, can be seen on fuel rods in an experimental set-up using neutron tomography. The spatial resolution needed to obtain this result is around 300 pm. A suitable detector system for this kind of experiment would be plastic fiber scintillators combined with a CCD camera. As a neutron source it would be possible to use a D-D neutron generator, which generates neutrons with energies of 2.5 MeV. Using a neutron generator with a high enough neutron yield and a detector with high enough detection efficiency, a neutron tomography to measure thin water films should take no longer than 25 - 30 minutes

  12. Investigation of water films on fuel rods in boiling water reactors using neutron tomography

    Energy Technology Data Exchange (ETDEWEB)

    Lanthen, Jonas

    2006-09-15

    In a boiling water reactor, thin films of liquid water around the fuel rods play a very important role in cooling the fuel, and evaporation of the film can lead to fuel damage. If the thickness of the water film could be measured accurately the reactor operation could be both safer and more economical. In this thesis, the possibility to use neutron tomography, to study thin water films on fuel rods in an experimental nuclear fuel set-up, has been investigated. The main tool for this has been a computer simulation software. The simulations have shown that very thin water films, down to around 20 pm, can be seen on fuel rods in an experimental set-up using neutron tomography. The spatial resolution needed to obtain this result is around 300 pm. A suitable detector system for this kind of experiment would be plastic fiber scintillators combined with a CCD camera. As a neutron source it would be possible to use a D-D neutron generator, which generates neutrons with energies of 2.5 MeV. Using a neutron generator with a high enough neutron yield and a detector with high enough detection efficiency, a neutron tomography to measure thin water films should take no longer than 25 - 30 minutes.

  13. CFD modeling of secondary flows in fuel rod bundles

    International Nuclear Information System (INIS)

    Baglietto, Emilio; Ninokata, Hisashi

    2004-01-01

    An optimized non-linear eddy viscosity model is introduced, for calculations of detailed coolant velocity distribution in a tight lattice fuel bundle. The low Reynolds formulation has been optimized based on DNS data for channel flow. The non-linear stress-strain relationship has been modified in the coefficients to model the flow anisotropy, which causes the formation of turbulence driven secondary flows inside the bundle subchannels. Predictions of the model are first compared to experimental measurements of secondary flows in a triangularly arrayed rod bundle with p/d=1.3. Subsequently wall shear stress and velocity predictions are compared with different experimental data for a rod bundle with p/d=1.17. The model shows to be able to correctly reproduce the scale of the secondary motion, and to accurately reproduce both wall shear stress and velocity distributions inside the rod bundle subchannels. (author)

  14. Nuclear fuel cycle

    International Nuclear Information System (INIS)

    1993-01-01

    Status of different nuclear fuel cycle phases in 1992 is discussed including the following issues: uranium exploration, resources, supply and demand, production, market prices, conversion, enrichment; reactor fuel technology; spent fuel management, as well as trends of these phases development up to the year 2010. 10 refs, 11 figs, 15 tabs

  15. Nuclear reactor fuel assembly

    International Nuclear Information System (INIS)

    Sasaki, Y.; Tashima, J.

    1975-01-01

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

  16. 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)

  17. Dismantling method for nuclear fuel assembly

    International Nuclear Information System (INIS)

    Yamazaki, Shuji; Kato, Akihiro; Yoshida, Masafumi.

    1993-01-01

    An upper nozzle is detached from a control rod guide tube and an instrumentation tube. Subsequently, slots (slits) having a predetermined width are formed longitudinally at enlarged diameter portions of the control rod guide tube and the instrumentation tube. Then, the control rod guide tube and the instrumentation tube are separated from a lower nozzle, and pulled out from the lattice space of each of the support lattices. Thereafter, a predetermined key is inserted to a key insertion window formed at each of the support lattices, to distort a spring and take the fuel rod out of the lattice space of each of the support lattices. With such procedures, when the control rod guide tube and the instrumentation tube are pulled out of the lattice space of the support lattice, the enlarged diameter portion is narrowed to reduce the diameter, thereby enabling to take them out easily. Accordingly, since the space for inserting the key can be ensured, the nuclear fuel assemblies can easily be dismantled. In addition, fuel rods can be taken out smoothly and in an intact state. (I.N.)

  18. TRANSURANUS: A fuel rod analysis code ready for use

    Energy Technology Data Exchange (ETDEWEB)

    Lassmann, K; O` Carroll, C; Van de Laar, J [Commission of the European Communities, Karlsruhe (Germany). European Inst. for Transuranium Elements; Ott, C [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

    1994-12-31

    The basic concepts of fuel rod performance codes are discussed. The TRANSURANUS code developed at the Institute for Transuranium Elements, Karlsruhe (GE) is presented. It is a quasi two-dimensional (1{sub 1/2}-D) code designed for treatment of a whole fuel rod for any type of reactor and any situation. The fuel rods found in the majority of test- or power reactors can be analyzed for very different situations (normal, off-normal and accidental). The time scale of the problems to be treated may range from milliseconds to years. The TRANSURANUS code consists of a clearly defined mechanical/mathematical framework into which physical models can easily be incorporated. This framework has been extensively tested and the programming very clearly reflects this structure. The code is well structured and easy to understand. It has a comprehensive material data bank for different fuels, claddings, coolants and their properties. The code can be employed in a deterministic and a statistical version. It is written in standard FORTRAN 77. The code system includes: 2 preprocessor programs (MAKROH and AXORDER) for setting up new data cases; the post-processor URPLOT for plotting all important quantities as a function of the radius, the axial coordinate or the time; the post-processor URSTART evaluating statistical analyses. The TRANSURANUS code exhibits short running times. A new WINDOWS-based interactive interface is under development. The code is now in use in various European institutions and is available to all interested parties. 7 figs., 15 refs.

  19. Treatment of wastes from a central spent-fuel rod consolidation facility

    International Nuclear Information System (INIS)

    Ross, W.A.

    1986-01-01

    The consolidation of commercial spent-fuel rods at a central treatment facility (such as the proposed Monitored Retrievable Storage Facility) will generate several types of waste, which may require treatment and disposal. Eight alternatives for the treatment of the wastes have been evaluated as part of DOE's Nuclear Waste Treatment Program at the Pacific Northwest Laboratory. The evaluation considered the system costs, potential waste form requirements, and processing characteristics

  20. Nuclear fuel storage facility

    International Nuclear Information System (INIS)

    Matsumoto, Takashi; Isaka, Shinji.

    1987-01-01

    Purpose: To increase the spent fuel storage capacity and reduce the installation cost in a nuclear fuel storage facility. Constitution: Fuels handled in the nuclear fuel storage device of the present invention include the following four types: (1) fresh fuels, (2) 100 % reactor core charged fuels, (3) spent fuels just after taking out and (4) fuels after a certain period (for example one half-year) from taking out of the reactor. Reactivity is high for the fuels (1), and some of fuels (2), while low in the fuels (3) (4), Source intensity is strong for the fuels (3) and some of the fuels (2), while it is low for the fuels (1) and (4). Taking notice of the fact that the reactivity, radioactive source intensity and generated after heat are different in the respective fuels, the size of the pool and the storage capacity are increased by the divided storage control. While on the other hand, since the division is made in one identical pool, the control method becomes important, and the working range is restricted by means of a template, interlock, etc., the operation mode of the handling machine is divided into four, etc. for preventing errors. (Kamimura, M.)

  1. Study on flow-induced vibration of the fuel rod in HTTR

    International Nuclear Information System (INIS)

    Takase, Kazuyuki

    1988-03-01

    This study was performed in order to investigate flow-induced vibration characteristics of a fuel rod in HTTR (High Temperature engineering Test Reactor) from both an experiment and a numerical simulation. Two kinds of fuel rods were used in this experiment: one was a graphite rod which simulated a specification of the HTTR's fuel rod and the other was an aluminum rod whose weight was a half of the graphite one. The experiment was carried out up to Re = 31000 using air at room temperature and pressure. Air flowed downstream in an annular passage which consisted of the fuel rod and the graphite channel. Numerical simulations by fluid and frequency equations were also carried out. Numerical and experimental results were then compared. The following conclusions were drived: (1) The fuel rod amplitudes increase with the flow rate and with a decrease of the fuel rod weight. (2) The fuel rod amplitudes are obtained by δ/De = 2.22 x 10 -10 Re 1.43 , 9000 ≤ Re ≤ 31000, where δ is a vibration amplitude, De is a hydraulic diameter and Reis Reynolds number. (3) The fuel rod frequencies shift from lower natural frequency to higher as the flow rate increases. (4) The flow-induced vibration behavior of the fuel rod can simulate well by simultaneous equations which used the turbulence model for fluid and the mass model for vibration of the fuel rod. (author)

  2. A probabilistic design method for LMFBR fuel rods

    International Nuclear Information System (INIS)

    Peck, S.O.; Lovejoy, W.S.

    1977-01-01

    Fuel rod performance analyses for design purposes are dependent upon material properties, dimensions, and loads that are statistical in nature. Conventional design practice accounts for the uncertainties in relevant parameters by designing to a 'safety factor', set so as to assure safe operation. Arbitrary assignment of these safety factors, based upon a number of 'worst case' assumptions, may result in costly over-design. Probabilistic design methods provide a systematic way to reflect the uncertainties in design parameters. PECS-III is a computer code which employs Monte Carlo techniques to generate the probability density and distribution functions for time-to-failure and cumulative damage for sealed plenum LMFBR fuel rods on a single rod or whole core basis. In Monte Carlo analyses, a deterministic model (that maps single-valued inputs into single-valued outputs) is coupled to a statistical 'driver'. Uncertainties in the input are reflected by assigning probability densities to the input parameters. Dependent input variables are considered multivariate normal. Independent input variables may be arbitrarily distributed. Sample values are drawn from these input densities, and a complete analysis is done by the deterministic model to generate a sample point in the output distribution. This process is repeated many times, and the number of times each output value occurs is accumulated. The probability that some measure of rod performance will fall within given limits is estimated by the relative frequency with which the Monte Carlo samples fall within tho

  3. Reactor core with rod-shaped fuel cells

    International Nuclear Information System (INIS)

    Dworak, A.

    1976-01-01

    The proposal refers to the optimization of the power distribution in a reactor core which is provided with several successive rod-shaped fuel cells. A uniform power output - especially in radial direction - is aimed at. This is achieved by variation of the dwelling periods of the fuel cells, which have, for this purpose, a fuel mixture changing from layer to layer. The fuel cells with the shortest dwelling period are arranged near the coolant inlet side of the reactor core. The dwelling periods of the fuel cells are adapted to the given power distribution. As neighboring cells have equal dwelling periods, the exchange can be performed much easier then with the composition currently known. (UWI) [de

  4. DANCOFF-3, Dancoff Correction for Cylindrical Fuel Rod at H2O Gaps and for Fuel Clusters

    International Nuclear Information System (INIS)

    Sauer, A.

    1989-01-01

    1 - Nature of physical problem solved: Calculation of the Dancoff correction for cylindrical fuel rods in square and hexagonal infinite lattices, for fuel element rods near water gaps, and for fuel rod clusters. 2 - Method of solution: Evaluation by direct numerical integration over the moderator region. 3 - Restrictions on the complexity of the problem: For every rod arrangement at most 100 cases with different materials cross- sections

  5. Thermoacoustic sensor for nuclear fuel temperaturemonitoring and heat transfer enhancement

    Energy Technology Data Exchange (ETDEWEB)

    James A. Smith; Dale K. Kotter; Randall A. Alli; Steven L. Garrett

    2013-05-01

    A new acoustical sensing system for the nuclear power industry has been developed at The Pennsylvania State University in collaboration with Idaho National Laboratories. This sensor uses the high temperatures of nuclear fuel to convert a nuclear fuel rod into a standing-wave thermoacoustic engine. When a standing wave is generated, the sound wave within the fuel rod will be propagated, by acoustic radiation, through the cooling fluid within the reactor or spent fuel pool and can be monitored a remote location external to the reactor. The frequency of the sound can be correlated to an effective temperature of either the fuel or the surrounding coolant. We will present results for a thermoacoustic resonator built into a Nitonic-60 (stainless steel) fuel rod that requires only one passive component and no heat exchangers.

  6. Nuclear Fuels: Present and Future

    Directory of Open Access Journals (Sweden)

    Donald R. Olander

    2009-02-01

    Full Text Available The important new developments in nuclear fuels and their problems are reviewed and compared with the status of present light-water reactor fuels. The limitations of these fuels and the reactors they power are reviewed with respect to important recent concerns, namely provision of outlet coolant temperatures high enough for use in H2 production, destruction of plutonium to eliminate proliferation concerns, and burning of the minor actinides to reduce the waste repository heat load and long-term radiation hazard. In addition to current oxide-based fuel-rod designs, the hydride fuel with liquid metal thermal bonding of the fuel-cladding gap is covered. Finally, two of the most promising Generation IV reactor concepts, the Very High Temperature Reactor and the Sodium Fast Reactor, and the accompanying reprocessing technologies, aqueous-based UREX and pyrometallurgical, are summarized. In all of the topics covered, the thermodynamics involved in the material's behavior under irradiation and in the reprocessing schemes are emphasized.

  7. Nuclear fuel pellet loading machine

    International Nuclear Information System (INIS)

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

    1975-01-01

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

  8. Cladding tube of fuel rod for a BWR type reactor

    International Nuclear Information System (INIS)

    Nakayama, Hitoshi; Fujie, Kunio; Kuwahara, Heikichi; Hirai, Tadamasa; Kakizaki, Kimio.

    1976-01-01

    Object: To form a cladding tube wall with tunnels in communication with the exterior through a number of small-diameter openings to rapidly disperse a large quantity of heat thereby providing high density of the fuel rod. Structure: Tunnels adjacent to each other are provided under the skin in contact with cooling liquid of a cladding tube, and a number of openings through which said tunnels and the periphery of the cladding tube are placed in communication are formed, said openings each having its section smaller than that of said tunnel. With this arrangement, the cooling water entered the tunnel through some of small diameter openings absorbs heat of the fuel rod to be vaporized, which is flown out into the cooling water through the other small diameter openings and formed into vapor bubbles which move up for release of heat. (Taniai, N.)

  9. Probabilistic fuel rod analyses using the TRANSURANUS code

    Energy Technology Data Exchange (ETDEWEB)

    Lassmann, K; O` Carroll, C; Laar, J Van De [CEC Joint Research Centre, Karlsruhe (Germany)

    1997-08-01

    After more than 25 years of fuel rod modelling research, the basic concepts are well established and the limitations of the specific approaches are known. However, the widely used mechanistic approach leads in many cases to discrepancies between theoretical predictions and experimental evidence indicating that models are not exact and that some of the physical processes encountered are of stochastic nature. To better understand uncertainties and their consequences, the mechanistic approach must therefore be augmented by statistical analyses. In the present paper the basic probabilistic methods are briefly discussed. Two such probabilistic approaches are included in the fuel rod performance code TRANSURANUS: the Monte Carlo method and the Numerical Noise Analysis. These two techniques are compared and their capabilities are demonstrated. (author). 12 refs, 4 figs, 2 tabs.

  10. Study of development of non-destructive method for determining FGR from high burned PWR type fuel rod

    International Nuclear Information System (INIS)

    Yanagisawa, Kazuaki; Miyanishi, Hideyuki; Kitagawa, Isamu; Iida, Shozo; Ito, Tadaharu; Amano, Hidetoshi.

    1991-11-01

    Experimental study was made to evaluate the FGR (Fission Product Gas Release) from high burned PWR type fuel rods by means of non-destructive method through measurement of the gamma activity of 85 Kr isotope which was accumulated in the fuel top plenum. Experimental result shows that it is possible to know the amounts of FGR at fuel plenum by the equations given in the followings. FGR = 0.28C/V f or FGR = 0.07C where, FGR (%) is the amounts of Xe and Kr released from UO 2 fuel, C (counts/h) the radioactivity of 85 Kr at plenum of the tested fuel rod and V f (ml) the plenum volume of the tested fuel rod, respectively. The present study was made by using 14 x 14 PWR type fuel rods preirradiated up to the burn-up of 42.1 MWd/kgU, followed by the pulse irradiation at Nuclear Safety Research Reactor of Japan Atomic Energy Research Institute (JAERI). The FGR of the tested segmented fuel rods were measured by puncturing and found to range from 0.6% to 12% according to the magnitude of the deposited energy given by pulse. Estimated experimental error bands against the above equations were within plus minus 30%. (author)

  11. Study and modeling of fluctuating fluid forces exerted on fuel rods in pressurized water reactors

    International Nuclear Information System (INIS)

    Bhattacharjee, Saptarshi

    2016-01-01

    Flow-induced vibrations in a pressurized water reactor (PWR) core can cause fretting wear in the fuel rods. Due to friction, wear occurs at the contact locations between the spacer grid and the fuel rod. This could compromise the first safety barrier of the nuclear reactor by damaging the fuel rod cladding. In order to ensure the integrity of the cladding, it is necessary to know the random fluctuating forces acting on the rods. However, the spectra for these fluid forces are not well known. The goal of this PhD thesis was to use simple geometrical elements to check the reproducibility of realistic pressurized water reactor spacer grids. As a first step, large eddy simulations were performed on a concentric annular pipe for different mesh refinements using the CFD code Trio CFD (previously Trio U) developed by CEA. A mesh sensitivity study was performed to obtain an acceptable mesh for reproducing standard literature results. This information on mesh resolution was used when carrying out simulations using various geometric obstacles inside the pipe, namely, mixing vanes, circular spacer grid and a combination of square spacer grid with mixing vanes. The last of the three configurations is the closest to a realistic PWR fuel assembly. Structured mesh was generated for the annular pipe case and circular grid case. An innovative hybrid mesh was used for the two remaining cases of the mixing vanes and the square grid: keeping unstructured mesh around the obstacles and structured mesh in the rest of the domain. The inner wall of the domain was representative of the fuel rod cladding. Both hydraulic and wall pressure characteristics were analyzed for each case. The results for the square grid case were found to be an approximate combination of the mixing vane case and circular grid case. Simulation results were compared with experiments performed at CEA Cadarache. Some preliminary comparisons were also made with classical semi-empirical models. (author) [fr

  12. Nuclear fuel activities in Belgium

    Energy Technology Data Exchange (ETDEWEB)

    Bairiot, H

    1997-12-01

    In his presentation on nuclear fuel activities in belgium the author considers the following directions of this work: fuel fabrication, NPP operation, fuel performance, research and development programmes.

  13. Preliminary Content Evaluation of the North Anna High Burn-Up Sister Fuel Rod Segments for Transportation in the 10-160B and NAC-LWT

    Energy Technology Data Exchange (ETDEWEB)

    Ketusky, E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-08-09

    The U.S. Department of Energy’s (DOE’s) Used Fuel Disposition Campaign (UFDC) Program has transported high-burnup nuclear sister fuel rods from a commercial nuclear power plant for purposes of evaluation and testing. The evaluation and testing of high-burnup used nuclear fuel is integral to DOE initiatives to collect information useful in determining the integrity of fuel cladding for future safe transportation of the fuel, and for determining the effects of aging, on the integrity of UNF subjected to extended storage and subsequent transportation. The UFDC Program, in collaboration with the U.S. Nuclear Regulatory Commission and the commercial nuclear industry, has obtained individual used nuclear fuel rods for testing. The rods have been received at Oak Ridge National Laboratory (ORNL) for both separate effects testing (SET) and small-scale testing (SST). To meet the research objectives, testing on multiple 6 inch fuel rod pins cut from the rods at ORNL will be performed at Pacific Northwest National Laboratory (PNNL). Up to 10 rod equivalents will be shipped. Options were evaluated for multiple shipments using the 10-160B (based on 4.5 rod equivalents) and a single shipment using the NAC-LWT. Based on the original INL/Virginia Power transfer agreement, the rods are assumed to 152 inches in length with a 0.374-inch diameter. This report provides a preliminary content evaluation for use of the 10-160B and NAC-LWT for transporting those fuel rod pins from ORNL to PNNL. This report documents the acceptability of using these packagings to transport the fuel segments from ORNL to PNNL based on the following evaluations: enrichment, A2 evaluation, Pu-239 FGE evaluation, heat load, shielding (both gamma and neutron), and content weight/structural evaluation.

  14. Nuclear reactor internals and control rod handling device

    International Nuclear Information System (INIS)

    Betancourt, G.N.; Etzel, W.W.

    1981-01-01

    A method and apparatus for removing, in an essentially continuous operation, the control rods and the upper guide structure from a nuclear reactor vessel during refueling. The apparatus includes a rigid frame which is secured to the upper guide structure after the vessel head is removed. A platform is vertically reciprocable within the frame and is adapted to engage and lift simultaneously all control rod drive shafts to a maximum elevation within the frame. A mechanical interface between the platform and the frame is provided so that continuation of the lifting force on the platform transfers the lift force to the frame whereby the upper guide structure is lifted out of the vessel. Automatically operated stop means are provided to lock the platform and rods in the maximum elevation within the frame in order to prevent accidental dropping of the rods during transfer of the upper guide structure and control rods to a temporary storage area

  15. A nuclear reactor with buffered control rods

    International Nuclear Information System (INIS)

    Bevilacqua, F.

    1974-01-01

    The control rods for, e.g., water-cooled reactors are fastened as units on common crossbars in vertical downward direction. The fastening on the crossbar is achieved by means of cross-shaped parts, e.g., in the shape of a double 'H'. A cylinder connected with a drive rod in normal operation is joined to each of the crossbars. In an emergency shut-down, this connection is interrupted and the control rod unit drops into the core through the action of gravity. Its fall is slowed down by a cushion or shock absorbing unit. For this purpose a piston is provided mounted on the supporting plate below the cylinder and guided within it. In the cylinder, the coolant is contained as damping medium. An upper opening in the cylinder serves as a ventilation hole. The movement of the piston is limited by a stopping part within the cylinder and slowed down by a spiral spring. (DG) [de

  16. Fuel rod failure during film boiling (PCM-1 test in the PBF)

    International Nuclear Information System (INIS)

    Domenico, W.F.; Stanley, C.J.; Mehner, A.S.

    1978-01-01

    The Power-Cooling-Mismatch (PCM) Test, PCM-1 was conducted in the Power Burst Facility (PFB) in March of 1978. The PCM Test Series is being conducted at the Idaho National Engineering Laboratory by EG and G Idaho, Inc., under contract to the USNRC and is designed to characterize the behavior of nuclear fuel rods operating under conditions of high power or low coolant flow or both leading to departure from nucleate boiling. The PCM-1 test was performed to provide in-pile data for a ''worst case'' PCM incident. The objective of this experiment was to study the behavior of a single pressurized water reactor (PWR) fuel rod subjected to a high-power and low flow environment which would result in cladding failure at full power. The ''worst case'' conditions established for the experiment consisted of a rod peak power of 78.7 kW/m and a coolant mass flux of 1356 kg/s.m 2 . Fuel temperatures at the stipulated operating conditions were such that a significant volume of molten fuel was present when failure occurred which produced a high probability of molten fuel-coolant interaction (MFCI) with the possibility of a vapor explosion

  17. Secondary hydriding of defected zircaloy-clad fuel rods

    International Nuclear Information System (INIS)

    Olander, D.R.; Vaknin, S.

    1993-01-01

    The phenomenon of secondary hydriding in LWR fuel rods is critically reviewed. The current understanding of the process is summarized with emphasis on the sources of hydrogen in the rod provided by chemical reaction of water (steam) introduced via a primary defect in the cladding. As often noted in the literature, the role of hydrogen peroxide produced by steam radiolysis is to provide sources of hydrogen by cladding and fuel oxidation that are absent without fission-fragment irradiation of the gas. Quantitative description of the evolution of the chemical state inside the fuel rod is achieved by combining the chemical kinetics of the reactions between the gas and the fuel and cladding with the transport by diffusion of components of the gas in the gap. The chemistry-gas transport model provides the framework into which therate constants of the reactions between the gases in the gap and the fuel and cladding are incorporated. The output of the model calculation is the H 2 0/H 2 ratio in the gas and the degree of claddingand fuel oxidation as functions of distance from the primary defect. This output, when combined with a criterion for the onset of massive hydriding of the cladding, can provide a prediction of the time and location of a potential secondary hydriding failure. The chemistry-gas transport model is the starting point for mechanical and H-in-Zr migration analyses intended to determine the nature of the cladding failure caused by the development of the massive hydride on the inner wall

  18. Multidimensional simulations of hydrides during fuel rod lifecycle

    International Nuclear Information System (INIS)

    Stafford, D.S.

    2015-01-01

    In light water reactor fuel rods, waterside corrosion of zirconium-alloy cladding introduces hydrogen into the cladding, where it is slightly soluble. When the solubility limit is reached, the hydrogen precipitates into crystals of zirconium hydride which decrease the ductility of the cladding and may lead to cladding failure during dry storage or transportation events. The distribution of the hydride phase and the orientation of the crystals depend on the history of the spatial temperature and stress profiles in the cladding. In this work, we have expanded the existing hydride modeling capability in the BISON fuel performance code with the goal of predicting both global and local effects on the radial, azimuthal and axial distribution of the hydride phase. We compare results from 1D simulations to published experimental data. We demonstrate the new capability by simulating in 2D a fuel rod throughout a lifecycle that includes irradiation, short-term storage in the spent fuel pool, drying, and interim storage in a dry cask. Using the 2D simulations, we present qualitative predictions of the effects of the inter-pellet gap and the drying conditions on the growth of a hydride rim. - Highlights: • We extend BISON fuel performance code to simulate lifecycle of fuel rods. • We model hydrogen evolution in cladding from reactor through dry storage. • We validate 1D simulations of hydrogen evolution against experiments. • We show results of 2D axisymmetric simulations predicting hydride formation. • We show how our model predicts formation of a hydride rim in the cladding.

  19. Preliminary design report for the prototypical fuel rod consolidation system

    International Nuclear Information System (INIS)

    Rosa, J.M.

    1986-01-01

    This report documents NUTECH's preliminary design of a dry, spent fuel rod consolidation system. This preliminary design is the result of Phase I of a planned four phase project. The present report on this project provides a considerable amount of detail for a preliminary design effort. The design and all of its details are described in this Preliminary Design Report (PDR). The NUTECH dry rod consolidation system described herein is remotely operated. It provides for automatic operation, but with operator hold points between key steps in the process. The operator has the ability to switch to a manual operation mode at any point in the process. The system is directed by the operator using an executive computer which controls and coordinates the operation of the in-cell equipment. The operator monitors the process using an in-cell closed circuit television (CCTV) system with audio output and equipment status displays on the computer monitor. The in-cell mechanical equipment consists of the following: (1) two overhead cranes with manipulators; (2) a multi-degree of freedom fuel handling table and its clamping equipment; (3) a fuel assembly end fitting removal station and its tools; (4) a consolidator (which pulls rods, assembles the consolidated bundle and loads the canister); (5) a canister end cap welder and weld inspection system; (6) decontamination systems; and (7) the CCTV and microphone systems

  20. Numerically predicting horizontally oriented spent fuel rod surface temperatures

    International Nuclear Information System (INIS)

    Wix, S.D.; Koski, J.A.

    1993-01-01

    A comparison between numerical calculations with use of commercial thermal analysis software packages and experimental data simulating a horizontally oriented spent fuel rod array was performed. Twelve cases were analyzed using air and helium for the fill gas, with three different heat dissipation levels. The numerically predicted temperatures are higher than the experimental data for all levels of heat dissipation with air as the fill gas. The temperature differences are 4 degrees C and 23 degrees C for the low heat dissipation and high dissipation, respectively. The temperature predictions using helium as a fill gas are lower than the experimental data for the low and medium heat dissipation levels. The temperature predictions are 1 degrees C and 6 degrees C lower than the experimental data for the low and medium heat dissipation, respectively. For the high heat dissipation level, the temperature predictions are 16 degrees C higher than the experimental data. Differences between the predicted and experimental temperatures can be attributed to several factors. These factors include a experimental uncertainity in the temperature and heat dissipation measurements, actual convection effects not included in the model, and axial heat flow in the experimental data. This works demonstrates that horizontally oriented spent fuel rod surface temperature predictions can be made using existing commercial software packages. This work also shows that end effects, such as axial heat transfer through the spent fuel rods, will be increasingly important as the amount of dissipated heat increases

  1. Numerically predicting horizontally oriented spent fuel rod surface temperatures

    International Nuclear Information System (INIS)

    Wix, S.D.; Koski, J.A.

    1992-01-01

    A comparison between numerical calculations with use of commercial thermal analysis software packages and experimental data simulating a horizontally oriented spent fuel rod array was performed. Twelve cases were analyzed using air and helium for the fill gas, with three different heat dissipation levels. The numerically predicted temperatures are higher than the experimental data for all levels of heat dissipation with air as the fill gas. The temperature differences are 4 degree C and 23 degree C for the low heat dissipation and high heat dissipation, respectively. The temperature predictions using helium as a fill gas are lower than the experimental data for the low and medium heat dissipation levels. The temperature predictions are 1 degree C and 6 degree C lower than the experimental data for the low and medium heat dissipation, respectively. For the high heat dissipation level, the temperature predictions are 16 degree C higher than the experimental data. Differences between the predicted and experimental temperatures can be attributed to several factors. These factors include experimental uncertainty in the temperature and heat dissipation measurements, actual convection effects not included in the model, and axial heat flow in the experimental data. This work demonstrates that horizontally oriented spent fuel rod surface temperature predictions can be made using existing commercial software packages. This work also shows that end effects, such as axial heat transfer through the spent fuel rods, will be increasingly important as the amount of dissipated heat increases

  2. Technique of manufacturing specimen of irradiated fuel rods

    International Nuclear Information System (INIS)

    Min, Duck Seok; Seo, Hang Seok; Min, Duck Kee; Koo, Dae Seo; Lee, Eun Pyo; Yang, Song Yeol

    1999-04-01

    Technique of manufacturing specimen of irradiated fuel rods to perform efficient PIE is developed by analyzing the relation between requiring time of manufacturing specimen and manufacturing method in irradiated fuel rods. It takes within an hour to grind 1 mm of specimen thickness under 150 rpm in speed of grinding, 600 g gravity in force using no.120, no.240, no.320 of grinding paper. In case of no.400 of grinding paper, it takes more an hour to grind the same thickness as above. It takes up to a quarter to grind 80-130 μm in specimen thickness using no.400 of grinding paper. When grinding time goes beyond 15 minutes, the grinding thickness of specimen does not exist. The polishing of specimen with 150 Rpms in speed of grinding machine, 600 g gravity in force, 10 minutes in polishing time using diamond paste 15 μm on polishing cloths amounts to 50 μm in specimen thickness. In case of diamond paste 9 μm on polishing cloth, the polishing of specimen amounts to 20 μm. The polishing thickness of specimen with 15 minutes in polishing time using 6 μm, 3 μm, 1 μm, 1/4 μm does not exist. Technique of manufacturing specimen of irradiated fuel rods will have application to the destructive examination of PIE. (author). 6 refs., 1 tab., 10 figs

  3. Large eddy simulation of a fuel rod subchannel

    International Nuclear Information System (INIS)

    Mayer, Gusztav

    2007-01-01

    In a VVER-440 reactor the measured outlet temperature is related to fuel limit parameters and the power upgrading plans of VVER-440 reactors motivated us to obtain more information on the mixing process of the fuel assemblies. In a VVER-440 rod bundle the fuel rods are arranged in triangular array. Measurement shows (Krauss and Meyer, 1998) that the classical engineering approach, which tries to trace the characterization of such systems back to equivalent (hydraulic diameter) pipe flows, does not give reasonable results. Due to the different turbulence characteristics, the mixing is more intensive in rod bundles than it would be expected based on equivalent pipe flow correlations. As a possible explanation of the high mixing, secondary flow was deduced from measurements by several experimentalists (Trupp and Azad, 1975). Another candidate to explain the high mixing is the so-called flow pulsation phenomenon (Krauss and Meyer, 1998). In this paper we present subchannel simulations (Mayer et al. 2007) using large eddy simulation (LES) methodology and the lattice Boltzmann method (LBM) without the spacers at Reynolds number 21000. The simulation results are compared with the measurements of Trupp and Azad (1975). The mean axial velocity profile shows good agreement with the measurement data. Secondary flow has been observed directly in the simulation results. Reasonable agreement has been achieved for most Reynolds stresses. Nevertheless, the calculated normal stresses show small, but systematic deviation from the measurement data. (author)

  4. Vibration mechanism of fuel rod in axial flow

    International Nuclear Information System (INIS)

    Kang, Heung Seok; Yoon, Kyung Ho; Kim, Hyung Kyu; Song, Kee Nam

    1998-08-01

    This is a review on the previous researches for the vibration of fuel rod induced by axial flow. The analysis methods are classified into three categories accordingly as the researchers postulate the vibration to be self-excited, forced and parametric; the self-excited mechanism by Burgreen and Quinn, the forced one by Reavis, Gorman, kanazawa, and S. Chen, and the parametric one by Y. Chen. Quinn supposed that the centrifugal force by flow exaggerated the natural bow in the cylinder, and the flexural force by it diminished the bow by turns; this interactive motion leaded cylinder to vibration. The supporters to the forced mechanism considered the forces arising from pressure perturbation within the boundary layers as vibrating sources. Y. Chen insisted that the cylinder could only be excited to vibration in resonance by the small oscillation of mean flow velocity. The previous studies were based on the simple boundary conditions such as hinged-hinged or fixed-fixed single span. Therefore, for the more accurate prediction of the fuel rod vibration in reactor, the further studies need to reflect the actual boundary conditions of the fuel rod like axial force and continuous supports by grids. (author). 25 refs

  5. 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

  6. Simulation of fuel rod irradiation capsules in water loops by electric heater rods

    International Nuclear Information System (INIS)

    Lopez, J.; Montes, M.; Serrano, J.; Haefner, H.E.

    1984-01-01

    The out of pile simulation of irradiation devices was carried out by J.E.N. in the frame of the KfK-JEN joint experiment for irradiation of fast reactor fuel rods (IVO-FR2-Vg7). A typical single-wall-Nak (22% Na, 78% K) electrical heated capsule was fabricated and hydraulical tests were done. The capsule was instrumented with 10 thermocouples in order to obtain the radial temperature profile into the capsule in function of the electrical rod power (max. 215 w/cm), flow rate (max. 2,4 m 3 /h) and coolant temperature (max. 60degC). The experimental values are compared to the Tecap-Code results. (author)

  7. A Study on the Structural Integrity Issues of a Dual-Cooled Fuel Rod

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyung-Kyu; Lee, Kang-Hee; Lee, Young-Ho; Yoon, Kyung-Ho; Kim, Jae-Yong; Song, Kun-Woo [Korea Atomic Energy Research Institute, 1045 Daedeokdaero Yuseong Daejeon 305-353 (Korea, Republic of)

    2009-06-15

    A dual-cooled fuel rod has an internal coolant flow passage in addition to the external one. A remarkable power up-rate can be achieved due to the increased surface area, which may draw great interests from the fuel researchers, designers and vendors. However, it requires effective resolution to the difficult technical issues when a fuel assembly is to be realized. It becomes much more difficult if a tough boundary condition needs to be satisfied such as a compatibility with the existing reactor internal structures. This kind of challenge is tackled through a national R and D project in Korea: to develop the structural components of a dual-cooled fuel that should be compatible with the current OPR 1000 (Korea Standard Nuclear Power Plant) internal structures. Fuel rod supporting structures, top and bottom end pieces and guide tubes are the components. Besides, the fuel rod components have to be developed as well since the fuel rod's geometry becomes much different from the conventional rod's one. The dimension change may well affect the above mentioned structural components. As a part of the work, structural integrity of the components of a dual-cooled fuel rod is studied in this paper. The investigated topics are: i) the thickness determination of a cladding tube (especially outer tube of a large diameter), ii) vibration issue of an inner cladding tube, iii) design concern of plenum spring and spacer. The cladding thickness issue arises due to the increased outside diameter of a fuel rod, which is caused by an internal flow passage formation. Among the criteria for the thickness determination, an elastic buckling criteria was focused on. Theoretical background for the well-known formula (such as a stability problem) was revisited. Verification tests were carried out independently with using a cladding tube of PHWR fuel rod. Results showed that the formula was not conservative to apply for the cladding thickness determination. Minimum thickness for the

  8. COBRA-IV-I: an interim version of COBRA for thermal-hydraulic analysis of rod bundle nuclear fuel elements and cores

    Energy Technology Data Exchange (ETDEWEB)

    Wheeler, C.L.; Stewart, C.W.; Cena, R.J.; Rowe, D.S.; Sutey, A.M.

    1976-03-01

    The COBRA-IV-I computer code uses the subchannel analysis approach to determine the enthalpy and flow distribution in rod bundles for both steady-state and transient conditions. The steady-state and transient solution schemes used in COBRA-IIIC are still available in COBRA-IV-I as the implicit solution scheme option. In addition to these techniques, a new explicit solution scheme is now available which allows the calculation of severe transients involving flow reversals, recirculations, expulsion and reentry flows, with a pressure or flow boundary condition specified. Significant storage compaction and reduced running times have been achieved to allow the calculation of problems involving hundreds of subchannels.

  9. Heat removal in gas-cooled fuel rod clusters

    International Nuclear Information System (INIS)

    Rehme, K.

    1975-01-01

    For a thermo- and fluid-dynamic analysis of fuel rod cluster subchannels for gas-cooled breeder reactors, the following values must be verified: a) friction coefficient as flow parameter; b) Stanton number as heat transfer parameter; c) influence of spacers on friction coefficient and Stanton number; d) heat and mass exchange between subchannels with different temperatures. These parameters are established by combining results of single experiments and of integral experiments. Mention is made of further studies to be performed in order to determine the heat removal from gas-cooled fast breeder fuel elements. (HR) [de

  10. Improved nuclear fuel element

    International Nuclear Information System (INIS)

    Klepfer, H.H.

    1974-01-01

    A nuclear fuel element is described which comprises: 1) an elongated clad container, 2) a layer of high lubricity material being disposed in and adjacent to the clad container, 3) a low neutron capture cross section metal liner being disposed in the clad container and adjacent to the layer, 4) a central core of a body of nuclear fuel material disposed in and partially filling the container and forming an internal cavity in the container, 5) an enclosure integrally secured and sealed at each end of the container, and a nuclear fuel material retaining means positioned in the cavity. (author)

  11. Modelling of WWER-440 fuel rod behaviour under operational conditions with the PIN-micro code

    Energy Technology Data Exchange (ETDEWEB)

    Stefanova, S; Vitkova, M; Simeonova, V; Passage, G; Manolova, M [Institute for Nuclear Research and Nuclear Energy, Sofia (Bulgaria); Haralampieva, Z [National Electric Company Ltd., Kozloduy (Bulgaria); Scheglov, A; Proselkov, V [Institute of Nuclear Reactors, RSC Kurchatov Inst., Moscow (Russian Federation)

    1997-08-01

    The report summarizes the first practical experience obtained by fuel rod performance modelling at the Institute for Nuclear Research and Nuclear Energy, Bulgarian Academy of Sciences. The results of application of the PIN-micro code and the code modification PINB1 for thermomechanical analysis of WWER-440 fuel assemblies (FAs) are presented. The aim of this analysis is to study the fuel rod behaviour of the operating WWER reactors. The performance of two FAs with maximal linear power and varying geometrical and technological parameters is analyzed. On the basis of recent publications on WWER fuel performance modelling at extended burnup, a modified PINB1 version of the standard PIN-micro code is shortly described and applied for the selected FAs. Comparison of the calculated results is performed. The PINB1 version predicts higher fuel temperatures and more adequate FGR rate, accounting for the extended burnup. The results presented in this paper prove the existence of sufficient safety margins, for the fuel performance limiting parameters during the whole considered period of core operation. (author). 8 refs, 16 figs, 1 tab.

  12. Modelling of WWER-440 fuel rod behaviour under operational conditions with the PIN-micro code

    International Nuclear Information System (INIS)

    Stefanova, S.; Vitkova, M.; Simeonova, V.; Passage, G.; Manolova, M.; Haralampieva, Z.; Scheglov, A.; Proselkov, V.

    1997-01-01

    The report summarizes the first practical experience obtained by fuel rod performance modelling at the Institute for Nuclear Research and Nuclear Energy, Bulgarian Academy of Sciences. The results of application of the PIN-micro code and the code modification PINB1 for thermomechanical analysis of WWER-440 fuel assemblies (FAs) are presented. The aim of this analysis is to study the fuel rod behaviour of the operating WWER reactors. The performance of two FAs with maximal linear power and varying geometrical and technological parameters is analyzed. On the basis of recent publications on WWER fuel performance modelling at extended burnup, a modified PINB1 version of the standard PIN-micro code is shortly described and applied for the selected FAs. Comparison of the calculated results is performed. The PINB1 version predicts higher fuel temperatures and more adequate FGR rate, accounting for the extended burnup. The results presented in this paper prove the existence of sufficient safety margins, for the fuel performance limiting parameters during the whole considered period of core operation. (author). 8 refs, 16 figs, 1 tab

  13. Temperature measurement in cans of fuel rods and fuel rod simulators

    International Nuclear Information System (INIS)

    Tschoeke, H.; Moeller, R.

    1977-01-01

    On the sodium-cooled 19-rod cluster model for the SNR 300 the can wall temperature distributions of the non-uniformly cooled rods were measured with thermocouples mounted in outer grooves in the peripheral zone, permitting, in connection with Ni solder, a practically undisturbed measurement. For a more exact determination of the local surface temperature a calibration method, the so-called double-wall method, was developed and applied. The description of this calibration method and the experimental results achieved until now are presented. (orig./RW) [de

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

    International Nuclear Information System (INIS)

    De Mario, E.E.

    1988-01-01

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

  15. Using Finite Model Analysis and Out of Hot Cell Surrogate Rod Testing to Analyze High Burnup Used Nuclear Fuel Mechanical Properties

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [ORNL; Jiang, Hao [ORNL; Wang, Hong [ORNL

    2014-07-01

    Based on a series of FEA simulations, the discussions and the conclusions concerning the impact of the interface bonding efficiency to SNF vibration integrity are provided in this report; this includes the moment carrying capacity distribution between pellets and clad, and the impact of cohesion bonding on the flexural rigidity of the surrogate rod system. As progressive de-bonding occurs at the pellet-pellet interfaces and at the pellet-clad interface, the load ratio of the bending moment carrying capacity gradually shifts from the pellets to the clad; the clad starts to carry a significant portion of the bending moment resistance until reaching the full de-bonding state at the pellet-pellet interface regions. This results in localized plastic deformation of the clad at the pellet-pellet-clad interface region; the associated plastic deformations of SS clad leads to a significant degradation in the stiffness of the surrogate rod. For instance, the flexural rigidity was reduced by 39% from the perfect bond state to the de-bonded state at the pellet-pellet interfaces.

  16. Nuclear Fuel Reprocessing

    International Nuclear Information System (INIS)

    Simpson, Michael F.; Law, Jack D.

    2010-01-01

    This is a submission for the Encyclopedia of Sustainable Technology on the subject of Reprocessing Spent Nuclear Fuel. Nuclear reprocessing is the chemical treatment of spent fuel involving separation of its various constituents. Principally, it is used to recover useful actinides from the spent fuel. Radioactive waste that cannot be re-used is separated into streams for consolidation into waste forms. The first known application of nuclear reprocessing was within the Manhattan Project to recover material for nuclear weapons. Currently, reprocessing has a peaceful application in the nuclear fuel cycle. A variety of chemical methods have been proposed and demonstrated for reprocessing of nuclear fuel. The two most widely investigated and implemented methods are generally referred to as aqueous reprocessing and pyroprocessing. Each of these technologies is described in detail in Section 3 with numerous references to published articles. Reprocessing of nuclear fuel as part of a fuel cycle can be used both to recover fissionable actinides and to stabilize radioactive fission products into durable waste forms. It can also be used as part of a breeder reactor fuel cycle that could result in a 14-fold or higher increase in energy utilization per unit of natural uranium. Reprocessing can also impact the need for geologic repositories for spent fuel. The volume of waste that needs to be sent to such a repository can be reduced by first subjecting the spent fuel to reprocessing. The extent to which volume reduction can occur is currently under study by the United States Department of Energy via research at various national laboratories and universities. Reprocessing can also separate fissile and non-fissile radioactive elements for transmutation.

  17. Spent nuclear fuel storage

    International Nuclear Information System (INIS)

    Romanato, Luiz Sergio

    2005-01-01

    When a country becomes self-sufficient in part of the nuclear cycle, as production of fuel that will be used in nuclear power plants for energy generation, it is necessary to pay attention for the best method of storing the spent fuel. Temporary storage of spent nuclear fuel is a necessary practice and is applied nowadays all over the world, so much in countries that have not been defined their plan for a definitive repository, as well for those that already put in practice such storage form. There are two main aspects that involve the spent fuels: one regarding the spent nuclear fuel storage intended to reprocessing and the other in which the spent fuel will be sent for final deposition when the definitive place is defined, correctly located, appropriately characterized as to several technical aspects, and licentiate. This last aspect can involve decades of studies because of the technical and normative definitions at a given country. In Brazil, the interest is linked with the storage of spent fuels that will not be reprocessed. This work analyses possible types of storage, the international panorama and a proposal for future construction of a spent nuclear fuel temporary storage place in the country. (author)

  18. Nuclear fuel element

    International Nuclear Information System (INIS)

    Yamamoto, Seigoro.

    1994-01-01

    Ultrafine particles of a thermal neutron absorber showing ultraplasticity is dispersed in oxide ceramic fuels by more than 1% to 10% or lower. The ultrafine particles of the thermal neutron absorber showing ultrafine plasticity is selected from any one of ZrGd, HfEu, HfY, HfGd, ZrEu, and ZrY. The thermal neutron absorber is converted into ultrafine particles and solid-solubilized in a nuclear fuel pellet, so that the dispersion thereof into nuclear fuels is made uniform and an absorbing performance of the thermal neutrons is also made uniform. Moreover, the characteristics thereof, for example, physical properties such as expansion coefficient and thermal conductivity of the nuclear fuels are also improved. The neutron absorber, such as ZrGd or the like, can provide plasticity of nuclear fuels, if it is mixed into the nuclear fuels for showing the plasticity. The nuclear fuel pellets are deformed like an hour glass as burning, but, since the end portion thereof is deformed plastically within a range of a repulsive force of the cladding tube, there is no worry of damaging a portion of the cladding tube. (N.H.)

  19. Evolution of fuel rod support under irradiation consequences on the mechanical behavior of fuel assembly

    International Nuclear Information System (INIS)

    Billerey, A.; Bouffioux, P.

    2002-01-01

    The complete paper follows. According to the fuel management policy in French PWR with respect to high burn-up, the prediction of the mechanical behavior of the irradiated fuel assembly is required as far as excessive deformations of fuel assembly might lead to incomplete Rod Cluster Control Assembly insertion (safety problems) and fretting wear lead to leaking rods (plant operation problems). One of the most important parameter is the evolution of the fuel rod support in the grid cell as it directly governs the mechanical behavior of the fuel assembly and consequently allows to predict the behavior of irradiated structure in terms of (i) axial and lateral deformation (global behavior of the assembly) and (ii) fretting wear (local behavior of the rod). Fuel rod support is provided by a spring-dimple system fixed on the grid. During irradiation, the spring force decreases and a gap between the rod and the spring might open. This phenomenon is due to (i) irradiation-induced stress relaxation for the spring and for the dimples, (ii) grid growth and (iii) reduction of rod diameter. Two models have been developed to predict the behavior of the rod in the grid cell. The first model is able to evaluate the spring force relaxation during irradiation. The second one is able to evaluate the rotation characteristic of the fuel rod in the cell, function of the spring force. The main input parameters are (i) the creep laws of the grid materials, (ii) the growth law of the grid, (iii) the evolution of rod diameter and (iv) the design of the fuel rod support. The objectives of this paper are to: (i) evaluate the consequences of grid support design modifications on the fretting sensitivity in terms of predicted maximum gap during irradiation and operational time to gap appearance; (ii) evaluate, using a non-linear Finite Element assembly model, the impact of the evolution of grid support under irradiation on the mechanical behavior of the full assembly in terms of axial and

  20. Fuel element for a nuclear reactor

    International Nuclear Information System (INIS)

    Linning, D.L.

    1977-01-01

    An improvement of the fuel element for a fast nuclear reactor described in patent 15 89 010 is proposed which should avoid possible damage due to swelling of the fuel. While the fuel element according to patent 15 89 010 is made in the form of a tube, here a further metal jacket is inserted in the centre of the fuel rod and the intermediate layer (ceramic uranium compound) is provided on both sides, so that the nuclear fuel is situated in the centre of the annular construction. Ceramic uranium or plutonium compounds (preferably carbide) form the fuel zone in the form of circular pellets, which are surrounded by annular gaps, so that gaseous fission products can escape. (UWI) [de