WorldWideScience

Sample records for fuel elements transportation

  1. Fuel element transport container

    International Nuclear Information System (INIS)

    Benna, P.; Neuenfeldt, W.

    1979-01-01

    The reprocessing system includes a large number of waterfilled ponds next to each other for the intermediate storage of fuel elements from LWR's. The fuel element transport device is allocated to a middle pond. The individual ponds are separated from each other by walls, and are only accessible from the middle pond via narrow passages. The transport device includes a telescopic running rail for a trolley with a grab device for the fuel element. The running rail is supported in turn by a second trolley, which can be moved by wheels on rails. Part of the drive of the first trolley is arranged on the second one. Using this transport device, adjacent ponds can be served through the passage openings. (DG) [de

  2. Transportation of irradiated fuel elements

    International Nuclear Information System (INIS)

    Preece, A.H.

    1980-01-01

    The report falls under the headings: introduction (explaining the special interest of the London Borough of Brent, as forming part of the route for transportation of irradiated fuel elements); nuclear power (with special reference to transport of spent fuel and radioactive wastes); the flask aspect (design, safety regulations, criticisms, tests, etc.); the accident aspect (working manual for rail staff, train formation, responsibility, postulated accident situations); the emergency arrangements aspect; the monitoring aspect (health and safety reports); legislation; contingency plans; radiation - relevant background information. (U.K.)

  3. Transportation of irradiated fuel elements

    International Nuclear Information System (INIS)

    1980-01-01

    A critique is presented of current methods of transporting spent nuclear fuel and the inadequacies of the associated contingency plans, with particular reference to the transportation of irradiated fuel through London. Anti-nuclear and pro-nuclear arguments are presented on a number of factors, including tests on flasks, levels of radiation exposure, routine transport arrangements and contingency arrangements. (U.K.)

  4. Remarks on the transportation of spent fuel elements

    International Nuclear Information System (INIS)

    Krull, W.

    1992-01-01

    Information and data are provided on several aspects of the transportation of spent fuel elements. These aspects include contract, transportation, reprocessing batch size, and economical considerations. (author)

  5. An improved assembly for the transport of fuel elements

    International Nuclear Information System (INIS)

    Myers, G.

    1979-01-01

    An improved assembly for the transport and storage of radioactive nuclear fuel elements is described. The fuel element transport canister is of the type in which the fuel elements are submerged in liquid with a self regulating ullage system, so that the fuel elements are always submerged in the liquid even when the assembly is used in one orientation during loading and another orientation during transportation. (UK)

  6. Assembly for transport and storage of radioactive nuclear fuel elements

    International Nuclear Information System (INIS)

    Myers, G.

    1978-01-01

    The invention concerns the self-control of coolant deficiencies on the transport of spent fuel elements from nuclear reactors. It guarantees that drying out of the fuel elements is prevented in case of a change of volume of the fluid contained in storage tanks and accumulators and serving as coolant and shielding medium. (TK) [de

  7. Method to mount defect fuel elements i transport casks

    International Nuclear Information System (INIS)

    Borgers, H.; Deleryd, R.

    1996-01-01

    Leaching or otherwise failed fuel elements are mounted in special containers that fit into specially designed chambers in a transportation cask for transport to reprocessing or long-time storage. The fuel elements are entered into the container under water in a pool. The interior of the container is dried before transfer to the cask. Before closing the cask, its interior, and the exterior of the container are dried. 2 figs

  8. The permission of transport of irradiated nuclear fuel elements

    International Nuclear Information System (INIS)

    Klomberg, T.J.M.

    2000-01-01

    In July and October 2000 the Dutch government granted permits for the transportation of irradiated nuclear fuel elements. The environmental organization Greenpeace objected against the permit, but that was rejected by the Dutch Council of State. A brief overview is given of the judgements and the state-of-the-art with respect to the transportation of the elements from Dutch reactors and storage facilities in Petten, Dodewaard and Borssele to Cogema in La Hague, France and BNFL in Sellafield, England

  9. Fuel element transport container with a removable cover

    International Nuclear Information System (INIS)

    Dannehl, G.; Fink, W.; Haenle, G.

    1980-01-01

    The cover of the fuel element transport container is removably fixed with screws on a flange as mechanical loads have to be expected during the transfer to the disposal plant. A ring-shaped or star-shaped clamping device grips over the cover. It has a clamp claw to lock the cover and permits unscrewing without unlocking the cover. (DG) [de

  10. Remarks on the transportation of spent fuel elements

    International Nuclear Information System (INIS)

    Krull, W.

    1986-01-01

    In this chapter topics discussed are the need for contracts, a transport company and risk insurance. Also, a section on transportation covers cranes, subpressure, contamination, cask limitations, physical protection and shipping. Reprocessing discusses minimum reprocessing batch and spent fuel. Finally, economical considerations concerning transportation and reprocessing are given

  11. Dynamic analysis and application of fuel elements pneumatic transportation in a pebble bed reactor

    International Nuclear Information System (INIS)

    Liu, Hongbing; Du, Dong; Han, Zandong; Zou, Yirong; Pan, Jiluan

    2015-01-01

    Almost 10,000 spherical fuel elements are transported pneumatically one by one in the pipeline outside the core of a pebble bed reactor every day. Any failure in the transportation will lead to the shutdown of the reactor, even safety accidents. In order to ensure a stable and reliable transportation, it's of great importance to analyze the motion and force condition of the fuel element. In this paper, we focus on the dynamic analysis of the pneumatic transportation of the fuel element and derive kinetic equations. Then we introduce the design of the transportation pipeline. On this basis we calculate some important data such as the velocity of the fuel element, the force between the fuel element and the pipeline and the efficiency of the pneumatic transportation. Then we analyze these results and provide some suggestions for the design of the pipeline. The experiment was carried out on an experimental platform. The velocities of the fuel elements were measured. The experimental results were consistent with and validated the theoretical analysis. The research may offer the basis for the design of the transportation pipeline and the optimization of the fuel elements transportation in a pebble bed reactor. - Highlights: • The kinetic equations of the fuel element in pneumatic transportation are derived. • The dynamic characteristics of the fuel element are analyzed. • Some important parameters are calculated based on the kinetic equations. • The experimental results were consistent with the analysis and verified the analysis. • This paper may offer an important guide to the research of a pebble bed reactor

  12. Fuel element

    International Nuclear Information System (INIS)

    Armijo, J.S.

    1976-01-01

    A fuel element for nuclear reactors is proposed which has a higher corrosion resisting quality in reactor operations. The zirconium alloy coating around the fuel element (uranium or plutonium compound) has on its inside a protection layer of metal which is metallurgically bound to the substance of the coating. As materials are namned: Alluminium, copper, niobium, stainless steel, and iron. This protective metallic layer has another inner layer, also metallurgically bound to its surface, which consists usually of a zirconium alloy. (UWI) [de

  13. Fuel element

    International Nuclear Information System (INIS)

    Kennedy, S.T.

    1982-01-01

    A nuclear reactor fuel element wherein a stack of nuclear fuel is prevented from displacement within its sheath by a retainer comprising a tube member which is radially expanded into frictional contact with the sheath by means of a captive ball within a tapered bore. (author)

  14. Process and device for cooling of nuclear reactor fuel elements enclosed in a transport container

    International Nuclear Information System (INIS)

    Stiefel, M.

    1986-01-01

    In order to remove the post-decay heat of the fuel elements contained in them, transport containers for burnt-up fuel elements can be connected to a water cooling circuit. In order to avoid thermal shocks, a tenside forming foam and air are introduced into the cooling circuit before its entry into the transport container in the direction of flow. The tenside and air continue to be supplied until the temperature inside the transport container has fallen below the temperature at which the foam is destroyed. By adding tenside and air, a two phase mixture is produced, which foams greatly when it enters the transport container and which cools the fuel elements so as to protect them.(orig./HP) [de

  15. Apparatus for the storage of transport- and storage-containers containing radioactive fuel elements

    International Nuclear Information System (INIS)

    Vox, A.

    1983-01-01

    The invention concerns an apparatus for the storage of transport and storage containers containing radioactive fuel elements. For each transport or storage container there is a separate silo-type container of steel, concrete, prestressed concrete or suchlike breakproof and fireproof material, to be placed in the open, that can be opened for removal and placing of the transport or storage container respectively. (orig.) [de

  16. Magnesium transport extraction of transuranium elements from LWR fuel

    International Nuclear Information System (INIS)

    Ackerman, J.P.; Battles, J.E.; Johnson, T.R.; Miller, W.E.; Pierce, R.D.

    1992-01-01

    This patent describes a process of separating transuranium actinide values from uranium values present in spent nuclear oxide fuel containing rare earth and noble metal fission products as well as fission products of alkali metals, alkaline earth metals and iodine. It comprises reducing the oxide fuel with Ca metal in the presence of Ca halide; separating the Ca halide with the CaO and the fission products contained therein from the U-Fe alloy and the metal values dissolved therein and electrolytically contacting the calcium salts with a carbon electrode; contacting the liquid U-Fe alloy having dissolved therein reduced metals from the spent nuclear fuel with liquid Mg metal, thereafter separating the liquid Mg and the metals dissolved therein from the U-Fe alloy and the metal dissolved therein, distilling the Mg from the transuranium actinide and rare earth metals, recontacting the U-Fe alloy with liquid Mg metal a sufficient number of times until not less than about 99% by weight of the transuranium actinide values have been removed from the U-Fe alloy

  17. Development of a transport cask for spent fuel elements of research reactors

    International Nuclear Information System (INIS)

    Quintana, F.; Saliba, R.O.; Furnari, J.C.; Mourao, R.P; Leite da Silva, L.; Novara, O.; Alexandre Miranda, C.; Mattar Neto, M.

    2012-01-01

    This article presents an overview of the development of a research reactor spent fuel transport cask. Through a project funded by the IAEA, Argentina, Brazil and Chile have collaborated to enhance regional capacity in the management of spent fuel elements from research reactors operated in the region. A packaging for the transport of research reactors spent fuel was developed. It was designed by a team of researchers from the countries mentioned and a 1:2 scale model for MTR type fuel was constructed in Argentina and subsequently tested in CDTN facilities in Belo Horizonte, Brazil. There were three test sequences to test the cask for normal transport and hypothetical accident conditions. It has successfully passed the tests and the overall performance was considered satisfactory. As part of the licensing process, a test sequence with the presence of regulatory authorities is scheduled for December, 2012 (author)

  18. Thermal simulations and tests in the development of a helmet transport spent fuel elements Research Reactor

    International Nuclear Information System (INIS)

    Saliba, R.; Quintana, F.; Márquez Turiello, R.; Furnari, J.C.; Pimenta Mourão, R.

    2013-01-01

    A packaging for the transport of irradiated fuel from research reactors was designed by a group of researchers to improve the capability in the management of spent fuel elements from the reactors operated in the region. Two half-scale models for MTR fuel were constructed and tested so far and a third one for both MTR and TRIGA fuels will be constructed and tested next. Four test campaigns have been carried out, covering both normal and hypothetical accident conditions of transportation. The thermal test is part of the requirements for the qualification of transportation packages for nuclear reactors spent fuel elements. In this paper both the numerical modelling and experimental thermal tests performed are presented and discussed. The cask is briefly described as well as the finite element model developed and the main adopted hypotheses for the thermal phenomena. The results of both numerical runs and experimental tests are discussed as a tool to validate the thermal modelling. The impact limiters, attached to the cask for protection, were not modelled. (author) [es

  19. Fuel element

    International Nuclear Information System (INIS)

    Hirose, Yasuo.

    1982-01-01

    Purpose: To increase the plenum space in a fuel element used for a liquid metal cooled reactor. Constitution: A fuel pellet is secured at one end with an end plug and at the other with a coil spring in a tubular container. A mechanism for fixing the coil spring composed of a tubular unit is mounted by friction with the inner surface of the tubular container. Accordingly, the recoiling force of the coil spring can be retained by fixing mechanism with a small volume, and since a large amount of plenum space can be obtained, the internal pressure rise in the cladding tube can be suppressed even if large quantities of fission products are discharged. (Kamimura, M.)

  20. Convective-diffusive transport of fission products in the gap of a failed fuel element

    International Nuclear Information System (INIS)

    Lian, Z.W.; Carlucci, L.N.; Arimescu, V.I.

    1995-03-01

    A model is presented to describe the transport behaviour of gaseous fission products along the axial fuel-to-sheathe gap of a failed fuel element to the coolant system. The model is applicable to an element having failed under normal operating conditions or loss-of coolant-accident conditions. Because of the large differences in operating parameters, the transport characteristics of gaseous fission products in a failed element under these two operating conditions are significantly different. However, in both cases the transport process can be described by convection-diffusion caused by the continuous release of fission products from the fuel to the gap. Under normal operating conditions, the bulk-flow velocity is found to be negligible, due to the low release rate of fission products from fuel. The process can be well approximated by the diffusion of fission products in a stagnant gas-steam mixture. The effect of convection on the fission product transport, however, becomes significant under loss-of-coolant-accident conditions, where the release rates of fission products from fuel can be several orders of magnitude higher that that under normal operating conditions. The convection of the mixture in the gap not only contributes an additional flux to the gas-mixture transport, but also increases the gradient of fission products concentration across the opening, and therefore increases the diffusion flux to the coolant. As a result of the bulk flow, the transport of fission products along the gap is accelerated and the hold-up of short-lived isotopes in the gap is significantly reduced. Steam ingress through the opening into the gap is obstructed by the bulk flow, resulting in low steam concentrations in the gap under loss-of-coolant-accident conditions. (author). 6 refs., 8 figs

  1. Transport of volatile fission products in the fuel-to-sheath gap of defective fuel elements during normal and reactor accident conditions

    International Nuclear Information System (INIS)

    Lewis, B.J.; Bonin, H.W.

    1995-01-01

    An analytical treatment has been used to model the vapour transport of radioactive fission products released into the fuel-to-sheath gap of defective nuclear fuel elements. The model accounts for both diffusive and bulk-convective transport. Convective transport becomes important as a result of a significant release of gaseous fission products into the gap during a high-temperature reactor accident. However, during normal reactor operation, diffusion is shown to be the dominant process of transport. The model is based on an analysis of several in-reactor tests with operating defective fuel elements, and high-temperature annealing experiments with irradiated fuel specimens. ((orig.))

  2. An approach for the design of closure bolts of spent fuel elements transportation packages

    International Nuclear Information System (INIS)

    Mattar Neto, Miguel; Miranda, Carlos A.J.; Fainer, Gerson

    2009-01-01

    The spent fuel elements transportation packages must be designed for severe conditions including significant fire and impact loads corresponding to hypothetical accident conditions. In general, these packages have large flat lids connected to cylindrical bodies by closure bolts that can be the weak link in the containment system. The bolted closure design depends on the geometrical characteristics of the flat lid and the cylindrical body, including their flanges, on the type of the gaskets and their dimensions, and on the number, strength, and tightness of the bolts. There are well established procedures for the closure bolts design used in pressure vessels and piping. They can not be used directly in the bolts design applied to transportation packages. Prior to the use of these procedures, it is necessary consider the differences in the main loads (pressure for the pressure vessels and piping and impact loads for the transportation packages) and in the geometry (large flat lids are not used in pressure vessels and piping). So, this paper presents an approach for the design of the closure bolts of spent fuel elements transportation packages considering the impact loads and the typical geometrical configuration of the transportation packages. (author)

  3. Fuel element

    International Nuclear Information System (INIS)

    1974-01-01

    A new fuel can with a loose bottom and head is described. The fuel bar is attached to the loose bottom and head with two grid poles keeping the distance between bottom and head. A bow-shaped handle is attached to the head so that the fuel bar can be lifted from the can

  4. Transport fuel

    DEFF Research Database (Denmark)

    Ronsse, Frederik; Jørgensen, Henning; Schüßler, Ingmar

    2014-01-01

    Worldwide, the use of transport fuel derived from biomass increased four-fold between 2003 and 2012. Mainly based on food resources, these conventional biofuels did not achieve the expected emission savings and contributed to higher prices for food commod - ities, especially maize and oilseeds...

  5. Process and container system for transferring or transporting fuel elements from a nuclear power station to a store

    International Nuclear Information System (INIS)

    Vox, A.J.

    1984-01-01

    A system of containers with three types of containers (an inside container, a transport container and a storage container) is used. One either sets the inside container open on the lid side into the transport container first in the water pond of the nuclear power station, and one then sets the fuel elements into the inside container, or one places the inside container, loaded with fuel elements away from the transport container, into the transport container. Both containers are then closed and are transported to the store as a unit. The storage container open on the lid side is prepared there, the floor of the transport container is opened and this, together with the inside container, is lifted above the storage container or set above the storage container. The inside container is then lowered onto the storage container, the transport container is removed and the lid of the storage container is closed. (orig./HP) [de

  6. Fuel element loading system

    International Nuclear Information System (INIS)

    Arya, S.P; s.

    1978-01-01

    A nuclear fuel element loading system is described which conveys a plurality of fuel rods to longitudinal passages in fuel elements. Conveyor means successively position the fuel rods above the longitudinal passages in axial alignment therewith and adapter means guide the fuel rods from the conveyor means into the longitudinal passages. The fuel elements are vibrated to cause the fuel rods to fall into the longitudinal passages through the adapter means

  7. Method of transporting fuel assemblies

    International Nuclear Information System (INIS)

    Okada, Katsutoshi.

    1979-01-01

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

  8. Suggestions of radiation protection instruments in ships used for transporting spent fuel elements from nuclear power plants to central stores and further to fuel reprocessing plants

    International Nuclear Information System (INIS)

    Warenmo, G.

    1979-01-01

    Some radiation protection measures are necessary in ships which will be used for transporting spent fuel elements from nuclear power plants to central stores and further to fuel reprocessing plants in order to protect the crew from unnecessarily high radiation doses and to ensure that not allowable values occur. Such measures are discussed in this report as well as suitable radiation protection instruments for such ships. (E.R.)

  9. 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

  10. 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

  11. 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.)

  12. 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

  13. Concept study for interim storage of research reactor fuel elements in transport and storage casks. Transport and storage licensing procedure for the CASTOR MTR 2 cask. Final report

    International Nuclear Information System (INIS)

    Weiss, M.

    2001-01-01

    As a result of the project, a concept was to be developed for managing spent fuel elements from research reactors on the basis of the interim storage technology existing in Germany, in order to make the transition to direct disposal possible in the long term. This final report describes the studies for the spent fuel management concept as well as the development of a transport and storage cask for spent fuel elements from research reactors. The concept analyses were based on data of the fuel to be disposed of, as well as the handling conditions for casks at the German research reactors. Due to the quite different conditions for handling of casks at the individual reactors, it was necessary to examine different cask concepts as well as special solutions for loading the casks outside of the spent fuel pools. As a result of these analyses, a concept was elaborated on the basis of a newly developed transport and storage cask as well as a mobile fuel transfer system for the reactor stations, at which a direct loading of the cask is not possible, as the optimal variant. The cask necessary for this concept with the designation CASTOR trademark MTR 2 follows in ist design the tried and tested principles of the CASTOR trademark casks for transport and interim storage of spent LWR fuel. With the CASTOR trademark MTR 2, it is possible to transport and to place into long term interim storage various fuel element types, which have been and are currently used in German research reactors. The technical development of the cask has been completed, the documents for the transport license as type B(U)F package design and for obtaining the storage license at the interim storage facility of Ahaus have been prepared, submitted to the licensing authorities and to a large degree already evaluated positively. The transport license of the CASTOR trademark MTR 2 has been issued for the shipment of VKTA-contents and FRM II compact fuel elements. (orig.)

  14. Nuclear fuel element

    International Nuclear Information System (INIS)

    Hirayama, Satoshi; Kawada, Toshiyuki; Matsuzaki, Masayoshi.

    1980-01-01

    Purpose: To provide a fuel element for reducing the mechanical interactions between a fuel-cladding tube and the fuel element and for alleviating the limits of the operating conditions of a reactor. Constitution: A fuel element having mainly uranium dioxide consists of a cylindrical outer pellet and cylindrical inner pellet inserted into the outer pellet. The outer pellet contains two or more additives selected from aluminium oxide, beryllium oxide, magnesium oxide, silicon oxide, sodium oxide, phosphorus oxide, calcium oxide and iron oxide, and the inner pellet contains nuclear fuel substance solely or one additive selected from calcium oxide, silicon oxide, aluminium oxide, magnesium oxide, zirconium oxide and iron oxide. The outer pellet of the fuel thus constituted is reduced in mechanical strength and also in the mechanical interactions with the cladding tube, and the plastic fluidity of the entire pellet is prevented by the inner pellet increased in the mechanical strength. (Kamimura, M.)

  15. Nuclear fuel element

    International Nuclear Information System (INIS)

    Grossman, L.N.; Levin, H.A.

    1975-01-01

    A nuclear fuel element has disposed therein an alloy having the essential components of nickel, titanium and zirconium, and the alloy reacts with water, water vapor and reactive gases at reactor ambient temperatures. The alloy is disposed in the plenum of the fuel element in the form of particles in a hollow gas permeable container having a multiplicity of openings of size smallr than the size of the particles. The container is preferably held in the spring in the plenum of the fuel element. (E.C.B.)

  16. Cross-sections for homogenized BWR fuel elements in 2d-diffusion theory by 1d-transport calculations

    International Nuclear Information System (INIS)

    Ambrosius, G.

    1980-01-01

    Leakage has a large influence on the thermal spectrum in a fuel rod cell of a BWR and originates: a) from rods with different absorptions and; b) from the different distances to the water gaps. Due to reason a) Gd-rods are treated together with a ring of the homogenized eight nearest neighbours. The often used definition of homogenized cross-sections as the ratio of the integrated reaction rate to the integrated flux proved to be inadequate. This homogenization method is exact as far as the flux is constant over the boundary and as the leakag e during calculating the homogenized cross-sections is similar to that during application. With respect to the condition b) a 1d-transport calculation for the whole fuel element with rings or slabs of homogenized fuel rod cells is performed. With the definition above the flux distribution is that of the fluxes in the moderator regions. The spectrum within each fuel rod cell which includes the leakage is calculated by superimposing at each energy on the flux distribution in the cell the flux at the cell position from the bundle calculation. Changes in the flux ratio between fuel and moderator due to the leakage are taken into account in a final few group 2d-diffusion calculation with fuel and (moderator + cladding) taken separately

  17. 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

  18. 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

  19. Nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    Butterfield, C.E.; Waite, E.

    1982-01-01

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

  20. Nuclear fuel elements

    International Nuclear Information System (INIS)

    Kawada, Toshiyuki; Hirayama, Satoshi; Yoneya, Katsutoshi.

    1980-01-01

    Purpose: To enable load-depending operation as well as moderation for the restriction of operation conditions in the present nuclear reactors, by specifying the essential ingredients and the total weight of the additives to UO 2 fuel substances. Constitution: Two or more additives selected from Al 2 O 3 , B 2 O, CaO, MgO, SiO 2 , Na 2 O and P 2 O 5 are added by the total weight of 2 - 5% to fuel substances consisting of UO 2 or a mixture of UO 2 and PuO 2 . When the mixture is sintered, the strength of the fuel elements is decreased and the fuel-cladding interactions due to the difference in the heat expansion coefficients between the ceramic fuel elements and the metal claddings are decreased to a substantially harmless degree. (Horiuchi, T.)

  1. German physical protection concept for the storage of spent fuel elements in transport and storage casks

    International Nuclear Information System (INIS)

    Weil, L.; Maier, R.

    2005-01-01

    Full text: In Germany, the legal regulations and requirements derived from rules and guidelines for the protection of storage facilities for spent fuel elements from disruptive action or other inference by third parties are structured hierarchically. The Atomic Energy Act constitutes the top level. It is supported by federal ordinances. The next level down is formed by the rules and guidelines. The storage of nuclear fuels may only be authorized, according to the provisions of the Atomic Energy Act, if the required protection from disruptive action or other interference by third parties can be guaranteed as following: it must be possible to prevent any danger to life and health due to a substantial amount of direct radiation or due to the release of a substantial amount of radioactive material; it must be possible to prevent singular or repeated acts of stealing nuclear fuels in such amounts that a critical accumulation can be produced directly without reprocessing and enrichment. Knowing that nuclear installations cannot be protected from every possible interference, physical protection is focused on basic security standards, the so-called design basic threat (DBT), departing from the assumed interference. DBT is regularly reviewed by the competent federal authorities and authorities of the states and are revised on the basis of newly gained knowledge, if necessary, such as in the wake of the terrorist attacks in the U.S. on September 11, 2001. The operator must guarantee and give proof of a sufficient level of physical protection of the plant. The sole physical protection measures implemented by the operator cannot ensure the required protection from other interference by third parties for an unlimited time span. The concept therefore requires additional physical protection measures by the police. (author)

  2. Nuclear fuel element

    International Nuclear Information System (INIS)

    Penrose, R.T.; Thompson, J.R.

    1976-01-01

    A method of protecting the cladding of a nuclear fuel element from internal attack and a nuclear fuel element for use in the core of a nuclear reactor are disclosed. The nuclear fuel element has disposed therein an additive of a barium-containing material and the barium-containing material collects reactive gases through chemical reaction or adsorption at temperatures ranging from room temperature up to fuel element plenum temperatures. The additive is located in the plenum of the fuel element and preferably in the form of particles in a hollow container having a multiplicity of gas permeable openings in one portion of the container with the openings being of a size smaller than the size of the particles. The openings permit gases and liquids entering the plenum to contact the particles. The additive is comprised of elemental barium or a barium alloy containing one or more metals in addition to barium such as aluminum, zirconium, nickel, titanium and combinations thereof. 6 claims, 3 drawing figures

  3. Transport of irradiated nuclear fuel

    International Nuclear Information System (INIS)

    1980-01-01

    In response to public interest in the transport by rail through London of containers of irradiated fuel elements on their way from nuclear power stations to Windscale, the Central Electricity Generating Board and British Rail held three information meetings in London in January 1980. One meeting was for representatives of London Borough Councils and Members of Parliament with a known interest in the subject, and the others were for press, radio and television journalists. This booklet contains the main points made by the principal speakers from the CEGB and BR. (The points covered include: brief description of the fuel cycle; effect of the fission process in producing plutonium and fission products in the fuel element; fuel transport; the fuel flasks; protection against accidents; experience of transporting fuel). (U.K.)

  4. Fuel Element Technical Manual

    Energy Technology Data Exchange (ETDEWEB)

    Burley, H.H. [ed.

    1956-08-01

    It is the purpose of the Fuel Element Technical Manual to Provide a single document describing the fabrication processes used in the manufacture of the fuel element as well as the technical bases for these processes. The manual will be instrumental in the indoctrination of personnel new to the field and will provide a single data reference for all personnel involved in the design or manufacture of the fuel element. The material contained in this manual was assembled by members of the Engineering Department and the Manufacturing Department at the Hanford Atomic Products Operation between the dates October, 1955 and June, 1956. Arrangement of the manual. The manual is divided into six parts: Part I--introduction; Part II--technical bases; Part III--process; Part IV--plant and equipment; Part V--process control and improvement; and VI--safety.

  5. Nuclear fuel elements

    International Nuclear Information System (INIS)

    Ainsworth, K.F.

    1979-01-01

    A nuclear fuel element is described having a cluster of nuclear fuel pins supported in parallel, spaced apart relationship by transverse cellular braces within coaxial, inner and outer sleeves, the inner sleeve being in at least two separate axial lengths, each of the transverse braces having a peripheral portion which is clamped peripherally between the ends of the axial lengths of the inner sleeve. (author)

  6. Nuclear fuel element

    International Nuclear Information System (INIS)

    Yamanaka, Tsuneyasu.

    1976-01-01

    Purpose: To provide a mechanism for the prevention of fuel pellet dislocation in fuel can throughout fuel fablication, fuel transportation and reactor operation. Constitution: A plenum spacer as a mechanism for the prevention of fuel pellet dislocation inserted into a cladding tube comprises split bodies bundled by a frame and an expansion body being capable of inserting into the central cavity of the split bodies. The expansion body is, for example, in a conical shape and the split bodies are formed so that they define in the center portion, when disposed along the inner wall of the cladding tube, a gap capable of inserting the conical body. The plenum spacer is assembled by initially inserting the split bodies in a closed state into the cladding tube after the loading of the pellets, pressing their peripheral portions and then inserting the expansion body into the space to urge the split bodies to the inner surface of the cladding tube. (Kawakami, Y.)

  7. Nuclear fuel elements

    International Nuclear Information System (INIS)

    Obara, Hiroshi.

    1981-01-01

    Purpose: To suppress iodine release thereby prevent stress corrosion cracks in fuel cans by dispersing ferrous oxide at the outer periphery of sintered uranium dioxide pellets filled and sealed within zirconium alloy fuel cans of fuel elements. Constitution: Sintered uranium dioxide pellets to be filled and sealed within a zirconium alloy fuel can are prepared either by mixing ferric oxide powder in uranium dioxide powder, sintering and then reducing at low temperature or by mixing iron powder in uranium dioxide powder, sintering and then oxidizing at low temperature. In this way, ferrous oxide is dispersed on the outer periphery of the sintered uranium dioxide pellets to convert corrosive fission products iodine into iron iodide, whereby the iodine release is suppressed and the stress corrosion cracks can be prevented in the fuel can. (Moriyama, K.)

  8. 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)

  9. Fabrication of flasks for the transport of irradiated Magnox fuel elements

    International Nuclear Information System (INIS)

    Hamill, R.K.; Sandlan, R.

    1979-01-01

    A programme for the manufacture of rectangular box shaped fuel element flasks was undertaken and the paper sets out to describe some of the associated technical work. Alternative ways of making the flask bodies were considered in the light of the specified fracture toughness requirements and the successful achievement of these properties by the use of forgings is described in the paper. Welding to the body forgings, particularly of the fins attached around the four sides was characterised by extremely tight alignment tolerances. The reasons for adopting the method chosen and the success achieved are outlined. Particular attention was paid to preheat and the means of maintaining the requirement throughout manufacture are described. Procedure development work was conducted to establish and qualify the cladding of the top face by Type 347 stainless steel using the submerged arc strip method. A more unusual aspect of the manufacturing sequence was the attachment of thin nickel sheet to the base and cover plate of each flask to provide thermal insulation. The method adopted was spot welding using the MIG process and some of the problems which had to be overcome both in the welding itself and in the subsequent dressing of the spot welds are described. (author)

  10. Fuel element store

    International Nuclear Information System (INIS)

    Wieser, R.

    1987-01-01

    The spherical fuel elements are stored dry in cans. The cans themselves are stacked in parallel storage shafts, which are combined into a rectangular storage space. The storage space is made earthquake-proof by surrounding it with concrete. It consists of a ceiling assembled from several steel parts, which is connected to the floor by support elements. A cooling air ventilation station supplies the individual storage shaft and therefore the cans with cooling air via incoming and outgoing pipes. (DG) [de

  11. Improved 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 and has a metal liner disposed between the cladding and the nuclear fuel material and a high lubricity material in the form of a coating disposed between the liner and the cladding. The liner preferably has a thickness greater than the longest fission product recoil distance and is composed of a low neutron capture cross-section material. The liner is preferably composed of zirconium, an alloy of zirconium, niobium or an alloy of niobium. The liner serves as a preferential reaction site for volatile impurities and fission products and protects the cladding from contact and reaction with such impurities and fission products. The high lubricity material acts as an interface between the liner and the cladding and reduces localized stresses on the cladding due to fuel expansion and cracking of the fuel

  12. Transport of MOX fuel

    International Nuclear Information System (INIS)

    Porter, I.R.; Carr, M.

    1997-01-01

    The regulatory framework which governs the transport of MOX fuel is set out, including packages, transport modes and security requirements. Technical requirements for the packages are reviewed and BNFL's experience in plutonium and MOX fuel transport is described. The safety of such operations and the public perception of safety are described and the question of gaining public acceptance for MOX fuel transport is addressed. The paper concludes by emphasising the need for proactive programmes to improve the public acceptance of these operations. (Author)

  13. Fuel transporting device

    International Nuclear Information System (INIS)

    Shiratori, Hirozo.

    1979-01-01

    Purpose: In a liquid-metal cooled reactor, to reduce the waiting time of fuel handling apparatuses and shorten the fuel exchange time. Constitution: A fuel transporting machine is arranged between a reactor vessel and an out-pile storage tank, thereby dividing the transportation line of the pot for contracting fuel and transporting the same. By assuming such a construction, the flow of fuel transportation which has heretofore been carried out through fuel transportation pipes is not limited to one direction but the take-out of fuels from the reactor and the take-in thereof from the storage tank can be carried out constantly, and much time is not required for fuel exchange. (Kamimura, M.)

  14. 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

  15. 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.

  16. Thermal model of spent fuel transport cask

    International Nuclear Information System (INIS)

    Ahmed, E.E.M.; Rahman, F.A.; Sultan, G.F.; Khalil, E.E.

    1996-01-01

    The investigation provides a theoretical model to represent the thermal behaviour of the spent fuel elements when transported in a dry shipping cask under normal transport conditions. The heat transfer process in the spent fuel elements and within the cask are modeled which include the radiant heat transfer within the cask and the heat transfer by thermal conduction within the spent fuel element. The model considers the net radiant method for radiant heat transfer process from the inner most heated element to the surrounding spent elements. The heat conduction through fuel interior, fuel-clad interface and on clad surface are also presented. (author) 6 figs., 9 refs

  17. Nuclear reactor fuel element

    International Nuclear Information System (INIS)

    D'Eye, R.W.M.; Shennan, J.V.; Ford, L.H.

    1977-01-01

    Fuel element with particles from ceramic fissionable material (e.g. uranium carbide), each one being coated with pyrolitically deposited carbon and all of them being connected at their points of contact by means of an individual crossbar. The crossbar consists of silicon carbide produced by reaction of silicon metal powder with the carbon under the influence of heat. Previously the silicon metal powder together with the particles was kneaded in a solvent and a binder (e.g. epoxy resin in methyl ethyl ketone plus setting agent) to from a pulp. The reaction temperature lies at 1750 0 C. The reaction itself may take place in a nitrogen atmosphere. There will be produced a fuel element with a high overall thermal conductivity. (DG) [de

  18. Nuclear fuel element

    International Nuclear Information System (INIS)

    Hirama, H.

    1978-01-01

    A nuclear fuel element comprises an elongated tube having upper and lower end plugs fixed to both ends thereof and nuclear fuel pellets contained within the tube. The fuel pellets are held against the lower end plug by a spring which is supported by a setting structure. The setting structure is maintained at a proper position at the middle of the tube by a wedge effect caused by spring force exerted by the spring against a set of balls coacting with a tapered member of the setting structure thereby wedging the balls against the inner wall of the tube, and the setting structure is moved free by pushing with a push bar against the spring force so as to release the wedge effect

  19. Vented nuclear fuel element

    International Nuclear Information System (INIS)

    Oguma, M.; Hirose, Y.

    1976-01-01

    A description is given of a vented nuclear fuel element having a plenum for accumulation of fission product gases and plug means for delaying the release of the fission product gases from the plenum, the plug means comprising a first porous body wettable with a liquid metal and a second porous body non-wettable with the liquid metal, the first porous body being impregnated with the liquid metal and in contact with the liquid metal

  20. Integral nuclear fuel element assembly

    International Nuclear Information System (INIS)

    Schluderberg, D. C.

    1985-01-01

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

  1. Transfer flask for hot active fuel elements

    International Nuclear Information System (INIS)

    Aubert, Roger; Moutard, Daniel.

    1980-01-01

    This invention concerns a flask for transporting active fuel elements removed from a nuclear reactor vessel, after only a few days storage and hence cooling, either within a nuclear power station itself or between such a station and a near-by storage area. This containment system is not a flask for conveyance over long and medium distances. Specifically, the invention concerns a transport flask that enables hot fuel elements to be cooled, even in the event of accidents [fr

  2. Storage, handling and internal transport of radioactive materials (fuel elements excepted) in nuclear power plants

    International Nuclear Information System (INIS)

    1983-06-01

    The rule applies to storage and handling as well as to transport within the plant and to the exchange of - solid radioactive wastes, - liquid radioactive wastes, except for those covered by the rule KTA 3603, - radioactive components and parts which are planned to be mounted and dismounted until shutdown of the plant, - radioactive-contaminated tools and appliances, - radioactive preparations. The rule is to be applied within the fenced-in sites of stationary nuclear power plants with LWR or HTR including their transport load halls, as fas as these are situated so as to be approachable from the nuclear power station by local transport systems. (orig./HP) [de

  3. Nuclear fuel element

    International Nuclear Information System (INIS)

    Iwano, Yoshihiko.

    1993-01-01

    Microfine cracks having a depth of less than 10% of a pipe thickness are disposed radially from a central axis each at an interval of less than 100 micron over the entire inner circumferential surface of a zirconium alloy fuel cladding tube. For manufacturing such a nuclear fuel element, the inside of the cladding tube is at first filled with an electrolyte solution of potassium chloride. Then, electrolysis is conducted using the cladding tube as an anode and the electrolyte solution as a cathode, and the inner surface of the cladding tube with a zirconium dioxide layer having a predetermined thickness. Subsequently, the cladding tube is laid on a smooth steel plate and lightly compressed by other smooth steel plate to form microfine cracks in the zirconium dioxide layer on the inner surface of the cladding tube. Such a compressing operation is continuously applied to the cladding tube while rotating the cladding tube. This can inhibit progress of cracks on the inner surface of the cladding tube, thereby enabling to prevent failure of the cladding tube even if a pellet/cladding tube mechanical interaction is applied. Accordingly, reliability of the nuclear fuel elements is improved. (I.N.)

  4. Transportation of nuclear fuel

    International Nuclear Information System (INIS)

    Prowse, D.R.

    1979-01-01

    Shipment of used fuel from nuclear reactors to a central fuel management facility is discussed with particular emphasis on the assessment of the risk to the public due to these shipments. The methods of transporting used fuel in large shipping containers is reviewed. In terms of an accident scenario, it is demonstrated that the primary risk of transport of used fuel is due to injury and death in common road accidents. The radiological nature of the used fuel cargo is, for all practical purposes, an insignificant factor in the total risk to the public. (author)

  5. Nuclear fuel element

    Energy Technology Data Exchange (ETDEWEB)

    Armijo, J S; Coffing, L F

    1979-04-05

    The fuel element with circular cross-section for BWR and PWR consists of a core surrounded by a compound jacket container where there is a gap between the core and jacket during operation in the reactor. The core consists of U, Pu, Th compounds and mixtures of these. The compound jacket consists of zircaloy 2 or 4. In order to for example prevent the corrosion of the compound jacket, its inner surface has a metal barrier with smaller neutron absorbers than the jacket material in the form of a zirconium sponge. The zirconium of this metal barrier has impurities of various elements in the order of magnitude of 1000 to 5000 ppm. The oxygen content is in the range of 200 to 1200 ppm and the thickness of the metal barrier is 1-30% of the thickness of the jacket.

  6. Improved nuclear fuel element

    International Nuclear Information System (INIS)

    1980-01-01

    The invention is of a nuclear fuel element which comprises a central core of a body of nuclear fuel material selected from the group consisting of compounds of uranium, plutonium, thorium and mixtures thereof, and an elongated composite cladding container comprising a zirconium alloy tube containing constituents other than zirconium in an amount greater than about 5000 parts per million by weight and an undeformed metal barrier of moderate purity zirconium bonded to the inside surface of the alloy tube. The container encloses the core so as to leave a gap between the container and the core during use in a nuclear reactor. The metal barrier is of moderate purity zirconium with an impurity level on a weight basis of at least 1000ppm and less than 5000ppm. Impurity levels of specific elements are given. Variations of the invention are also specified. The composite cladding reduces chemical interaction, minimizes localized stress and strain corrosion and reduces the likelihood of a splitting failure in the zirconium alloy tube. Other benefits are claimed. (U.K.)

  7. Hydrogen in CANDU fuel elements

    International Nuclear Information System (INIS)

    Sejnoha, R.; Manzer, A.M.; Surette, B.A.

    1995-01-01

    Unirradiated and irradiated CANDU fuel cladding was tested to compare the role of stress-corrosion cracking and of hydrogen in the development of fuel defects. The results of the tests are compared with information on fuel performance in-reactor. The role of hydriding (deuteriding) from the coolant and from the fuel element inside is discussed, and the control of 'hydrogen gas' content in the element is confirmed as essential for defect-free fuel performance. Finally, implications for fuel element design are discussed. (author)

  8. Instrumentation of fuel elements and fuel plates

    International Nuclear Information System (INIS)

    Durand, J.P.; Fanjas, Y.

    1993-01-01

    When controlling the behaviour of a reactor or developing a new fuel concept, it is of utmost interest to have the possibility to confirm the thermohydraulic calculations by actual measurements in the fuel elements or in the fuel plates. For years, CERCA has developed the technology and supplied its customers with fuel elements equipped with pressure or temperature measuring devices according to the requirements. Recent customer projects have led to the development of a new method to introduce thermocouples directly into the fuel plate meat instead of the cladding. The purpose of this paper is to review the various instrumentation possibilities available at CERCA. (author)

  9. Instrumentation of fuel elements and fuel plates

    International Nuclear Information System (INIS)

    Durand, J.P.; Fanjas, Y.

    1994-01-01

    When controlling the behaviour of a reactor or developing a new fuel concept, it is of utmost interest to have the possibility to confirm the thermohydraulic calculations by actual measurements in the fuel elements or in the fuel plates. For years, CERCA has developed the technology and supplied its customers with fuel elements equipped with pressure or temperature measuring devices according to the requirements. Recent customer projects have lead to the development of a new method to introduce thermocouples directly into the fuel plate meat instead of the cladding. The purpose of this paper is to review the various instrumentation possibilities available at CERCA. (author)

  10. 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.)

  11. Transportation of radioactive elements

    International Nuclear Information System (INIS)

    Thubert, Francis; Rentien, Guy; Jacquet, Michel

    1981-01-01

    The production and marketing of artificial radioactive elements engaged in by the 'Office des Rayonnements Ionisants' requires the use of specially designed packagings and assorted means of transport. The authors begin by describing the different kinds of products involved and the forms of packagings needed, and go on to discuss the various means of transport used, underlining the fact that, in terms of number and gravity, the incidents that have occurred to date have indeed been few and far between [fr

  12. Nuclear reactor fuel element splitter

    International Nuclear Information System (INIS)

    Yeo, D.

    1976-01-01

    A method and apparatus are disclosed for removing nuclear fuel from a clad fuel element. The fuel element is power driven past laser beams which simultaneously cut the cladding lengthwise into at least two longitudinal pieces. The axially cut lengths of cladding are then separated, causing the nuclear fuel contained therein to drop into a receptacle for later disposition. The cut lengths of cladding comprise nuclear waste which is disposed of in a suitable manner. 6 claims, 10 drawing figures

  13. Nuclear fuel element

    International Nuclear Information System (INIS)

    Armijo, J.S.

    1977-01-01

    A nuclear fuel element for use in the core of a nuclear reactor is disclosed which has a composite cladding having a substrate, a metal barrier metallurgically bonded to the inside surface of the substrate and an inner layer metallurgically bonded to the inside surface of the metal barrier. In this composite cladding, the inner layer and the metal barrier shield the substrate from any impurities or fission products from the nuclear fuel material held within the composite cladding. The metal barrier forms about 1 to about 4 percent of the thickness of the cladding and is comprised of a metal selected from the group consisting of niobium, aluminum, copper, nickel, stainless steel, and iron. The inner layer and then the metal barrier serve as reaction sites for volatile impurities and fission products and protect the substrate from contact and reaction with such impurities and fission products. The substrate and the inner layer of the composite cladding are selected from conventional cladding materials and preferably are a zirconium alloy. Also in a preferred embodiment the substrate and the inner layer are comprised of the same material, preferably a zirconium alloy. 19 claims, 2 figures

  14. Approximate solutions of pulse transport in turbulent flow in narrow fuel element bundle geometries, using the FE method

    International Nuclear Information System (INIS)

    Kaiser, H.G.

    1985-01-01

    The author is concerned with the flow conditions in case of narrow fuel element grids of pressurised-water reactors. Starting from the mathematical formulation of the flow processes for incompressible, isothermal flows, models of the turbulence characteristics are being developed. Besides turbulence models, and network structure the finite element method is treated as numeric solution process. Finally the results are summarized and discussed. (HAG) [de

  15. Rack for nuclear fuel elements

    International Nuclear Information System (INIS)

    Rubinstein, H.J.; Gordon, C.B.; Robison, A.; Clark, P.M.

    1977-01-01

    Disclosed is a rack for storing spent nuclear fuel elements in which a plurality of aligned rows of upright enclosures of generally square cross-sectional areas contain vertically disposed spent fuel elements. Each fuel element is supported at the lower end thereof by a respective support that rests on the floor of the spent fuel pool for a nuclear power plant. An open rack frame is employed as an upright support for the enclosures containing the spent fuel elements. Legs at the lower corners of the frame rest on the floor of the pool to support the frame. In one exemplary embodiment, the support for the fuel element is in the form of a base on which a fuel element rests and the base is supported by legs. In another exemplary embodiment, each fuel element is supported on the pool floor by a self-adjusting support in the form of a base on which a fuel element rests and the base rests on a ball or swivel joint for self-alignment. The lower four corners of the frame are supported by legs adjustable in height for leveling the frame. Each adjustable frame leg is in the form of a base resting on the pool floor and the base supports a threaded post. The threaded post adjustably engages a threaded column on which rests the lower end of the frame. 16 claims, 14 figures

  16. Fuel cell water transport

    Science.gov (United States)

    Vanderborgh, Nicholas E.; Hedstrom, James C.

    1990-01-01

    The moisture content and temperature of hydrogen and oxygen gases is regulated throughout traverse of the gases in a fuel cell incorporating a solid polymer membrane. At least one of the gases traverses a first flow field adjacent the solid polymer membrane, where chemical reactions occur to generate an electrical current. A second flow field is located sequential with the first flow field and incorporates a membrane for effective water transport. A control fluid is then circulated adjacent the second membrane on the face opposite the fuel cell gas wherein moisture is either transported from the control fluid to humidify a fuel gas, e.g., hydrogen, or to the control fluid to prevent excess water buildup in the oxidizer gas, e.g., oxygen. Evaporation of water into the control gas and the control gas temperature act to control the fuel cell gas temperatures throughout the traverse of the fuel cell by the gases.

  17. Nuclear fuel elements design, fabrication and performance

    CERN Document Server

    Frost, Brian R T

    1982-01-01

    Nuclear Fuel Elements: Design, Fabrication and Performance is concerned with the design, fabrication, and performance of nuclear fuel elements, with emphasis on fast reactor fuel elements. Topics range from fuel types and the irradiation behavior of fuels to cladding and duct materials, fuel element design and modeling, fuel element performance testing and qualification, and the performance of water reactor fuels. Fast reactor fuel elements, research and test reactor fuel elements, and unconventional fuel elements are also covered. This volume consists of 12 chapters and begins with an overvie

  18. Thermal insulation of fuel elements

    International Nuclear Information System (INIS)

    Dubrovcak, P.; Pec, V.; Pitonak, J.

    1978-01-01

    The claim of the invention concerns thermal insulation of fuel elements heated for measurement of uranium fuel physical properties. For this, layers of aluminium film and of glass fibre are wound onto the inner tube of the element cladding. The space between the inner and the outer tubes is evacuated and the tubes are spaced using spacer wires. (M.S.)

  19. Increased burnup of fuel elements

    International Nuclear Information System (INIS)

    Ahlf, J.

    1983-01-01

    The specialists' group for fuel elements of the Kerntechnische Gesellschaft e.V. held a meeting on ''Increased Burnup of Fuel Elements'' on 9th and 10th of November 1982 at the GKSS Research Center Geesthacht. Most papers dealt with the problems of burnup increase of fuel elements for light water reactors with respect to fuel manufacturing, power plant operation and reprocessing. Review papers were given on the burnup limits for high temperature gas cooled reactors and sodium fast breeder reactors. The meeting ended with a presentation of the technical equipment of the hot laboratory of the GKSS and the programs which are in progress there. (orig.) [de

  20. Impact of thermal conductivity models on the coupling of heat transport, oxygen diffusion, and deformation in (U, Pu)O nuclear fuel elements

    Science.gov (United States)

    Mihaila, Bogdan; Stan, Marius; Crapps, Justin; Yun, Di

    2013-02-01

    We study the coupled thermal transport, oxygen diffusion, and thermal expansion in a generic nuclear fuel rod consisting of a (U) fuel pellet separated by a helium gap from zircaloy cladding. Steady-state and time-dependent finite-element simulations with a variety of initial- and boundary-value conditions are used to study the effect of the Pu content, y, and deviation from stoichiometry, x, on the temperature and deformation profiles in this fuel element. We find that the equilibrium radial temperature and deformation profiles are most sensitive to x at small values of y. For larger values of y, the effects of oxygen and Pu content are equally important. Following a change in the heat-generation rate, the centerline temperature, the radial deformation of the fuel pellet, and the centerline deviation from stoichiometry track each other closely in (U,Pu)O, as the characteristic time scales of the heat transport and oxygen diffusion are similar. This result is different from the situation observed in the case of UO fuels.

  1. Fuel element services

    International Nuclear Information System (INIS)

    Marta, H.; Alvarez, P.; Jimenez, J.

    2006-01-01

    Refuelling outages comprise a number of maintenance tasks scheduled long in advance to assure a reliable operation throughout the next cycle and, in the long run, a safer and more efficient plant. Most of these tasks are routine service of mechanical and electrical system and likewise fuel an be considered a critical component as to handling, inspection, cleaning and repair. ENUSA-ENWESA AIE has been working in this area since 1995 growing from fuel repair to a more integrated service that includes new and spent fuel handling, inserts, failed fuel rod detection systems, ultrasonic fuel cleaning, fuel repair and a comprehensive array of inspection and tests related to the reliability of the mechanical components in the fuel assembly, all this, performed in compliance with quality, safety, health physics and any other nuclear standard. (Author)

  2. Solution of the two dimensional diffusion and transport equations in a rectangular lattice with an elliptical fuel element using Fourier transform methods: One and two group cases

    International Nuclear Information System (INIS)

    Williams, M.M.R.; Hall, S.K.; Eaton, M.D.

    2014-01-01

    Highlights: • A rectangular reactor cell with an elliptical fuel element. • Solution of transport and diffusion equations by Fourier expansion. • Numerical examples showing convergence. • Two group cell problems. - Abstract: A method for solving the diffusion and transport equations in a rectangular lattice cell with an elliptical fuel element has been developed using a Fourier expansion of the neutron flux. The method is applied to a one group model with a source in the moderator. The cell flux is obtained and also the associated disadvantage factor. In addition to the one speed case, we also consider the two group equations in the cell which now become an eigenvalue problem for the lattice multiplication factor. The method of solution relies upon an efficient procedure to solve a large set of simultaneous linear equations and for this we use the IMSL library routines. Our method is compared with the results from a finite element code. The main drawback of the problem arises from the very large number of terms required in the Fourier series which taxes the storage and speed of the computer. Nevertheless, useful solutions are obtained in geometries that would normally require the use of finite element or analogous methods, for this reason the Fourier method is useful for comparison with that type of numerical approach. Extension of the method to more intricate fuel shapes, such as stars and cruciforms as well as superpositions of these, is possible

  3. Experience gathered from the transport of a fuel element prototype of the CNA-II (Atucha-II nuclear power plant) type

    International Nuclear Information System (INIS)

    Pastorini, A.; Belinco, C.G.; El Bis, E.D.; Sacchi, M.A.; Mayans, C.O.; Martin Ghiselli, A.; Marcora, G.R.

    1990-01-01

    This work describes the needs to materialize the transport of a fuel element prototype of the CNA-II (Atucha-II nuclear power plant) type, under special conditions, from the Fabrication Pilot Plant sited at the Constituyentes Atomic Center and the Ezeiza Atomic Center, for its subsequent analysis at the High Pressure Experimental Loop. The special conditions under which the transport has been made responded to the fact that the prototype presents a fragile adjustment between rods and separators, necessary to be preserved. (Author) [es

  4. REACTOR FUEL ELEMENTS TESTING CONTAINER

    Science.gov (United States)

    Whitham, G.K.; Smith, R.R.

    1963-01-15

    This patent shows a method for detecting leaks in jacketed fuel elements. The element is placed in a sealed tank within a nuclear reactor, and, while the reactor operates, the element is sparged with gas. The gas is then led outside the reactor and monitored for radioactive Xe or Kr. (AEC)

  5. 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

  6. Fuels processing for transportation fuel cell systems

    Science.gov (United States)

    Kumar, R.; Ahmed, S.

    Fuel cells primarily use hydrogen as the fuel. This hydrogen must be produced from other fuels such as natural gas or methanol. The fuel processor requirements are affected by the fuel to be converted, the type of fuel cell to be supplied, and the fuel cell application. The conventional fuel processing technology has been reexamined to determine how it must be adapted for use in demanding applications such as transportation. The two major fuel conversion processes are steam reforming and partial oxidation reforming. The former is established practice for stationary applications; the latter offers certain advantages for mobile systems and is presently in various stages of development. This paper discusses these fuel processing technologies and the more recent developments for fuel cell systems used in transportation. The need for new materials in fuels processing, particularly in the area of reforming catalysis and hydrogen purification, is discussed.

  7. 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.)

  8. Spent fuel transportation problems

    International Nuclear Information System (INIS)

    Kondrat'ev, A.N.; Kosarev, Yu.A.; Yulikov, E.A.

    1977-01-01

    In this paper, problems of transportation of nuclear spent fuel to reprocessing plants are discussed. The solutions proposed are directed toward the achievement of the transportation as economic and safe as possible. The increase of the nuclear power plants number in the USSR and the great distances between these plants and the reprocessing plants involve an intensification of the spent fuel transportation. Higher burnup and holdup time reduction cause the necessity of more bulky casks. In this connection, the economic problems become still more important. One of the ways of the problem solution is the development of rational and cheap cask designs. Also, the enforcement in the world of the environmental and personnel health protection requires to increase the transportation reliability and safety. The paper summarizes safe transportation rules with clarifying the following questions: the increase of the transport unit quantity of the spent fuel; rational shipment organization that minimizes vehicle turnover cycle duration; development of the reliable calculation methods to determine strength, thermal conditions and nuclear safety of transport packaging as applied to the vehicles of high capacity; maximum unification of vehicles, calculation methods and documents; and cask testing on models and in pilot scale on specific test rigs to assure that they meet the international safe fuel shipment rules. Besides, some considerations on the choice and use of structural materials for casks are given, and problems of manufacturing such casks from uranium and lead are considered, as well as problems of the development of fireproof shells, control instrumentation, vehicles decontamination, etc. All the problems are considered from the point of view of normal and accidental shipment conditions. Conclusions are presented [ru

  9. Nuclear reactor fuel element assemblies

    International Nuclear Information System (INIS)

    Krawiec, D.M.; Bevilacqua, F.

    1974-01-01

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

  10. Fuel element box inspection device

    International Nuclear Information System (INIS)

    Ortmayer, R.M.; Pick, W.

    1985-01-01

    The invention concerns a device for inspecting the outer geometry of a long fuel element box by measuring the surface contours over its longitudinal crossection and along its length by sensors. These are kept in a sledge which can be moved along the fuel element guide in a slot guide. The measurement signals reach an evaluation device outside the longitudinal box. (orig./HP) [de

  11. Nuclear fuel elements and assemblies

    International Nuclear Information System (INIS)

    Saito, Shozo; Maki, Hideo.

    1982-01-01

    Purpose: To facilitate the attainment of the uranium enrichment or gadolinia enrichment of a pellet filled in a fuel element. Constitution: The axial length of a pellet filled in a fuel element is set to predetermined sizes according to the uranium enrichment factor, gadolinia enrichment or their combination. Thus, the uranium enrichment factor or gadolinia enrichment can be identified by attaining the axial length of the pellet by using such a pellt. (Kamimura, M.)

  12. Agricultural transportation fuels

    International Nuclear Information System (INIS)

    Anon.

    1993-01-01

    The recommendations on the title subject are focused on the question whether advantages and disadvantages of agricultural fuels compared to fossil fuels justify the Dutch policy promotion of the use of agricultural products as basic materials for agricultural fuels. Attention is paid to energetic, environmental and economical aspects of both fuel types. Four options to apply agricultural transportation fuels are discussed: (1) 10% bio-ethanol in euro-unleaded gasoline for engines of passenger cars, equipped with a three-way catalyst; (2) the substitution of 15% methyl tertiair butyl ether (MTBE) by ethyl tertiair butyl ether (ETBE) as a substituent for lead in unleaded super plus gasoline (Sp 98) for engines of passenger cars, equipped with a three-way catalyst; (3) 50% KME (rapeseed oil ester) in low-sulfur diesel (0.05%S D) for engines of vans without a catalyst; and (4) the substitution of 0.05% S D by bio-ethanol or KME for buses with fuel-adjusted engines, equipped with a catalyst. Also the substitution by liquefied petroleum gas (LPG), compressed natural gas (CNG) or E 95 was investigated in option four. Each of the options investigated can contribute to a reduction of the use of fossil energy and the environmental effects of the use of fossil fuels, although some environmental effects from agricultural fuels must be taken into consideration. It is recommended to seriously pay attention to the promotion of agricultural fuels, not only in the Netherlands, but also in an international context. Policy instruments to be used in the stimulation of the use of such fuels are the existing European Community subsidies on fallow lands, exemption of the European Community energy levy, and the use of tax differentiation. Large-scale demonstration projects must be started to quantify hazardous emissions and to solve still existing technical problems. 8 figs., 3 tabs., refs., 4 appendices

  13. Nuclear fuel element

    International Nuclear Information System (INIS)

    Knowles, A.N.

    1979-01-01

    A nuclear fuel-containing body for a high temperature gas cooled nuclear reactor is described which comprises a flat plate in which the nuclear fuel is contained as a dispersion of fission product-retaining coated fuel particles in a flat sheet of graphitic or carbonaceous matrix material. The flat sheet is clad with a relatively thin layer of unfuelled graphite bonded to the sheet by being formed initially from a number of separate preformed graphitic artefacts and then platen-pressed on to the exterior surfaces of the flat sheet, both the matrix material and the artefacts being in a green state, to enclose the sheet. A number of such flat plates are supported edge-on to the coolant flow in the bore of a tube made of neutron moderating material. Where a number of tiers of plates are superimposed on one another, the abutting edges are chamfered to reduce vibration. (author)

  14. Quality assurance of fuel elements

    International Nuclear Information System (INIS)

    Hoerber, J.

    1980-01-01

    The quality assurance activities for reactor fuel elements are based on a quality assurance system which implies the requirements resulting from the specifications, regulations of the authorities, national standards and international rules and regulations. The quality assurance related to production of reactor fuel will be shown for PWR fuel elements in all typical fabrication steps as conversion into UO 2 -powder, pelletizing, rodmanufacture and assembling. A wide range of destructive and nondestructive techniques is applied. Quality assurance is not only verified by testing techniques but also by process monitoring by means of parameter control in production and testing procedures. (RW)

  15. Fuel element tomography by gammametry

    International Nuclear Information System (INIS)

    Simonet, G.; Pineira, T.

    1982-03-01

    As from transversal gamma determinations of a cylindrical fuel element, the TOMOGAM program reconstitutes the distribution of fission products in a section. This direct, fast and non destructive method, makes it possible to have access to the behaviour of the fuel at any time: - the soluble fission products in the matrix represent the fuel itself and the distribution of the fissions, - the migrating elements inform on the temperature reached in accordance with the permitted powers, - the volatile nuclides build up in particular points where physical-chemical phenomena of fuel-cladding interaction are liable to corrode the latter. Hence, gamma spectrometry extends its possibilities of analysis relative to the performance of reactor elements [fr

  16. Fuel cells in transportation

    Energy Technology Data Exchange (ETDEWEB)

    Erdmann, G [Technische Univ., Berlin (Germany); Hoehlein, B [Research Center Juelich (Germany)

    1996-12-01

    A promising new power source for electric drive systems is the fuel cell technology with hydrogen as energy input. The worldwide fuel cell development concentrates on basic research efforts aiming at improving this new technology and at developing applications that might reach market maturity in the very near future. Due to the progress achieved, the interest is now steadily turning to the development of overall systems such as demonstration plants for different purposes: electricity generation, drive systems for road vehicles, ships and railroads. This paper does not present results concerning the market potential of fuel cells in transportation but rather addresses some questions and reflections that are subject to further research of both engineers and economists. Some joint effort of this research will be conducted under the umbrella of the IEA Implementing Agreement 026 - Annex X, but there is a lot more to be done in this challenging but also promising fields. (EG) 18 refs.

  17. Reactor fuel element and fuel assembly

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  18. Fast breeder fuel element development

    International Nuclear Information System (INIS)

    Marth, W.; Muehling, G.

    1983-08-01

    This report is a compilation of the papers which have been presented during a seminar ''Fast Breeder Fuel Element Development'' held on November 15/16, 1982 at KfK. The papers give a survey of the status, of the obtained results and of the necessary work, which still has to be done in the frame of various development programmes for fast breeder fuel elements. In detail the following items were covered by the presentations: - the requirements and boundary conditions for the design of fuel pins and elements both for the reference concept of the SNR 300 core and for the large, commercial breeder type of the future (presentation 1,2 and 6); - the fabrication, properties and characterization of various mixed oxide fuel types (presentations 3,4 and 5); - the operational fuel pin behaviour, limits of different design concepts and possible mechanism for fuel pin failures (presentations (7 and 8); - the situation of cladding- and wrapper materials development especially with respect to the high burn-up values of commercial reactors (presentations 9 and 10); - the results of the irradiation experiments performed under steady-state and non-stationary operational conditions and with failed fuel pins (presentations 11, 12, 13 and 14). (orig./RW) [de

  19. Apparatus for locating defective nuclear fuel elements

    International Nuclear Information System (INIS)

    Lawrie, W.E.

    1979-01-01

    An ultrasonic search unit for locating defective fuel elements within a fuel assembly used in a water cooled nuclear reactor is presented. The unit is capable of freely traversing the restricted spaces between the fuel elements

  20. Spacer grid for fuel elements

    International Nuclear Information System (INIS)

    Hensolt, T.; Huenner, M.; Rau, P.; Veca, A.

    1978-01-01

    The spacer grid for fuel elements of a gas-cooled fast breeder reactor (but also for PWRs and BWRs) consists of a lattice field with dodecagonal meshes. These meshes are formed by three each adjacent hexagons grouped arround a central axis. The pairs of legs extending into the dodecagon and being staggered by 120 0 are designed as knubs with inclined abutting surfaces for the fuel rods. By this means there is formed a three-point bearing for centering the fuel rods. The spacer grid mentioned above is rough-worked from a single disc- resp. plate-shaped body (unfinished piece). (DG) [de

  1. Spacer grid for fuel elements

    International Nuclear Information System (INIS)

    Hensolt, T.; Huenner, M.; Rau, P.; Veca, A.

    1980-01-01

    The spacer grid for fuel elements of a gas-cooled fast breeder reactor (but also for PWRs and BWRs) consists of a lattice field with dodecagonal meshes. These meshes are formed by three each adjacent hexagons grouped arround a central axis. The pairs of legs extending into the dodecagon and being staggered by 120 are designed as knubs with inclined abutting surfaces for the fuel rods. By this means there is formed a three-point bearing for centering the fuel rods. The spacer grid mentioned above is rough-worked from a single disc- resp. plate-shaped body (unfinished piece). (orig.)

  2. Nuclear fuel transport and particularly spent fuel transport

    International Nuclear Information System (INIS)

    Lenail, B.

    1986-01-01

    Nuclear material transport is an essential activity for COGEMA linking the different steps of the fuel cycle transport systems have to be safe and reliable. Spent fuel transport is more particularly examined in this paper because the development of reprocessing plant. Industrial, techmical and economical aspects are reviewed [fr

  3. Nuclear fuel element

    International Nuclear Information System (INIS)

    Watarumi, Kazutoshi.

    1992-01-01

    Hollow fuel pellets are piled at multi-stages in a cladding tube to form a pellet stack. A bundle of metal fine wires made of zirconium or an alloy thereof is inserted passing through the hollow portion of each of the hollow pellets over a length of the pellet stack. The metal fine wires are bundled by securing ring at a joining portions of the pellets. Then, the portion between both of adjacent rings is expanded radially and has a spring function biasing in the radial direction. With such a constitution, even if the pellet is expanded radially due to pallet gas swelling, the hollow portion is not closed, and the gas flow channel is ensured. In addition, even if the pellet is cracked due to thermal shocks, the pellet piece is prevented from dropping to the hollow portion. In this case, the thermal conduction between the pellets and the cladding tube is kept satisfactorily by the spring function of the metal wire bundle. (I.N.)

  4. Unified fuel elements development for research reactors

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  5. Monitoring arrangement for vented nuclear fuel elements

    International Nuclear Information System (INIS)

    Campana, R.J.

    1981-01-01

    In a nuclear fuel reactor core, fuel elements are arranged in a closely packed hexagonal configuration, each fuel element having diametrically opposed vents permitting 180 0 rotation of the fuel elements to counteract bowing. A grid plate engages the fuel elements and forms passages for communicating sets of three, four or six individual vents with respective monitor lines in order to communicate vented radioactive gases from the fuel elements to suitable monitor means in a manner readily permitting detection of leakage in individual fuel elements

  6. Fuel element database: developer handbook

    International Nuclear Information System (INIS)

    Dragicevic, M.

    2004-09-01

    The fuel elements database which was developed for Atomic Institute of the Austrian Universities is described. The software uses standards like HTML, PHP and SQL. For the standard installation freely available software packages such as MySQL database or the PHP interpreter from Apache Software Foundation and Java Script were used. (nevyjel)

  7. Automatic welding of fuel elements

    International Nuclear Information System (INIS)

    Briola, J.

    1958-01-01

    The welding process depends on the type of fuel element, the can material and the number of cartridges to be welded: - inert-gas welding (used for G2 and the 1. set of EL3), - inert atmosphere arc welding (used for welding uranium and zirconium), - electronic welding (used for the 2. set of EL3 and the tank of Proserpine). (author) [fr

  8. The so-called 'Castor-Crisis': Transport of spent nuclear fuel elements and German 'Angst'. How to prevent the public relations catastrophe

    International Nuclear Information System (INIS)

    Suess, Werner

    1999-01-01

    Full text: 1. 'Castor-Crisis' - The Real Facts. - the background: radioactive contamination on the surface of transport containers for spent nuclear fuel elements; - legal aspects: transport limit values and notification obligations; - health aspects: radioactive contamination and ionising radiation; - the news media: divergence between technical facts and public perception. 2. 'Castor-Crisis' - The Reactions. 2.1 Technical measures: - 'action plan' of the Federal Ministry for Environment Protection and Reactor Safety; - IT-based European Information and Report System for the transport of nuclear combustibles => 'Transparent Transport Procedures'; - optimisation of decontamination procedures and transport organisation; - simplification of logistics, clearer responsibilities. 2.2 Communications measures: - defense strategy: 'we made a mistake...'; - information campaign: 'we owe you some answers...'; - regaining credibility: public testimonials of employees in newspaper ads, brochures etc.; - regaining credibility: neutral investigation of all relevant circumstances by KPMG. 3. 'Castor-Crisis' - The Lessons: - internal crisis management: improved co-ordination at company and branch level; - pro-active strategy: 'The benefits of nuclear energy' (avoidance Of CO 2 -emissions); - limits of communications; - communications efforts for nuclear energy - the European context. (author)

  9. Electricity as Transportation ``Fuel''

    Science.gov (United States)

    Tamor, Michael

    2013-04-01

    The personal automobile is a surprisingly efficient device, but its place in a sustainable transportation future hinges on its ability use a sustainable fuel. While electricity is widely expected to be such a ``fuel,'' the viability of electric vehicles rests on the validity of three assumptions. First, that the emissions from generation will be significantly lower than those from competing chemical fuels whether `renewable' or fossil. Second, that advances in battery technology will deliver adequate range and durability at an affordable cost. Third, that most customers will accept any functional limitations intrinsic to electrochemical energy storage. While the first two are subjects of active research and vigorous policy debate, the third is treated virtually as a given. Popular statements to the effect that ``because 70% of all daily travel is accomplished in less than 100 miles, mass deployment of 100 mile EVs will electrify 70% of all travel'' are based on collections of one-day travel reports such as the National Household Travel Survey, and so effectively ignore the complexities of individual needs. We have analyzed the day-to-day variations of individual vehicle usage in multiple regions and draw very different conclusions. Most significant is that limited EV range results in a level of inconvenience that is likely to be unacceptable to the vast majority of vehicle owners, and for those who would accept that inconvenience, battery costs must be absurdly low to achieve any economic payback. In contrast, the plug-in hybrid (PHEV) does not suffer range limitations and delivers economic payback for most users at realistic battery costs. More importantly, these findings appear to be universal in developed nations, with labor market population density being a powerful predictor of personal vehicle usage. This ``scalable city'' hypothesis may prove to a powerful predictor of the evolution of transportation in the large cities of the developing world.

  10. Device for manipulating a nuclear reactor fuel element in a fuel element pond containing water

    International Nuclear Information System (INIS)

    Jabsen, F.S.

    1977-01-01

    Using this device a fuel element can be manipulated inside a water filled storage pond for inspection purposes. A transport arrangement which is normally situated above such a pond is modified for this purpose. A crane bridge runs on rails on the upper edge of the pond. A type of trolley runs transversely to the direction of travel of the bridge between 2 wide flange supports forming the crane support. During movement this trolley moves a submerged combination of periscope and TV camera pendant from it at about half the pond height horizontally along the crane support. 2 vehicles move between these on 4 rollers each, on the under flanges of the crane support at spacings of about one fuel element length. A pendant arm of the same length as the periscope dips vertically into the pond from each vehicle. There is a bar of about fuel element length resting on the lower ends of both arms. The surface of a fuel element lying on this bar can be inspected through the periscope on longitudinal travel of the trolley. The bar with the fuel element can be rotated 90 0 downwards into a vertical position after removal of one or more rotating kingpins and release of a rope hanging on the end away from the kingpin. The rope is actuated by a winch on the crane support. The bar has vertical plates at both ends to hold the fuel element in its vertical position. (HP) [de

  11. 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

  12. Transport of encapsulated nuclear fuels

    International Nuclear Information System (INIS)

    Broman, Ulrika; Dybeck, Peter; Ekendahl, Ann-Mari

    2005-12-01

    The transport system for encapsulated fuel is described, including a preliminary drawing of a transport container. In the report, the encapsulation plant is assumed to be located to Oskarshamn, and the repository to Oskarshamn or Forsmark

  13. Grid for a fuel element

    International Nuclear Information System (INIS)

    1975-01-01

    An illustrative embodiment of the invention has one or more corrugations formed in the surface of a fuel element grid for a nuclear reactor. Not only does the corrugation enhance the strength of the grid plate in which it is formed, but it also provides a simple and convenient means for regulating the reactor coolant pressure drop through an appropriate choice of the corrugation depth

  14. Fuel elements of research reactors in China

    International Nuclear Information System (INIS)

    Zhou Yongmao; Chen Dianshan; Tan Jiaqiu

    1987-01-01

    This paper describes the current status of design, fabrication of fuel elements for research reactors in China, emphasis is placed on the technology of fuel elements for the High Flux Engineering Test Reactor (HFETR). (author)

  15. CNG: a potential transport fuel

    International Nuclear Information System (INIS)

    Anon.

    1992-01-01

    Compressed Natural Gas (CNG) is an alternative transport fuel. Advantages of its use are briefly described. Infra structural requirements, if it is to be used in India are outlined. Applications of CNG as transport fuel for buses and trucks in India are discussed. (P.R.K.). 5 refs

  16. Method of measuring distance between fuel element

    International Nuclear Information System (INIS)

    Urata, Megumu.

    1991-01-01

    The distance between fuel elements contained in a pool is measured in a contactless manner even for a narrow distance less than 1 mm. That is, the equipment for measuring the distance between spent fuel elements of a spent fuel assembly in a nuclear reactor comprises a optical fiber scope, a lens, an industrial TV camera and a monitor TV. The top end of the optical fiber scope is inserted between fuel elements to be measured. The state thereof is displayed on the TV screen to measure the distance between the fuel elements. The measured results are compared with a previously formed calibration curve to determine the value between the fuel elements. Then, the distance between the fuel elements can be determined in the pool of a power plant without dismantling the fuel assembly, to investigate the state of the bending and estimate the fuel working life. (I.S.)

  17. Nuclear fuel element and container

    International Nuclear Information System (INIS)

    Grubb, W.T.; King, L.H.

    1981-01-01

    The invention is based on the discovery that a substantial reduction in metal embrittlement or stress corrosion cracking from fuel pellet-cladding interaction can be achieved by the use of a copper layer or liner in proximity to the nuclear fuel, and an intermediate zirconium oxide barrier layer between the copper layer and the zirconium cladding substrate. The intermediate zirconia layer is a good copper diffusion barrier; also, if the zirconium cladding surface is modified prior to oxidation, copper can be deposited by electroless plating. A nuclear fuel element is described which comprises a central core of fuel material and an elongated container using the system outlined above. The method for making the container is again described. It comprises roughening or etching the surface of the zirconium or zirconium alloy container, oxidizing the resulting container, activating the oxidized surface to allow for the metallic coating of such surfaces by electroless deposition and further coating the activated-oxidized surface of the zirconium or zirconium alloy container with copper, iron or nickel or an alloy thereof. (U.K.)

  18. Detector for failed fuel elements

    International Nuclear Information System (INIS)

    Ito, Masaru.

    1979-01-01

    Purpose: To provide automatic monitor for the separation or reactor water and sampling water, in a failed fuel element detector using a sipping chamber. Constitution: A positional detector for the exact mounting of a sipping chamber on a channel box and a level detector for the detection of complete discharge of cooling water in the sipping chamber are provided in the sipping chamber. The positional detector is contacted to the upper end of the channel box and operated when the sipping chamber is correctly mounted to the fuel assemblies. The level detector comprises a float and a limit switch and it is operated when the water in the sipping chamber is discharged by a predetermined amount. Isolation of reactor water and sampling water are automatically monitored by the signal from these two detectors. (Ikeda, J.)

  19. Nuclear reactor fuel element assemblies

    International Nuclear Information System (INIS)

    Raven, L.F.

    1975-01-01

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

  20. Nuclear fuel element leak detection system

    International Nuclear Information System (INIS)

    John, C.D. Jr.

    1978-01-01

    Disclosed is a leak detection system integral with a wall of a building used to fabricate nuclear fuel elements for detecting radiation leakage from the nuclear fuel elements as the fuel elements exit the building. The leak detecting system comprises a shielded compartment constructed to withstand environmental hazards extending into a similarly constructed building and having sealed doors on both ends along with leak detecting apparatus connected to the compartment. The leak detecting system provides a system for removing a nuclear fuel element from its fabrication building while testing for radiation leaks in the fuel element

  1. Transportation of spent nuclear fuels

    International Nuclear Information System (INIS)

    Meguro, Toshiichi

    1976-01-01

    The spent nuclear fuel taken out of reactors is cooled in the cooling pool in each power station for a definite time, then transported to a reprocessing plant. At present, there is no reprocessing plant in Japan, therefore the spent nuclear fuel is shipped abroad. In this paper, the experiences and the present situation in Japan are described on the transport of the spent nuclear fuel from light water reactors, centering around the works in Tsuruga Power Station, Japan Atomic Power Co. The spent nuclear fuel in Tsuruga Power Station was first transported in Apr. 1973, and since then, about 36 tons were shipped to Britain by 5 times of transport. The reprocessing plant in Japan is expected to start operation in Apr. 1977, accordingly the spent nuclear fuel used for the trial will be transported in Japan in the latter half of this year. Among the permission and approval required for the transport of spent nuclear fuel, the acquisition of the certificate for transport casks and the approval of land and sea transports are main tasks. The relevant laws are the law concerning the regulations of nuclear raw material, nuclear fuel and reactors and the law concerning the safety of ships. The casks used in Tsuruga Power Station and EXL III type, and the charging of spent nuclear fuel, the decontamination of the casks, the leak test, land transport with a self-running vehicle, loading on board an exclusive carrier and sea transport are briefly explained. The casks and the ship for domestic transport are being prepared. (Kato, I.)

  2. CERCA's fuel elements instrumentation manufacturing

    International Nuclear Information System (INIS)

    Harbonnier, G.; Jarousse, C.; Pin, T.; Febvre, M.; Colomb, P.

    2005-01-01

    When research and test reactors wish to further understand the Fuel Elements behavior when operating as well as mastering their irradiation conditions, operators carry out neutron and thermo hydraulic analysis. For thermal calculation, the codes used have to be preliminary validated, at least in the range of the reactor safety operational limits. When some further investigations are requested either by safety authorities or for its own reactor needs, instrumented tools are the ultimate solution for providing representative measurements. Such measurements can be conducted for validating thermal calculation codes, at nominal operating condition as well as during transients ones, or for providing numerous and useful data in the frame of a new products qualification program. CERCA, with many years of experience for implanting thermocouples in various products design, states in this poster his manufacturing background on instrumented elements, plates or targets. (author)

  3. Process for changing fuel elements of a water-cooled nuclear reactor

    International Nuclear Information System (INIS)

    Fleischmann, R.; Rau, P.

    1986-01-01

    In order to change fuel elements, a water-filled duct can be installed between the rector pressure vessel and a space for accommodating the fuel elements. The fuel elements are transported there under water by a fuelling machine. The duct is installed as watertight connection closed on all sides between the reactor pressure vessel and a fuel element transport container brought close to it. The fuelling machine works in this duct. (orig./HP) [de

  4. System for assembling nuclear fuel elements

    International Nuclear Information System (INIS)

    1980-01-01

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

  5. Chilean fuel elements fabrication progress report

    International Nuclear Information System (INIS)

    Baeza, J.; Contreras, H.; Chavez, J.; Klein, J.; Mansilla, R.; Marin, J.; Medina, R.

    1993-01-01

    Due to HEU-LEU core conversion necessity for the Chilean MTR reactors, the Fuel Elements Plant is being implemented to LEU nuclear fuel elements fabrication. A glove box line for powder-compact processing designed at CCHEN, which supposed to operate under an automatic control system, is at present under initial tests. Results of first natural uranium fuel plates manufacturing runs are shown

  6. Transuranium element transport in agricultural systems (soil to food chain transfer of nuclear fuel cycle radionuclides). Annual progress report

    International Nuclear Information System (INIS)

    Wallace, A.

    1977-10-01

    Progress is reported on the following research projects: preparation of bibliography covering literature on plant uptake of transuranium elements; development of techniques for growth of agricultural crops in large containers that simulate field conditions; equipment for counting of alpha-emitting transuranium elements; studies on variability in concentration ratio of 241 Am under different environmental conditions; alpha radiation burn in bush beans exposed to 241 Am in solution; constancy of concentration ratio as a measure of plant uptake of 241 Am; growth of radishes in soil with and without DTPA, and radish peel as source of radionuclides; effects of varying levels of DTPA in loam soil on concentration ratio values; and a plant species (Atriplex hymenelytra--desert holly) with high C.R. values and search for other plants with high C.R. values

  7. Transportation of spent MTR fuels

    International Nuclear Information System (INIS)

    Raisonnier, D.

    1997-01-01

    This paper gives an overview of the various aspects of MTR spent fuel transportation and provides in particular information about the on-going shipment of 4 spent fuel casks to the United States. Transnucleaire is a transport and Engineering Company created in 1963 at the request of the French Atomic Energy Commission. The company followed the growth of the world nuclear industry and has now six subsidiaries and affiliated companies established in countries with major nuclear programs

  8. Transportation of spent MTR fuels

    Energy Technology Data Exchange (ETDEWEB)

    Raisonnier, D.

    1997-08-01

    This paper gives an overview of the various aspects of MTR spent fuel transportation and provides in particular information about the on-going shipment of 4 spent fuel casks to the United States. Transnucleaire is a transport and Engineering Company created in 1963 at the request of the French Atomic Energy Commission. The company followed the growth of the world nuclear industry and has now six subsidiaries and affiliated companies established in countries with major nuclear programs.

  9. Transportation fuels of the future?

    International Nuclear Information System (INIS)

    Piel, W.J.

    2001-01-01

    Society is putting more emphasis on the mobile transportation sector to achieve future goals of sustainability and a cleaner environment. To achieve these goals, does society need to jump to a new combination of fuel and vehicle technology or can we just continue to improve on the current fuels and drive train technology that has powered us the past 70 or more years? Do we need to move to more exotic energy conversion technology (fuel cell vehicles?), or can improving fuel properties further allow us to continue using combustion engines to power our vehicles? What fuel properties can still be improved in gasoline and diesel? Besides removing sulfur, should there be less aromatics in fuels? Should aromatics be eliminated? Is there a role for oxygenates in gasoline and diesel? Do blending oxygenates in fuels help or hinder in achieving the environmental goals? Can we and should we reduce our dependency on crude oil for transportation energy? Why have not the previous government-sponsored Alternative Fuel programs displaced crude oil? The marketplace will determine which fuel and vehicle technology combination will eventually be used in the future. Does the information we know today give us insight to this future? This paper will attempt to address some of the key issues and questions on the role fuels may play in that marketplace decision

  10. Transport device of spent fuel

    International Nuclear Information System (INIS)

    Watanabe, Takashi.

    1976-01-01

    Object: To provide a transport device of spent fuel particularly used in a fast breeder, which can enhance accessibility to travelling mechanism portions and exchangeability thereof to facilitate maintenance in the event of failure. Structure: On a travelling floor, which has a function to shield radioactive rays, extending in a direction of transporting spent fuel and being formed with a break passing through in a direction wall thickness, a travelling body is moved along the break. The travelling body has a support rod member mounted thereon, and the support rod member is moved within the break, the support rod member having a fuel support pocket suspended therefrom. (Furukawa, Y.)

  11. Getter for nuclear fuel elements

    International Nuclear Information System (INIS)

    Ross, W.T.; Williamson, H.E.

    1976-01-01

    A nuclear fuel element for use in the core of a nuclear reactor is disclosed and has disposed therein an improved getter capable of gettering reactive gases including a source of hydrogen. The getter comprises a composite with a substrate having thereon a coating capable of gettering reactive gases. The substrate has a greater coefficient of thermal expansion than does the coating, and over a period of time at reactor operating temperatures any protective film on the coating is fractured at various places and fresh portions of the coating are exposed to getter reactive gases. With further passage of time at reactor operating temperatures a fracture of the protective film on the coating will grow into a crack in the coating exposing further portions of the coating capable of gettering reactive gases. 13 claims, 5 drawing figures

  12. Getter for nuclear fuel elements

    International Nuclear Information System (INIS)

    Ross, W.T.; Williamson, H.E.

    1976-01-01

    A nuclear fuel element for use in the core of a nuclear reactor is disclosed and has disposed therein an improved getter capable of gettering reactive gases including a source of hydrogen. The getter comprises a composite with a substrate having thereon a coating capable of gettering reactive gases. The substrate has a greater coefficient of thermal expansion than does the coating, and over a period of time at reactor operating temperatures any protective film on the coating is fractured at various places and fresh portions of the coating are exposed to getter reactive gases. With further passage of time at reactor operating temperatures a fracture of the protective film on the coating will grow into a crack in the coating exposing further portions of the coating capable of gettering reactive gases

  13. Automated Fuel Element Closure Welding System

    International Nuclear Information System (INIS)

    Wahlquist, D.R.

    1993-01-01

    The Automated Fuel Element Closure Welding System is a robotic device that will load and weld top end plugs onto nuclear fuel elements in a highly radioactive and inert gas environment. The system was developed at Argonne National Laboratory-West as part of the Fuel Cycle Demonstration. The welding system performs four main functions, it (1) injects a small amount of a xenon/krypton gas mixture into specific fuel elements, and (2) loads tiny end plugs into the tops of fuel element jackets, and (3) welds the end plugs to the element jackets, and (4) performs a dimensional inspection of the pre- and post-welded fuel elements. The system components are modular to facilitate remote replacement of failed parts. The entire system can be operated remotely in manual, semi-automatic, or fully automatic modes using a computer control system. The welding system is currently undergoing software testing and functional checkout

  14. Spacer for supporting fuel element boxes

    International Nuclear Information System (INIS)

    Wild, E.

    1979-01-01

    A spacer plate unit arranged externally on each side and at a predetermined level of a polygonal fuel element box for mutually supporting, with respect to one another, a plurality of the fuel element boxes forming a fuel element bundle, is formed of a first and a second spacer plate part each having the same length and the same width and being constituted of unlike first and second materials, respectively. The first and second spacer plate parts of the several spacer plate units situated at the predetermined level are arranged in an alternating continuous series when viewed in the peripheral direction of the fuel element box, so that any two spacer plate units belonging to face-to-face oriented sides of two adjoining fuel element boxes in the fuel element bundle define interfaces of unlike materials

  15. Intermodal transportation of spent fuel

    International Nuclear Information System (INIS)

    Elder, H.K.

    1983-09-01

    Concepts for transportation of spent fuel in rail casks from nuclear power plant sites with no rail service are under consideration by the US Department of Energy in the Commercial Spent Fuel Management program at the Pacific Northwest Laboratory. This report identifies and evaluates three alternative systems for intermodal transfer of spent fuel: heavy-haul truck to rail, barge to rail, and barge to heavy-haul truck. This report concludes that, with some modifications and provisions for new equipment, existing rail and marine systems can provide a transportation base for the intermodal transfer of spent fuel to federal interim storage facilities. Some needed land transportation support and loading and unloading equipment does not currently exist. There are insufficient shipping casks available at this time, but the industrial capability to meet projected needs appears adequate

  16. Unification of fuel elements for research reactors

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  17. Experience feedback from the transportation of Framatome fuel assemblies

    International Nuclear Information System (INIS)

    Robin, M.E.; Gaillard, G.; Aubin, C.

    1998-01-01

    Framatome, the foremost world nuclear fuel manufacturer, has for 25 years been delivering fuel elements from its three factories (Dessel, Romans, Pierrelatte) to the various sites in France and abroad (Germany, Sweden, Belgium, China, Korea, South Africa, Switzerland). During this period, Framatome has built up experience and expertise in fuel element transportation by road, rail and sea. In this filed, the range of constraints is very wide: safety and environmental protection constraints; constraints arising from the control and protection of nuclear materials, contractual and financial constraints, media watchdogs. Through the experience feedback from the transportation of FRAMATOME assemblies, this paper addresses all the phases in the transportation of fresh fuel assemblies. (authors)

  18. Trunnions for spent fuel element shipping casks

    International Nuclear Information System (INIS)

    Cooke, B.

    1989-01-01

    Trunnions are used on spent fuel element shipping casks for one or more of a combination of lifting, tilting or securing to a transport vehicle. Within the nuclear transportation industry there are many different philosophies on trunnions, concerning the shape, manufacture, attachment, inspection, maintenance and repair. With the volume of international transport of spent fuel now taking place, it is recognized that problems are occurring with casks in international traffic due to the variance of the philosophies, national standards, and the lack of an international standard. It was agreed through the ISO that an international standard was required to harmonize. It was not possible to evolve an international standard. It was only possible to evolve an international guide. To evolve a standard would mean superseding any existing national standards which already cover particular aspects of trunnions i.e. deceleration forces imposed on trunnions used as tie down features. Therefore the document is a guide only and allows existing national standards to take precedence where they exist. The guide covers design, manufacture, maintenance, repair and quality assurance. The guide covers trunnions used on spent fuel casks transported by road, rail and sea. The guide details the considerations which should be taken account of by cask designers, i.e. stress intensity, design features, inspection and test methods etc. Manufacture, attachment and pre-service testing is also covered. The guide details user requirements which should also be taken account of, i.e. servicing frequency, content, maintenance and repair. The application of quality assurance is described separately although the principles are used throughout the guide

  19. Hydriding failure in water reactor fuel elements

    International Nuclear Information System (INIS)

    Sah, D.N.; Ramadasan, E.; Unnikrishnan, K.

    1980-01-01

    Hydriding of the zircaloy cladding has been one of the important causes of failure in water reactor fuel elements. This report reviews the causes, the mechanisms and the methods for prevention of hydriding failure in zircaloy clad water reactor fuel elements. The different types of hydriding of zircaloy cladding have been classified. Various factors influencing zircaloy hydriding from internal and external sources in an operating fuel element have been brought out. The findings of post-irradiation examination of fuel elements from Indian reactors, with respect to clad hydriding and features of hydriding failure are included. (author)

  20. Gamma irradiation plants using reactor fuel elements

    International Nuclear Information System (INIS)

    Suckow, W.

    1976-11-01

    Recent irradiation plants utilizing fuel elements are described. Criteria for optimizing such plants, evaluation of the plants realized so far, and applications for the facilities are discussed. (author)

  1. Methods of producing transportation fuel

    Science.gov (United States)

    Nair, Vijay [Katy, TX; Roes, Augustinus Wilhelmus Maria [Houston, TX; Cherrillo, Ralph Anthony [Houston, TX; Bauldreay, Joanna M [Chester, GB

    2011-12-27

    Systems, methods, and heaters for treating a subsurface formation are described herein. At least one method for producing transportation fuel is described herein. The method for producing transportation fuel may include providing formation fluid having a boiling range distribution between -5.degree. C. and 350.degree. C. from a subsurface in situ heat treatment process to a subsurface treatment facility. A liquid stream may be separated from the formation fluid. The separated liquid stream may be hydrotreated and then distilled to produce a distilled stream having a boiling range distribution between 150.degree. C. and 350.degree. C. The distilled liquid stream may be combined with one or more additives to produce transportation fuel.

  2. Spent fuel transport in fuel cycle

    International Nuclear Information System (INIS)

    Labrousse, M.

    1977-01-01

    The transport of radioactive substances is a minor part of the fuel cycle because the quantities of matter involved are very small. However the length and complexity of the cycle, the weight of the packing, the respective distances between stations, enrichment plants and reprocessing plants are such that the problem is not negligible. In addition these transports have considerable psychological importance. The most interesting is spent fuel transport which requires exceptionally efficient packaging, especially where thermal and mechanical resistance are concerned. To meet the safety criteria necessary for the protection of both public and users it was decided to use the maximum capacity consistent with rail transport and to avoid coolant fluids under pressure. Since no single type of packing is suitable for all existing stations an effort has been made to standardise handling accessories, and future trands are towards maximum automation. A discussion on the various technical solutions available for the construction of these packing systems is followed by a description of those used for the two types of packaging ordered by COGEMA [fr

  3. Loads on pebble bed fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Teuchert, E.; Maly, V.

    1974-03-15

    A comparison is made of key parameters for multi-recycle pebbles and single-pass once-through (OTTO) pebbles. The parameters analyzed include heat transfer characteristics with burn-up, temperature profiles, power per element as a function of axial position in the core, and burn-up. For the OTTO-scheme, the comparisons addressed the use of the conventional fuel element and the advanced "shell ball" designed to reduce the peak fuel temperature in the center of the fuel element. All studies addressed the uranium-thorium fuel cycle.

  4. International experience in conditioning spent fuel elements

    International Nuclear Information System (INIS)

    Ashton, P.

    1991-04-01

    The purpose of this report is to compile and present in a clear form international experience (USA, Canada, Sweden, FRG, UK, Japan, Switzerland) gained to date in conditioning spent fuel elements. The term conditioning is here taken to mean the handling and packaging of spent fuel elements for short- or long-term storage or final disposal. Plants of a varying nature fall within this scope, both in terms of the type of fuel element treated and the plant purpose eg. experimental or production plant. Emphasis is given to plants which bear some similarity to the concept developed in Germany for direct disposal of spent fuel elements. Worldwide, however, relatively few conditioning plants are in existence or have been conceived. Hence additional plants have been included where aspects of the experience gained are also of relevance eg. plants developed for the consolidation of spent fuel elements. (orig./HP) [de

  5. Worldwide spent fuel transportation logistics

    International Nuclear Information System (INIS)

    Best, R.E.; Garrison, R.F.

    1978-01-01

    This paper presents an overview of the worldwide transportation requirements for spent fuel. Included are estimates of numbers and types of shipments by mode and cask type for 1985 and the year 2000. In addition, projected capital and transportation costs are presented. For the year 1977 and prior years inclusive, there is a cumulative worldwide requirement for approximately 300 MTU of spent fuel storage at away-from-reactor (AFR) facilities. The cumulative requirements for years through 1985 are projected to be nearly 10,000 MTU, and for the years through 2000 the requirements are conservatively expected to exceed 60,000 MTU. These AFR requirements may be related directly to spent fuel transportation requirements. In total nearly 77,000 total cask shipments of spent fuel will be required between 1977 and 2000. These shipments will include truck, rail, and intermodal moves with many ocean and coastal water shipments. A limited number of shipments by air may also occur. The US fraction of these is expected to include 39,000 truck shipments and 14,000 rail shipments. European shipments to regional facilities are expected to be primarily by rail or water mode and are projected to account for 16,000 moves. Pacific basin shipments will account for 4500 moves. The remaining are from other regions. Over 400 casks will be needed to meet the transportation demands. Capital investment is expected to reach $800,000,000 in 1977 dollars. Cumulative transport costs will be a staggering $4.4 billion dollars

  6. 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)

  7. Fuel elements for LWR power plants

    International Nuclear Information System (INIS)

    Roepenack, H.

    1977-01-01

    About five times more expensive than the fabrication of a fuel element is the enriched uranium contained therein; soon the monthly interest charges for the uranium value of a fuel element reload will account for five percent of the fabrication costs, and much more expensive than all this together can it be if reactor operation has to be interrupted because of damaged elements. Thus, quality assurance comes first. (orig.) [de

  8. Experimental research of fuel element reliability

    International Nuclear Information System (INIS)

    Cech, B.; Novak, J.; Chamrad, B.

    1980-01-01

    The rate and extent of the damage of the can integrity for fission products is the basic criterion of reliability. The extent of damage is measurable by the fission product leakage into the reactor coolant circuit. An analysis is made of the causes of the fuel element can damage and a model is proposed for testing fuel element reliability. Special experiments should be carried out to assess partial processes, such as heat transfer and fuel element surface temperature, fission gas liberation and pressure changes inside the element, corrosion weakening of the can wall, can deformation as a result of mechanical interactions. The irradiation probe for reliability testing of fuel elements is described. (M.S.)

  9. High performance nuclear fuel element

    International Nuclear Information System (INIS)

    Mordarski, W.J.; Zegler, S.T.

    1980-01-01

    A fuel-pellet composition is disclosed for use in fast breeder reactors. Uranium carbide particles are mixed with a powder of uraniumplutonium carbides having a stable microstructure. The resulting mixture is formed into fuel pellets. The pellets thus produced exhibit a relatively low propensity to swell while maintaining a high density

  10. Inert matrix fuel in dispersion type fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Savchenko, A.M. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation)]. E-mail: sav@bochvar.ru; Vatulin, A.V. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation); Morozov, A.V. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation); Sirotin, V.L. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation); Dobrikova, I.V. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation); Kulakov, G.V. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation); Ershov, S.A. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation); Kostomarov, V.P. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation); Stelyuk, Y.I. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation)

    2006-06-30

    The advantages of using inert matrix fuel (IMF) as a dispersion fuel in an aluminium alloy matrix are considered, in particular, low temperatures in the fuel centre, achievable high burn-ups, serviceability in transients and an environmentally friendly process of fuel rod fabrication. Two main versions of IMF are under development at A.A. Bochvar Institute, i.e. heterogeneous or isolated distribution of plutonium. The out-of-pile results on IMF loaded with uranium dioxide as plutonium simulator are presented. Fuel elements with uranium dioxide composition fabricated at A.A. Bochvar Institute are currently under MIR tests (RIAR, Dimitrovgrad). The fuel elements reached a burn-up of 88 MW d kg{sup -1} (equivalent to the burn up of the standard uranium dioxide pelletized fuel) without loss of leak-tightness of the cladding. The feasibility of fabricating IMF of these particular types with plutonium dioxide is considered with a view to in-pile irradiation.

  11. Inert matrix fuel in dispersion type fuel elements

    Science.gov (United States)

    Savchenko, A. M.; Vatulin, A. V.; Morozov, A. V.; Sirotin, V. L.; Dobrikova, I. V.; Kulakov, G. V.; Ershov, S. A.; Kostomarov, V. P.; Stelyuk, Y. I.

    2006-06-01

    The advantages of using inert matrix fuel (IMF) as a dispersion fuel in an aluminium alloy matrix are considered, in particular, low temperatures in the fuel centre, achievable high burn-ups, serviceability in transients and an environmentally friendly process of fuel rod fabrication. Two main versions of IMF are under development at A.A. Bochvar Institute, i.e. heterogeneous or isolated distribution of plutonium. The out-of-pile results on IMF loaded with uranium dioxide as plutonium simulator are presented. Fuel elements with uranium dioxide composition fabricated at A.A. Bochvar Institute are currently under MIR tests (RIAR, Dimitrovgrad). The fuel elements reached a burn-up of 88 MW d kg-1 (equivalent to the burn up of the standard uranium dioxide pelletized fuel) without loss of leak-tightness of the cladding. The feasibility of fabricating IMF of these particular types with plutonium dioxide is considered with a view to in-pile irradiation.

  12. Spent fuel element storage facility

    International Nuclear Information System (INIS)

    Ukaji, Hideo; Yamashita, Rikuo.

    1981-01-01

    Purpose: To always keep water level of a spent fuel cask pit equal with water level of spent fuel storage pool by means of syphon principle. Constitution: The pool water of a spent fuel storage pool is airtightly communicated through a pipe with the pool water of a spent fuel cask, and a gate is provided between the pool and the cask. Since cask is conveyed into the cask pit as the gate close while conveying, the pool water level is raised an amount corresponding to the volume of the cask, and water flow through scattering pipe and the communication pipe to the storage pool. When the fuel is conveyed out of the cask, the water level is lowered in the amount corresponding to the volume in the cask pit, and the water in the pool flow through the communication pipe to the cask pit. (Sekiya, K.)

  13. MRT fuel element inspection at Dounreay

    Energy Technology Data Exchange (ETDEWEB)

    Gibson, J.

    1997-08-01

    To ensure that their production and inspection processes are performed in an acceptable manner, ie. auditable and traceable, the MTR Fuel Element Fabrication Plant at Dounreay operates to a documented quality system. This quality system, together with the fuel element manufacturing and inspection operations, has been independently certified to ISO9002-1987, EN29002-1987 and BS5750:Pt2:1987 by Lloyd`s Register Quality Assurance Limited (LRQA). This certification also provides dual accreditation to the relevant German, Dutch and Australian certification bodies. This paper briefly describes the quality system, together with the various inspection stages involved in the manufacture of MTR fuel elements at Dounreay.

  14. Requirements for materials of dispersion fuel elements

    International Nuclear Information System (INIS)

    Samojlov, A.G.; Kashtanov, A.I.; Volkov, V.S.

    1982-01-01

    Requirements for materials of dispersion fuel elements are considered. The necessity of structural and fissile materials compatibility at maximum permissible operation temperatures and temperatures arising in a fuel element during manufacture is pointed out. The fuel element structural material must be ductile, possess high mechanical strength minimum neutron absorption cross section, sufficient heat conductivity, good corrosion resistance in a coolant and radiation resistance. The fissile material must have high fissile isotope concentration, radiation resistance, high thermal conductivity, certain porosity high melting temperature must not change the composition under irradiation

  15. Method for inspecting nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    Jabsen, F.S.

    1979-01-01

    A technique for disassembling a nuclear reactor fuel element without destroying the individual fuel pins and other structural components from which the element is assembled is described. A traveling bridge and trolley span a water-filled spent fuel storage pool and support a strongback. The strongback is under water and provides a working surface on which the spent fuel element is placed for inspection and for the manipulation that is associated with disassembly and assembly. To remove, in a non-destructive manner, the grids that hold the fuel pins in the proper relative positions within the element, bars are inserted through apertures in the grids with the aid of special tools. These bars are rotated to flex the adjacent grid walls and, in this way relax the physical engagement between protruding portions of the grid walls and the associated fuel pins. With the grid structure so flexed to relax the physical grip on the individual fuel pins, these pins can be withdrawn for inspection or replacement as necessary without imposing a need to destroy fuel element components

  16. Transport and reprocessing of irradiated nuclear fuel

    International Nuclear Information System (INIS)

    Lenail, B.

    1981-01-01

    This contribution deals with transport and packaging of oxide fuel from and to the Cogema reprocessing plant at La Hague (France). After a general discussion of nuclear fuel and the fuel cycle, the main aspects of transport and reprocessing of oxide fuel are analysed. (Auth.)

  17. Fundamental aspects of nuclear reactor fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Olander, D.R.

    1976-01-01

    The book presented is designed to function both as a text for first-year graduate courses in nuclear materials and as a reference for workers involved in the materials design and performance aspects of nuclear power plants. The contents are arranged under the following chapter headings: statistical thermodynamics, thermal properties of solids, crystal structures, cohesive energy of solids, chemical equilibrium, point defects in solids, diffusion in solids, dislocations and grain boundaries, equation of state of UO/sub 2/, fuel element thermal performance, fuel chemistry, behavior of solid fission products in oxide fuel elements, swelling due to fission gases, pore migration and fuel restructuring kinetics, fission gas release, mechanical properties of UO/sub 2/, radiation damage, radiation effects in metals, interaction of sodium and stainless steel, modeling of the structural behavior of fuel elements and assemblies. (DG)

  18. Fundamental aspects of nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    Olander, D.R.

    1976-01-01

    The book presented is designed to function both as a text for first-year graduate courses in nuclear materials and as a reference for workers involved in the materials design and performance aspects of nuclear power plants. The contents are arranged under the following chapter headings: statistical thermodynamics, thermal properties of solids, crystal structures, cohesive energy of solids, chemical equilibrium, point defects in solids, diffusion in solids, dislocations and grain boundaries, equation of state of UO 2 , fuel element thermal performance, fuel chemistry, behavior of solid fission products in oxide fuel elements, swelling due to fission gases, pore migration and fuel restructuring kinetics, fission gas release, mechanical properties of UO 2 , radiation damage, radiation effects in metals, interaction of sodium and stainless steel, modeling of the structural behavior of fuel elements and assemblies

  19. Burnup measurements of leader fuel elements

    International Nuclear Information System (INIS)

    Henriquez, C; Navarro, G; Pereda, C

    2000-01-01

    Some time ago the CCHEN authorities decided to produce a set of 50 low enrichment fuel elements. These elements were produced in the PEC (Fuel Elements Plant), located at CCHEN offices in Lo Aguirre. These new fuel elements have basically the same geometrical characteristics of previous ones, which were British and made with raw material from the U.S. The principal differences between our fuel elements and the British ones is the density of fissile material, U-235, which was increased to compensate the reduction in enrichment. Last year, the Fuel Elements Plant (PEC) delivered the shipment's first four (4) fuel elements, called leaders, to the RECH1. A test element was delivered too, and the complete set was introduced into the reactor's nucleus, following the normal routine, but performing a special follow-up on their behavior inside the nucleus. This experimental element has only one outside fuel plate, and the remaining (15) structural plates are aluminum. In order to study the burnup, the test element was taken out of the nucleus, in mid- November 1999, and left to decay until June 2000, when it was moved to the laboratory (High Activity Cell), to start the burnup measurements, with a gamma spectroscopy system. This work aims to show the results of these measurements and in addition to meet the following objectives: (a) Visual test of the plate's general condition; (b) Sipping test of fission products; (c) Study of burn-up distribution in the plate; (d) Check and improve the calculus algorithm; (e) Comparison of the results obtained from the spectroscopy with the ones from neutron calculus

  20. Fuel elements handling device and method

    International Nuclear Information System (INIS)

    Jabsen, F.S.

    1976-01-01

    This invention relates to nuclear equipment and more particularly to methods and apparatus for the non-destructive inspection, manipulation, disassembly and assembly of reactor fuel elements and the like. (author)

  1. Inspection of nuclear fuel transport in Spain

    International Nuclear Information System (INIS)

    Lobo Mendez, J.

    1977-01-01

    The experience acquired in inspecting nuclear fuel shipments carried out in Spain will serve as a basis for establishing the regulations wich must be adhered to for future transports, as the transport of nuclear fuels in Spain will increase considerably within the next years as a result of the Spanish nuclear program. The experience acquired in nuclear fuel transport inspection is described. (author) [es

  2. Apparatus and method for assembling fuel elements

    International Nuclear Information System (INIS)

    Arya, S.P.

    1978-01-01

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

  3. 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

  4. A CAREM type fuel element dynamic analysis

    International Nuclear Information System (INIS)

    Magoia, J.E.

    1990-01-01

    A first analysis on the dynamic behaviour of a fuel element designed for the CAREM nuclear reactor (Central Argentina de Elementos Modulares) was performed. The model used to represent this dynamic behaviour was satisfactorily evaluated. Using primary estimations for some of its numerical parameters, a first approximation to its natural vibrational modes was obtained. Results obtained from fuel elements frequently used in nuclear power plants of the PWR (Pressurized Water Reactors) type, are compared with values resulting from similar analysis. (Author) [es

  5. Fluid pressure method for recovering fuel pellets from nuclear fuel elements

    International Nuclear Information System (INIS)

    John, C.D. Jr.

    1979-01-01

    A method is described for removing fuel pellets from a nuclear fuel element without damaging the fuel pellets or fuel element sheath so that both may be reused. The method comprises holding the fuel element while a high pressure stream internally pressurizes the fuel element to expand the fuel element sheath away from the fuel pellets therein so that the fuel pellets may be easily removed

  6. HTGR fuel element size reduction system

    International Nuclear Information System (INIS)

    Strand, J.B.; Cramer, G.T.

    1978-06-01

    Reprocessing of high-temperature gas-cooled reactor fuel requires development of a fuel element size reduction system. This report describes pilot plant testing of crushing equipment designed for this purpose. The test program, the test results, the compatibility of the components, and the requirements for hot reprocessing are discussed

  7. Safety assessment for Dragon fuel element production

    International Nuclear Information System (INIS)

    Price, M.S.T.

    1963-11-01

    This report shall be the Safety Assessment covering the manufacture of the First Charge of Fuel and Fuel Elements for the Dragon Reactor Experiment. It is issued in two parts, of which Part I is descriptive and Part II gives the Hazards Analysis, the Operating Limitations, the Standing Orders and the Emergency Drill. (author)

  8. Fuel element for a nuclear reactor

    International Nuclear Information System (INIS)

    Tanihiro, Yasunori; Sumita, Isao.

    1970-01-01

    An improved fuel element of the heat pipe type is disclosed in which the fuel element itself is given a heat pipe structure and filled with a coated particle fuel at the section thereof having a capillary tube construction, whereby the particular advantages of heat pipes and coated fuels are combined and utilized to enhance thermal control and reactor efficiency. In an embodiment, the fuel element of the present invention is filled at its lower capillary tube section with coated fuel and at its upper section with a granurated neutron absorber. Both sections are partitioned from the central shaft by a cylindrically shaped wire mesh defining a channel through which the working liquid is vaporized from below and condensed by the coolant external to the fuel element. If the wire mesh is chosen to have a melting point lower than that of the fuel but higher than that of the operating temperature of the heat pipe, the mesh will melt and release the neutron absorbing particles should hot spots develop, thus terminating fission. (Owens, K. J.)

  9. Alternatives to traditional transportation fuels: An overview

    Energy Technology Data Exchange (ETDEWEB)

    1994-06-01

    This report presents the first compilation by the Energy Information Administration (EIA) of information on alternatives to gasoline and diesel fuel. The purpose of the report is: (1) to provide background information on alternative transportation fuels and replacement fuels compared with gasoline and diesel fuel, and (2) to furnish preliminary estimates of alternative transportation fuels and alternative fueled vehicles as required by the Energy Policy Act of 1992 (EPACT), Title V, Section 503, ``Replacement Fuel Demand Estimates and Supply Information.`` Specifically, Section 503 requires the EIA to report annually on: (1) the number and type of alternative fueled vehicles in existence the previous year and expected to be in use the following year, (2) the geographic distribution of these vehicles, (3) the amounts and types of replacement fuels consumed, and (4) the greenhouse gas emissions likely to result from replacement fuel use. Alternative fueled vehicles are defined in this report as motorized vehicles licensed for on-road use, which may consume alternative transportation fuels. (Alternative fueled vehicles may use either an alternative transportation fuel or a replacement fuel.) The intended audience for the first section of this report includes the Secretary of Energy, the Congress, Federal and State agencies, the automobile manufacturing industry, the transportation fuel manufacturing and distribution industries, and the general public. The second section is designed primarily for persons desiring a more technical explanation of and background for the issues surrounding alternative transportation fuels.

  10. Grids for nuclear fuel elements

    International Nuclear Information System (INIS)

    Nicholson, G.

    1980-01-01

    This invention relates to grids for nuclear fuel assemblies with the object of providing an improved grid, tending to have greater strength and tending to offer better location of the fuel pins. It comprises sets of generally parallel strips arranged to intersect to define a structure of cellular form, at least some of the intersections including a strip which is keyed to another strip at more than one point. One type of strip may be dimpled along its length and another type of strip may have slots for keying with the dimples. (Auth.)

  11. Hydraulic modelling of the CARA Fuel element

    International Nuclear Information System (INIS)

    Brasnarof, Daniel O.; Juanico, Luis; Giorgi, M.; Ghiselli, Alberto M.; Zampach, Ruben; Fiori, Jose M.; Yedros, Pablo A.

    2004-01-01

    The CARA fuel element is been developing by the National Atomic Energy Commission for both Argentinean PHWRs. In order to keep the hydraulic restriction in their fuel channels, one of CARA's goals is to keep its similarity with both present fuel elements. In this paper is presented pressure drop test performed at a low-pressure facility (Reynolds numbers between 5x10 4 and 1,5x10 5 ) and rational base models for their spacer grid and rod assembly. Using these models, we could estimate the CARA hydraulic performance in reactor conditions that have shown to be satisfactory. (author) [es

  12. Research reactor fuel transport in the U.K

    Energy Technology Data Exchange (ETDEWEB)

    Panter, R [U.K. Atomic Energy Authority, Harwell (United Kingdom)

    1983-09-01

    This paper describes the containers currently used for transport of fresh or spent fuel elements for Research and Materials Test Reactors in the U.K., their status, operating procedures and some of the practical difficulties. In the U.K., MTR fuel cycle work is almost entirely the responsibility of the U.K. Atomic Energy Authority.

  13. Spent nuclear fuel transport problems

    International Nuclear Information System (INIS)

    Kondrat'ev, A.N.; Kosarev, Yu.A.; Yulikov, E.I.

    1977-01-01

    The paper considers the problems of shipping spent fuel from nuclear power stations to reprocessing plants and also the principal ways of solving these problems with a view to achieving maximum economy and safety in transport. The increase in the number of nuclear power plants in the USSR will entail an intensification of spent-fuel shipments. Higher burnup and the need to reduce cooling time call for heavier and more complex shipping containers. The problem of shipping spent fuel should be tackled comprehensively, bearing in mind the requirements of safety and economy. One solution to these problems is to develop rational and cheap designs of such containers. In addition, the world-wide trend towards more thorough protection of the environment against pollution and of the health of the population requires the devotion of constant attention to improving the reliability and safety of shipments. The paper considers the prospects for nuclear power development in the USSR and in other member countries of the CMEA (1976-1980), the composition and design of some Soviet packaging assemblies, the appropriate cooling time for spent fuel from thermal reactor power stations, procedures for reducing fuel-shipping costs, some methodological problems of container calculation and design, and finally problems of testing and checking containers on test rigs. (author)

  14. Computer simulation of fuel element performance

    Energy Technology Data Exchange (ETDEWEB)

    Sukhanov, G I

    1979-01-01

    The review presents reports made at the Conference on the Bahaviour and Production of Fuel for Water Reactors on March 13-17, 1979. Discussed at the Conference are the most developed and tested calculation models specially evolved to predict the behaviour of fuel elements of water reactors. The following five main aspects of the problem are discussed: general conceptions and programs; mechanical mock-ups and their applications; gas release, gap conductivity and fuel thermal conductivity; analysis of nonstationary processes; models of specific phenomena. The review briefly describes the physical principles of the following models and programs: the RESTR, providing calculation of the radii of zones of columnar and equiaxial grains as well as the radius of the internal cavity of the fuel core; programs for calculation of fuel-can interaction, based on the finite elements method; a model predicting the behaviour of the CANDU-PHW fuel elements in transient conditions. General results are presented of investigations of heat transfer through a can-fuel gap and thermal conductivity of UO/sub 2/ with regard for cracking and gas release of the fuel. Many programs already suit the accepted standards and are intensively tested at present.

  15. Fuel element for a nuclear reactor

    International Nuclear Information System (INIS)

    Rau, P.

    1981-01-01

    Fuel elements which consist of parallel longitudinal fuel rods of circular crossection, can be provided with spiral distance pieces, by which the fuel rods support one another, if they are collected together by an outer enclosure. According to the invention, the enclosure includes several strips extending over a small fraction of the rod length, which are connected together by a skeleton rod instead of a fuel rod. The strips can be composed of flat parts which are connected together by the skeleton rod acting as a hinge. The invention is particularly suitable for breeder or converter reactors. (orig.) [de

  16. Technique for mass-spectrometric determination of moisture content in fuel elements and fuel element claddings

    International Nuclear Information System (INIS)

    Kurillovich, A.N.; Pimonov, Yu.I.; Biryukov, A.S.

    1988-01-01

    A technique for mass-spectroimetric determination of moisture content in fuel elements and fuek claddings in the 2x10 -4 -1.5x10 -2 g range is developed. The relative standard deviation is 0.13. A character of moisture extraction from oxide uranium fuels in the 20-700 deg C temperature range is studied. Approximately 80% of moisture is extracted from the fuels at 300 deg C. The moisture content in fuel elements with granular uranium oxide fuels is measured. Dependence of fuel element moisture content on conditions of hot vacuum drying is shown. The technique permits to optimize the fuel element fabrication process to decrease the moisture content in them. 4 refs.; 3 figs.; 2 tabs

  17. Reactor physics assessment of modified 37-element CANDU fuel bundles

    International Nuclear Information System (INIS)

    Pristavu, R.; Rizoiu, A.

    2016-01-01

    Reducing the central element diameter in order to improve the total flow area of CANDU fuel bundle and redistribute the power density of all remaining elements was studied in Canada and Korea when considering the effect of aging pressure tube diametral creep. The aim of this paper is to study the modified bundle behavior using the transport codes WIMS and DRAGON. In calculations, a WIMS nuclear data library on 172 energy groups was used. 2-D transport calculations were performed with WIMS and DRAGON, leading to similar results in estimated cell parameters. Additionally, 3-D DRAGON calculations were carried on in order to evaluate the local flux distribution shift, as well as the incremental cross sections for supercells containing modified CANDU bundles and reactivity devices. The overall effect of using modified fuel bundles was meaningless for both cell and supercell parameters, thus ensuring this possibility of fuel improvement for thermal-hydraulic purposes only. (authors)

  18. Nuclear fuel element end fitting

    International Nuclear Information System (INIS)

    Jabsen, F.S.

    1980-01-01

    An invention is described whereby end fittings are formed from lattices of mutually perpendicular plates. At the plate intersections, sockets are secured to the end fittings in a manner that permits the longitudinal axes of each of the sockets to align with the respective lines of intersection of the plates. The sockets all protrude above one of the surfaces of the end fitting. Further, a detent is formed in the proturding sides of each of the sockets. Annular grooves are formed in each of the ends of the fuel rods that are to be mounted between the end fittings. The socket detents protrude into the respective annular grooves, thus engaging the grooves and retaining the fuel rods and end fittings in one integral structure. (auth)

  19. Sensor system for fuel transport vehicle

    Science.gov (United States)

    Earl, Dennis Duncan; McIntyre, Timothy J.; West, David L.

    2016-03-22

    An exemplary sensor system for a fuel transport vehicle can comprise a fuel marker sensor positioned between a fuel storage chamber of the vehicle and an access valve for the fuel storage chamber of the vehicle. The fuel marker sensor can be configured to measure one or more characteristics of one or more fuel markers present in the fuel adjacent the sensor, such as when the marked fuel is unloaded at a retail station. The one or more characteristics can comprise concentration and/or identity of the one or more fuel markers in the fuel. Based on the measured characteristics of the one or more fuel markers, the sensor system can identify the fuel and/or can determine whether the fuel has been adulterated after the marked fuel was last measured, such as when the marked fuel was loaded into the vehicle.

  20. Fuel element performance computer modelling

    International Nuclear Information System (INIS)

    Locke, D.H.

    1978-01-01

    The meeting was attended by 88 participants from 17 countries. Altogether 47 papers were presented. The majority of the presentations contained a description of the equations and solutions used to describe and evaluate some of the physical processes taking place in water reactor fuel pins under irradiation. At the same time, particular attention was paid to the ''bench marking'' of the codes wherein solutions arrived at for particular experiments are compared with the results at the experiments

  1. Inserts for nuclear fuel elements

    International Nuclear Information System (INIS)

    Cragg, P.J.

    1982-01-01

    An insert for a nuclear fuel pin which comprises a strip. The strip carries notches, which enable a coding arrangement to be carried on the strip. The notches may be of differing sizes and the coding on the strip includes identification and identification checking data. Each notch on the strip may give rise to a signal pulse which is counted by a detector to avoid errors. (author)

  2. Nuclear fuels and development of nuclear fuel elements

    International Nuclear Information System (INIS)

    Sundaram, C.V.; Mannan, S.L.

    1989-01-01

    Safe, reliable and economic operation of nuclear fission reactors, the source of nuclear power at present, requires judicious choice, careful preparation and specialised fabrication procedures for fuels and fuel element structural materials. These aspects of nuclear fuels (uranium, plutonium and their oxides and carbides), fuel element technology and structural materials (aluminium, zircaloy, stainless steel etc.) are discussed with particular reference to research and power reactors in India, e.g. the DHRUVA research reactor at BARC, Trombay, the pressurised heavy water reactors (PHWR) at Rajasthan and Kalpakkam, and the Fast Breeder Test Reactor (FBTR) at Kalpakkam. Other reactors like the gas-cooled reactors operating in UK are also mentioned. Because of the limited uranium resources, India has opted for a three-stage nuclear power programme aimed at the ultimate utilization of her abundant thorium resources. The first phase consists of natural uranium dioxide-fuelled, heavy water-moderated and cooled PHWR. The second phase was initiated with the attainment of criticality in the FBTR at Kalpakkam. Fast Breeder Reactors (FBR) utilize the plutonium and uranium by-products of phase 1. Moreover, FBR can convert thorium into fissile 233 U. They produce more fuel than is consumed - hence, the name breeders. The fuel parameters of some of the operating or proposed fast reactors in the world are compared. FBTR is unique in the choice of mixed carbides of plutonium and uranium as fuel. Factors affecting the fuel element performance and life in various reactors e.g. hydriding of zircaloys, fuel pellet-cladding interaction etc. in PHWR and void swelling; irradiation creep and helium embrittlement of fuel element structural materials in FBR are discussed along with measures to overcome some of these problems. (author). 15 refs., 9 tabs., 23 figs

  3. Methods of making transportation fuel

    Science.gov (United States)

    Roes, Augustinus Wilhelmus Maria [Houston, TX; Mo, Weijian [Sugar Land, TX; Muylle, Michel Serge Marie [Houston, TX; Mandema, Remco Hugo [Houston, TX; Nair, Vijay [Katy, TX

    2012-04-10

    A method for producing alkylated hydrocarbons is disclosed. Formation fluid is produced from a subsurface in situ heat treatment process. The formation fluid is separated to produce a liquid stream and a first gas stream. The first gas stream includes olefins. The liquid stream is fractionated to produce at least a second gas stream including hydrocarbons having a carbon number of at least 3. The first gas stream and the second gas stream are introduced into an alkylation unit to produce alkylated hydrocarbons. At least a portion of the olefins in the first gas stream enhance alkylation. The alkylated hydrocarbons may be blended with one or more components to produce transportation fuel.

  4. Structural analysis of reactor fuel elements

    International Nuclear Information System (INIS)

    Weeks, R.W.

    1977-01-01

    An overview of fuel-element modeling is presented that traces the development of codes for the prediction of light-water-reactor and fast-breeder-reactor fuel-element performance. It is concluded that although the mathematical analysis is now far advanced, the development and incorporation of mechanistic constitutive equations has not kept pace. The resultant reliance on empirical correlations severely limits the physical insight that can be gained from code extrapolations. Current efforts include modeling of alternate fuel systems, analysis of local fuel-cladding interactions, and development of a predictive capability for off-normal behavior. Future work should help remedy the current constitutive deficiencies and should include the development of deterministic failure criteria for use in design

  5. Reliability analysis of dispersion nuclear fuel elements

    Science.gov (United States)

    Ding, Shurong; Jiang, Xin; Huo, Yongzhong; Li, Lin an

    2008-03-01

    Taking a dispersion fuel element as a special particle composite, the representative volume element is chosen to act as the research object. The fuel swelling is simulated through temperature increase. The large strain elastoplastic analysis is carried out for the mechanical behaviors using FEM. The results indicate that the fission swelling is simulated successfully; the thickness increments grow linearly with burnup; with increasing of burnup: (1) the first principal stresses at fuel particles change from tensile ones to compression ones, (2) the maximum Mises stresses at the particles transfer from the centers of fuel particles to the location close to the interfaces between the matrix and the particles, their values increase with burnup; the maximum Mises stresses at the matrix exist in the middle location between the two particles near the mid-plane along the length (or width) direction, and the maximum plastic strains are also at the above region.

  6. Reliability analysis of dispersion nuclear fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Ding Shurong [Department of Mechanics and Engineering Science, Fudan University, Shanghai 200433 (China)], E-mail: dsr1971@163.com; Jiang Xin [Department of Mechanics and Engineering Science, Fudan University, Shanghai 200433 (China); Huo Yongzhong [Department of Mechanics and Engineering Science, Fudan University, Shanghai 200433 (China)], E-mail: yzhuo@fudan.edu.cn; Li Linan [Department of Mechanics, Tianjin University, Tianjin 300072 (China)

    2008-03-15

    Taking a dispersion fuel element as a special particle composite, the representative volume element is chosen to act as the research object. The fuel swelling is simulated through temperature increase. The large strain elastoplastic analysis is carried out for the mechanical behaviors using FEM. The results indicate that the fission swelling is simulated successfully; the thickness increments grow linearly with burnup; with increasing of burnup: (1) the first principal stresses at fuel particles change from tensile ones to compression ones, (2) the maximum Mises stresses at the particles transfer from the centers of fuel particles to the location close to the interfaces between the matrix and the particles, their values increase with burnup; the maximum Mises stresses at the matrix exist in the middle location between the two particles near the mid-plane along the length (or width) direction, and the maximum plastic strains are also at the above region.

  7. HTGR fuel element structural design consideration

    International Nuclear Information System (INIS)

    Alloway, R.; Gorholt, W.; Ho, F.; Vollman, R.; Yu, H.

    1987-01-01

    The structural design of the large HTGR prismatic core fuel elements involve the interaction of four engineering disciplines: nuclear physics, thermo-hydraulics, structural and material science. Fuel element stress analysis techniques and the development of structural criteria are discussed in the context of an overview of the entire design process. The core of the proposed 2240 MW(t) HTGR is described as an example where the design process was used. Probabilistic stress analysis techniques coupled with probabilistic risk analysis (PRA) to develop structural criteria to account for uncertainty are described. The PRA provides a means for ensuring that the proposed structural criteria are consistant with plant investment and safety risk goals. The evaluation of cracked fuel elements removed from the Fort St. Vrain reactor in the U.S.A. is discussed in the context of stress analysis uncertainty and structural criteria development. (author)

  8. HTGR fuel element structural design considerations

    International Nuclear Information System (INIS)

    Alloway, R.; Gorholt, W.; Ho, F.; Vollman, R.; Yu, H.

    1986-09-01

    The structural design of the large HTGR prismatic core fuel elements involve the interaction of four engineering disciplines: nuclear physics, thermo-hydraulics, structural and material science. Fuel element stress analysis techniques and the development of structural criteria are discussed in the context of an overview of the entire design process. The core of the proposed 2240 MW(t) HTGR is described as an example where the design process was used. Probabalistic stress analysis techniques coupled with probabalistic risk analysis (PRA) to develop structural criteria to account for uncertainty are described. The PRA provides a means for ensuring that the proposed structural criteria are consistent with plant investment and safety risk goals. The evaluation of cracked fuel elements removed from the Fort St. Vrain reactor in the USA is discussed in the context of stress analysis uncertainty and structural criteria development

  9. Fuels and Combustion | Transportation Research | NREL

    Science.gov (United States)

    Fuels and Combustion Fuels and Combustion This is the March 2015 issue of the Transportation and , combustion strategy, and engine design hold the potential to maximize vehicle energy efficiency and performance of low-carbon fuels in internal combustion engines with a whole-systems approach to fuel chemistry

  10. Upgraded HFIR Fuel Element Welding System

    International Nuclear Information System (INIS)

    Sease, John D.

    2010-01-01

    The welding of aluminum-clad fuel plates into aluminum alloy 6061 side plate tubing is a unique design feature of the High Flux Isotope Reactor (HFIR) fuel assemblies as 101 full-penetration circumferential gas metal arc welds (GMAW) are required in the fabrication of each assembly. In a HFIR fuel assembly, 540 aluminum-clad fuel plates are assembled into two nested annular fuel elements 610 mm (24-inches) long. The welding process for the HFIR fuel elements was developed in the early 1960 s and about 450 HFIR fuel assemblies have been successfully welded using the GMAW process qualified in the 1960 s. In recent years because of the degradation of the electronic and mechanical components in the old HFIR welding system, reportable defects in plate attachment or adapter welds have been present in almost all completed fuel assemblies. In October 2008, a contract was awarded to AMET, Inc., of Rexburg, Idaho, to replace the old welding equipment with standard commercially available welding components to the maximum extent possible while maintaining the qualified HFIR welding process. The upgraded HFIR welding system represents a major improvement in the welding system used in welding HFIR fuel elements for the previous 40 years. In this upgrade, the new inner GMAW torch is a significant advancement over the original inner GMAW torch previously used. The innovative breakthrough in the new inner welding torch design is the way the direction of the cast in the 0.762 mm (0.030-inch) diameter aluminum weld wire is changed so that the weld wire emerging from the contact tip is straight in the plane perpendicular to the welding direction without creating any significant drag resistance in the feeding of the weld wire.

  11. Fuel elements and safety engineering goals

    International Nuclear Information System (INIS)

    Schulten, R.; Bonnenberg, H.

    1990-01-01

    There are good prospects for silicon carbide anti-corrosion coatings on fuel elements to be realised, which opens up the chance to reduce the safety engineering requirements to the suitable design and safe performance of the ceramic fuel element. Another possibility offered is combined-cycle operation with high efficiencies, and thus good economic prospects, as with this design concept combining gas and steam turbines, air ingress due to turbine malfunction is an incident that can be managed by the system. This development will allow economically efficient operation also of nuclear power reactors with relatively small output, and hence contribute to reducing CO 2 emissions. (orig./DG) [de

  12. Probability of spent fuel transportation accidents

    International Nuclear Information System (INIS)

    McClure, J.D.

    1981-07-01

    The transported volume of spent fuel, incident/accident experience and accident environment probabilities were reviewed in order to provide an estimate of spent fuel accident probabilities. In particular, the accident review assessed the accident experience for large casks of the type that could transport spent (irradiated) nuclear fuel. This review determined that since 1971, the beginning of official US Department of Transportation record keeping for accidents/incidents, there has been one spent fuel transportation accident. This information, coupled with estimated annual shipping volumes for spent fuel, indicated an estimated annual probability of a spent fuel transport accident of 5 x 10 -7 spent fuel accidents per mile. This is consistent with ordinary truck accident rates. A comparison of accident environments and regulatory test environments suggests that the probability of truck accidents exceeding regulatory test for impact is approximately 10 -9 /mile

  13. Fission product release from defected nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    Lewis, B.J.

    1983-01-01

    The release of gaseous (krypton and xenon) and iodine radioactive fission products from defective fuel elements is described with a semi-empirical model. The model assumes precursor-corrected 'Booth diffusional release' in the UO 2 and subsequent holdup in the fuel-to-sheath gap. Transport in the gap is separately modelled with a phenomenological rate constant (assuming release from the gap is a first order rate process), and a diffusivity constant (assuming transport in the gap is dominated by a diffusional process). Measured release data from possessing various states of defection are use in this analysis. One element (irradiated in an earlier experiment by MacDonald) was defected with a small drilled hole. A second element was machined with 23 slits while a third element (fabricated with a porous end plug) displayed through-wall sheath hydriding. Comparison of measured release data with calculated values from the model yields estimates of empirical diffusion coefficients for the radioactive species in the UO 2 (1.56 x 10 -10 to 7.30 x 10 -9 s -1 ), as well as escape rate constants (7.85 x 10 -7 to 3.44 x 10 -5 s -1 ) and diffusion coefficients (3.39 x 10 -5 to 4.88 x 10 -2 cm 2 /s) for these in the fuel-to-sheath gap. Analyses also enable identification of the various rate-controlling processes operative in each element. For the noble gas and iodine species, the rate-determining process in the multi-slit element is 'Booth diffusion'; however, for the hydrided element an additional delay results from diffusional transport in the fuel-to-heath gap. Furthermore, the iodine species exhibit an additional holdup in the drilled element because of significant trapping on the fuel and/or sheath surfaces. Using experimental release data and applying the theoretical results of this work, a systematic procedure is proposed to characterize fuel failures in commercial power reactors (i.e., the number of fuel failures and average leak size)

  14. Store and process for intermediate or final storage of used fuel elements from a nuclear reactor

    International Nuclear Information System (INIS)

    Kumpf, H.

    1986-01-01

    The fuel elements are enclosed in boxes at the nuclear reactor and transported in these to the incoming station. Transport is a by truck, which makes it possible for the transport container to move in a vertical position, where the upper side is on the top side of the truck. The fuel elements in their boxes are handed over to a magazine there, which can be reached by a loading machine serving the storage room. (orig./HP) [de

  15. Spent fuels transportation coming from Australia

    International Nuclear Information System (INIS)

    2002-01-01

    Maritime transportation of spent fuels from Australia to France fits into the contract between COGEMA and ANSTO, signed in 1999. This document proposes nine information cards in this domain: HIFAR a key tool of the nuclear, scientific and technological australian program; a presentation of the ANSTO Australian Nuclear Science and Technology Organization; the HIFAR spent fuel management problem; the COGEMA expertise in favor of the research reactor spent fuel; the spent fuel reprocessing at La Hague; the transports management; the transport safety (2 cards); the regulatory framework of the transports. (A.L.B.)

  16. Fuel cell development for transportation: Catalyst development

    Energy Technology Data Exchange (ETDEWEB)

    Doddapaneni, N. [Sandia National Lab., Albuquerque, NM (United States)

    1996-04-01

    Fuel cells are being considered as alternate power sources for transportation and stationary applications. With proton exchange membrane (PEM) fuel cells the fuel crossover to cathodes causes severe thermal management and cell voltage drop due to oxidation of fuel at the platinized cathodes. The main goal of this project was to design, synthesize, and evaluate stable and inexpensive transition metal macrocyclic catalysts for the reduction of oxygen and be electrochemically inert towards anode fuels such as hydrogen and methanol.

  17. Storage container for radioactive fuel elements

    International Nuclear Information System (INIS)

    1984-01-01

    The interim storage cask for spent fuel elements or the glass moulds for high-level radioactive waste are made up of heat-resistant, reinforced concrete with chambers and highgrade steel lining. Cooling systems with natural air circulation are connected with the chambers. (HP) [de

  18. Prevention of criticality accidents. Fuel elements storage

    International Nuclear Information System (INIS)

    Canavese, S.I.; Capadona, N.M.

    1990-01-01

    Before the need to store fuel elements of the plate type MTR (Materials Testing Reactors), produced with enriched uranium at 20% in U235 for research reactors, it requires the design of a deposit for this purpose, which will give intrinsic security at a great extent and no complaints regarding its construction, is required. (Author) [es

  19. Nondestructive examination techniques on Candu fuel elements

    International Nuclear Information System (INIS)

    Gheorghe, G.; Man, I.

    2013-01-01

    During irradiation in nuclear reactor, fuel elements undergo dimensional and structural changes, and changes of surface conditions sheath as well, which can lead to damages and even loss of integrity. Visual examination and photography of Candu fuel elements are among the non-destructive examination techniques, next to dimensional measurements that include profiling (diameter, bending, camber) and length, sheath integrity control with eddy currents, measurement of the oxide layer thickness by eddy current techniques. Unirradiated Zircaloy-4 tubes were used for calibration purposes, whereas irradiated Zircaloy-4 tubes were actually subjected to visual inspection and dimensional measurements. We present results of measurements done by eddy current techniques on Zircaloy- 4 tubes, unirradiated, but oxidized in an autoclave prior to examinations. The purpose of these nondestructive examination techniques is to determine those parameters that characterize the behavior and performance of nuclear fuel operation. (authors)

  20. 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

  1. Research and Test Reactor Fuel Elements (RTRFE)

    International Nuclear Information System (INIS)

    Pace, Brett W.; Marinak, Edward A.

    1999-01-01

    BWX Technologies Inc. (BWXT) has experienced several production improvements over the past year. The homogeneity yields in 4.8 gU/cc U 3 Si 2 plates have increased over last year's already high yields. Through teamwork and innovative manufacturing techniques, maintaining high quality surface finishes on plates and elements is becoming easier and less expensive. Currently, BWXT is designing a fabrication development plan to reach a fuel loading of 9 gU/cc within 2 - 4 years. This development will involve a step approach requested by ANL to produce plates using U-8Mo at a loading of 6 gU/cc first and qualify the fuel at those levels. In achieving the goal of a very high-density fuel loading of 9 gU/cc, BWXT is considering employing several new, state of the art, ultrasonic testing techniques for fuel core evaluation. (author)

  2. Catalogue of fuel elements - 1. addendum October 1958

    International Nuclear Information System (INIS)

    Even, A.

    1957-01-01

    This document contains sheets presenting various characteristics of nuclear fuel elements which are distinguished with respect to their shape: cylinder bar, plate, tube. Each sheet comprises an indication of the atomic pile in which the fuel element is used, dimensions, cartridge data, data related to cooling, to combustion rate, and to fuel handling. A drawing of the fuel element is also given

  3. Certification test for safety of new fuel transportation package

    International Nuclear Information System (INIS)

    Aritomi, Masanori; Sugawa, Osami; Suga, Masao.

    1993-01-01

    The objective of this certification test is to prove the safety of new fuel transportation package against a fire of actual size caused by traffic accidents. After the fire test, the fuel assemblies were covered with coal-tar like material vaporized from anti-shock material used in the container. Surface color of BWR-type fuel assembly was dark grey that is supposed to be the color of oxide of Zircaloy. As for PWR-type fuel assembly, the condition encountered during fire test caused no change to the outlook of the rod element. Both the BWR and PWR type fuel rod elements showed no deformation and were completely sound. Therefore it may be concluded that the container protected the mimic fuel assemblies against fire of 30 minutes duration and caused no damage. This report is the result of the above experiments and examinations, and we appreciate the cooperation of those who are concerned. (J.P.N.)

  4. Spacer device for nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    Anthony, A.J.; Gaines, A.L.; Krawiec, D.M.

    1974-01-01

    The grid-type spacer device consists of two rows of main spacers arranged parallel to each other with some space in between, the first row extending perpendicular to the second row. Parallel to the respective rows of main spacers there are rows of secondary spacers interlocked with the main spacers. The individual spacers are welded together at their points of intersection. A large number of spring cages are installed within the spacer device to hold in place the main spacers which are oriented at right angles relative to each other. In addition, the spring cages serve for supporting the fuel elements. The spacers are made of zirconium which does not greatly influence the neutron capture cross section of the reactor. The material of the spring cages with the spring elements is a nickel alloy. It has the necessary stress relaxation properties to be able to force the fuel elements against the spacers under the action of the spring. (DG) [de

  5. Fuel removal, transport, and storage

    International Nuclear Information System (INIS)

    Reno, H.W.

    1986-01-01

    The March 1979 accident at Unit 2 of the Three Mile Island Nuclear Power Station (TMI-2) which damaged the core of the reactor resulted in numerous scientific and technical challenges. Some of those challenges involve removing the core debris from the reactor, packaging it into canisters, loading canisters into a rail cask, and transporting the debris to the Idaho National Engineering Laboratory (INEL) for storage, examination, and preparation for final disposal. This paper highlights how some challenges were resolved, including lessons learned and benefits derived therefrom. Key to some success at TMI was designing, testing, fabricating, and licensing two rail casks, which each provide double containment of the damaged fuel. 10 refs., 12 figs

  6. Nuclear reactor core and fuel element therefor

    International Nuclear Information System (INIS)

    Fortescue, P.

    1986-01-01

    This patent describes a nuclear reactor core. This core consists of vertical columns of disengageable fuel elements stacked one atop another. These columns are arranged in side-by-side relationship to form a substantially continuous horizontal array. Each of the fuel elements include a block of refractory material having relatively good thermal conductivity and neutron moderating characteristics. The block has a pair of parallel flat top and bottom end faces and sides which are substantially prependicular to the end faces. The sides of each block is aligned vertically within a vertical column, with the sides of vertically adjacent blocks. Each of the blocks contains fuel chambers, including outer rows containing only fuel chambers along the sides of the block have nuclear fuel material disposed in them. The blocks also contain vertical coolant holes which are located inside the fuel chambers in the outer rows and the fuel chambers which are not located in the outer rows with the fuel chambers and which extend axially completely through from end face to end face and form continuous vertical intracolumn coolant passageways in the reactor core. The blocks have vertical grooves extending along the sides of the blocks form interblock channels which align in groups to form continuous vertical intercolumn coolant passsageways in the reactor core. The blocks are in the form of a regular hexagonal prism with each side of the block having vertical gooves defining one half of one of the coolant interblock channels, six corner edges on the blocks have vertical groves defining one-third of an interblock channel, the vertical sides of the blocks defining planar vertical surfaces

  7. Laser assisted decontamination of nuclear fuel elements

    International Nuclear Information System (INIS)

    Padma Nilaya, J.; Biswas, Dhruba J.; Kumar, Aniruddha

    2010-04-01

    Laser assisted removal of loosely bound fuel particulates from the clad surface following the process of pellet loading has decided advantages over conventional methods. It is a dry and noncontact process that generates very little secondary waste and can occur inside a glove box without any manual interference minimizing the possibility of exposure to personnel. The rapid rise of the substrate/ particulate temperature owing to the absorption of energy from the incident laser pulse results in a variety of processes that may lead to the expulsion of the particulates. As a precursor to the cleaning of the fuel elements, initial experiments were carried out on contamination simulated on commonly used clad surfaces to gain a first hand experience on the various laser parameters for which as efficient cleaning can be obtained without altering the properties of the clad surface. The cleaning of a dummy fuel element was subsequently achieved in the laboratory by integrating the laser with a work station that imparted simultaneous rotational and linear motion to the fuel element. (author)

  8. Design of JMTR high-performance fuel element

    International Nuclear Information System (INIS)

    Sakurai, Fumio; Shimakawa, Satoshi; Komori, Yoshihiro; Tsuchihashi, Keiichiro; Kaminaga, Fumito

    1999-01-01

    For test and research reactors, the core conversion to low-enriched uranium fuel is required from the viewpoint of non-proliferation of nuclear weapon material. Improvements of core performance are also required in order to respond to recent advanced utilization needs. In order to meet both requirements, a high-performance fuel element of high uranium density with Cd wires as burnable absorbers was adopted for JMTR core conversion to low-enriched uranium fuel. From the result of examination of an adaptability of a few group constants generated by a conventional transport-theory calculation with an isotropic scattering approximation to a few group diffusion-theory core calculation for design of the JMTR high-performance fuel element, it was clear that the depletion of Cd wires was not able to be predicted accurately using group constants generated by the conventional method. Therefore, a new generation method of a few group constants in consideration of an incident neutron spectrum at Cd wire was developed. As the result, the most suitable high-performance fuel element for JMTR was designed successfully, and that allowed extension of operation duration without refueling to almost twice as long and offer of irradiation field with constant neutron flux. (author)

  9. Alternatives to traditional transportation fuels 1996

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-12-01

    Interest in alternative transportation fuels (ATF`s) has increased in recent years due to the drives for cleaner air and less dependence upon foreign oil. This report, Alternatives to Traditional Transportation Fuels 1996, provides information on ATFs, as well as the vehicles that consume them.

  10. Ecosystem element transport model for Lake Eckarfjaerden

    Energy Technology Data Exchange (ETDEWEB)

    Konovalenko, L.; Bradshaw, C. [The Department of Ecology, Environment and Plant Sciences, Stockholm University (Sweden); Andersson, E.; Kautsky, U. [Swedish Nuclear Fuel and Waste Management Co. - SKB (Sweden)

    2014-07-01

    The ecosystem transport model of elements was developed for Lake Eckarfjaerden located in the Forsmark area in Sweden. Forsmark has currently a low level repository (SFR) and a repository for spent fuel is planned. A large number of data collected during site-investigation program 2002-2009 for planning the repository were available for the creation of the compartment model based on carbon circulation, physical and biological processes (e.g. primary production, consumption, respiration). The model is site-specific in the sense that the food web model is adapted to the actual food web at the site, and most estimates of biomass and metabolic rates for the organisms and meteorological data originate from site data. The functional organism groups of Lake Eckarfjaerden were considered as separate compartments: bacterio-plankton, benthic bacteria, macro-algae, phytoplankton, zooplankton, fish, benthic fauna. Two functional groups of bacteria were taken into account for the reason that they have the highest biomass of all functional groups during the winter, comprising 36% of the total biomass. Effects of ecological parameters, such as bacteria and algae biomass, on redistribution of a hypothetical radionuclide release in the lake were examined. The ecosystem model was used to estimate the environmental transfer of several elements (U, Th, Ra) and their isotopes (U-238, U-234,Th-232, Ra-226) to various aquatic organisms in the lake, using element-specific distribution coefficients for suspended particle and sediment. Results of chemical analyses of the water, sediment and biota were used for model validation. The model gives estimates of concentration factors for fish based on modelling rather on in situ measurement, which reduces the uncertainties for many radionuclides with scarce of data. Document available in abstract form only. (authors)

  11. Automatic inspection for remotely manufactured fuel elements

    International Nuclear Information System (INIS)

    Reifman, J.; Vitela, J.E.; Gibbs, K.S.; Benedict, R.W.

    1995-01-01

    Two classification techniques, standard control charts and artificial neural networks, are studied as a means for automating the visual inspection of the welding of end plugs onto the top of remotely manufactured reprocessed nuclear fuel element jackets. Classificatory data are obtained through measurements performed on pre- and post-weld images captured with a remote camera and processed by an off-the-shelf vision system. The two classification methods are applied in the classification of 167 dummy stainless steel (HT9) fuel jackets yielding comparable results

  12. Failure analysis for WWER-fuel elements

    International Nuclear Information System (INIS)

    Boehmert, J.; Huettig, W.

    1986-10-01

    If the fuel defect rate proves significantly high, failure analysis has to be performed in order to trace down the defect causes, to implement corrective actions, and to take measures of failure prevention. Such analyses are work-consuming and very skill-demanding technical tasks, which require examination methods and devices excellently developed and a rich stock of experience in evaluation of features of damage. For that this work specifies the procedure of failure analyses in detail. Moreover prerequisites and experimental equipment for the investigation of WWER-type fuel elements are described. (author)

  13. Nuclear criticality assessment of LEU and HEU fuel element storage

    International Nuclear Information System (INIS)

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

    1984-01-01

    Criticality aspects of storing LEU (20%) and HEU (93%) fuel elements have been evaluated as a function of 235 U loading, element geometry, and fuel type. Silicide, oxide, and aluminide fuel types have been evaluated ranging in 235 U loading from 180 to 620 g per element and from 16 to 23 plates per element. Storage geometry considerations have been evaluated for fuel element separations ranging from closely packed formations to spacings of several centimeters between elements. Data are presented in a form in which interpolations may be made to estimate the eigenvalue of any fuel element storage configuration that is within the range of the data. (author)

  14. Solid fuel applications to transportation engines

    Energy Technology Data Exchange (ETDEWEB)

    Rentz, Richard L.; Renner, Roy A.

    1980-06-01

    The utilization of solid fuels as alternatives to liquid fuels for future transportation engines is reviewed. Alternative liquid fuels will not be addressed nor will petroleum/solid fuel blends except for the case of diesel engines. With respect to diesel engines, coal/oil mixtures will be addressed because of the high interest in this specific application as a result of the large number of diesel engines currently in transportation use. Final assessments refer to solid fuels only for diesel engines. The technical assessments of solid fuels utilization for transportation engines is summarized: solid fuel combustion in transportation engines is in a non-developed state; highway transportation is not amenable to solid fuels utilization due to severe environmental, packaging, control, and disposal problems; diesel and open-cycle gas turbines do not appear worthy of further development, although coal/oil mixtures for slow speed diesels may offer some promise as a transition technology; closed-cycle gas turbines show some promise for solid fuels utilization for limited applications as does the Stirling engine for use of cleaner solid fuels; Rankine cycle engines show good potential for limited applications, such as for locomotives and ships; and any development program will require large resources and sophisticated equipment in order to advance the state-of-the-art.

  15. Fuel element cluster for nuclear reactors

    International Nuclear Information System (INIS)

    Anthony, A.J.; Hutchinson, J.J.

    1976-01-01

    The claim refers to the constructional design of a fuel element cluster the elements of which are held by upper and lower end plates connected to each other in upright position, the bearing being formed by a screw connection between at least one guide tube for control rods and the two end plates. The claims are directed, especially, to the connection of the parts as well as to the materials selection which are determined to a high degree by the thermal expansion coefficients. (UA) [de

  16. Method of detecting a fuel element failure

    International Nuclear Information System (INIS)

    Cohen, P.

    1975-01-01

    A method is described for detecting a fuel element failure in a liquid-sodium-cooled fast breeder reactor consisting of equilibrating a sample of the coolant with a molten salt consisting of a mixture of barium iodide and strontium iodide (or other iodides) whereby a large fraction of any radioactive iodine present in the liquid sodium coolant exchanges with the iodine present in the salt; separating the molten salt and sodium; if necessary, equilibrating the molten salt with nonradioactive sodium and separating the molten salt and sodium; and monitoring the molten salt for the presence of iodine, the presence of iodine indicating that the cladding of a fuel element has failed. (U.S.)

  17. Nuclear fuel element nut retainer cup

    International Nuclear Information System (INIS)

    Walton, L.A.

    1977-01-01

    A typical embodiment has an end fitting for a nuclear reactor fuel element that is joined to the control rod guide tubes by means of a nut plate assembly. The nut plate assembly has an array of nuts, each engaging the respective threaded end of the control rod guide tubes. The nuts, moreover, are retained on the plate during handling and before fuel element assembly by means of hollow cylindrical locking cups that are brazed to the plate and loosely circumscribe the individual enclosed nuts. After the nuts are threaded onto the respective guide tube ends, the locking cups are partially deformed to prevent one or more of the nuts from working loose during reactor operation. The locking cups also prevent loose or broken end fitting parts from becoming entrained in the reactor coolant

  18. Tests on CANDU fuel elements sheath samples

    International Nuclear Information System (INIS)

    Ionescu, S.; Uta, O.; Mincu, M.; Prisecaru, I.

    2016-01-01

    This work is a study of the behavior of CANDU fuel elements after irradiation. The tests are made on ring samples taken from fuel cladding in INR Pitesti. This paper presents the results of examinations performed in the Post Irradiation Examination Laboratory. By metallographic and ceramographic examination we determinate that the hydride precipitates are orientated parallel to the cladding surface. A content of hydrogen of about 120 ppm was estimated. After the preliminary tests, ring samples were cut from the fuel rod, and were subject of tensile test on an INSTRON 5569 model machine in order to evaluate the changes of their mechanical properties as consequence of irradiation. Scanning electron microscopy was performed on a microscope model TESCAN MIRA II LMU CS with Schottky FE emitter and variable pressure. The analysis shows that the central zone has deeper dimples, whereas on the outer zone, the dimples are tilted and smaller. (authors)

  19. Fuel element radiometry system for quality control

    International Nuclear Information System (INIS)

    Bhattacharya, Sadhana; Gaur, Swati; Sridhar, Padmini; Mukhopadhyay, P.K.; Vaidya, P.R.; Das, Sanjoy; Sinha, A.K.; Bhatt, Sameer

    2010-01-01

    An indigenous and fully automatic PC based radiometry system has been designed and developed. The system required a vibration free scanning with various automated sequential movements to scan the fuel pin of size 5.8 mm (OD) x 1055 mm (L) along its full length. A mechanical system with these requirements and precision controls has been designed. The system consists of a tightly coupled and collimated radiation source-detector unit and data acquisition and control system. It supports PLC based control electronics to control and monitor the movement of fuel element, nuclear data acquisition and analysis system and feedback system to the mechanical scanner to physically accept or reject the fuel pin based on the decision derived by the software algorithms. (author)

  20. Fuel temperature characteristics of the 37-element and CANFLEX fuel bundle

    International Nuclear Information System (INIS)

    Bae, Jun Ho; Rho, Gyu Hong; Park, Joo Hwan

    2009-10-01

    This report describes the fuel temperature characteristics of CANFLEX fuel bundles and 37-element fuel bundles for a different burnup of fuel. The program was consisted for seeking the fuel temperature of fuel bundles of CANFLEX fuel bundles and 37-element fuel bundles by using the method in NUCIRC. Fuel temperature has an increasing pattern with the burnup of fuel for CANFLEX fuel bundles and 37-element fuel bundles. For all the case of burnup, the fuel temperature of CANFLEX fuel bundles has a lower value than that of 37-element fuel bundles. Especially, for the high power channel, the CANFLEX fuel bundles show a lower fuel temperature as much as about 75 degree, and the core averaged fuel temperature has a lower fuel temperature of about 50 degree than that of 37-element fuel bundles. The lower fuel temperature of CANFLEX fuel bundles is expected to enhance the safety by reducing the fuel temperature coefficient. Finally, for each burnup of CANFLEX fuel bundles and 37-element fuel bundles, the equation was present for predicting the fuel temperature of a bundle in terms of a coolant temperature and bundle power

  1. Alternatives to traditional transportation fuels 1993

    Energy Technology Data Exchange (ETDEWEB)

    1995-01-01

    In recent years, gasoline and diesel fuel have accounted for about 80 percent of total transportation fuel and nearly all of the fuel used in on-road vehicles. Growing concerns about the environmental effects of fossil fuel use and the Nation`s high level of dependence on foreign oil are providing impetus for the development of replacements or alternatives for these traditional transportation fuels. (The Energy Policy Act of 1992 definitions of {open_quotes}replacement{close_quotes} and {open_quotes}alternative{close_quotes} fuels are presented in the following box.) The Alternative Motor Fuels Act of 1988, the Clean Air Act Amendments of 1990 (CAAA90) and the Energy Policy Act of 1992 (EPACT) are significant legislative forces behind the growth of replacement fuel use. Alternatives to Traditional Transportation Fuels 1993 provides the number of on-road alternative fueled vehicles in use in the United States, alternative and replacement fuel consumption, and information on greenhouse gas emissions resulting from the production, delivery, and use of replacement fuels for 1992, 1993, and 1995.

  2. Development of the Fuel Element Database of PUSPATI TRIGA Reactor

    International Nuclear Information System (INIS)

    Nurhayati Ramli; Naim Syauqi Hamzah; Nurfazila Husain; Yahya Ismail; Mat Zin Mat Husin; Mohd Fairus Abd Farid

    2015-01-01

    Since June 28th, 1982, the PUSPATI TRIGA Reactor (RTP) operates safely with an accumulated energy release of about 17,200 MWhr, which corresponds to about 882 g of uranium burn-up. The reactor core has been reconfigured 15th times. Presently, there are 111 TRIGA fuel elements in the core, which 66 of the fuel elements are from the initial criticality while the rest of the fuel elements have been added to compensate the uranium consumption. As 59 % of the fuel elements are older than 30 years old, it is necessary to put the history of every fuel element in a database for easy access of the fuel element movement, inspection results history and integrity status. This paper intends to describe how the fuel element database is developed and related formulae used in determining the RTP fuel element elongation. (author)

  3. Design of fuel element for RA10

    International Nuclear Information System (INIS)

    Estevez, Esteban A.; Markiewicz, Mario; Gerding, Roberto

    2012-01-01

    The RA-10 reactor is an open pool multipurpose reactor. It is intended for radioisotopes production, fuel irradiation and use of neutron beam experiments. The nominal configuration core consists of 19 fuel elements (FE) and 6 in-core irradiation positions. With regard to the FE, although both conceptual design and manufacturing technology are similar to the already developed and qualified by CNEA (MTR fuel flat plate), the conditions imposed by the new reactor on FE's are more demanding that previous supplies. Here it should be mentioned the magnitude of the hydrodynamic forces acting on the FE caused by coolant flow through the core (upward) and mainly by the high coolant velocity between fuel plates (greater than 5 times than those currently in operation). Moreover, the high power density results in higher heat flux in fuel plates and greater temperature gradient. As a result of these increased demands present during irradiation, and in order to maintain a high level of reliability, it is necessary carry out some modifications in the mechanical design of the FE (with respect to the so-called ECBE design or s tandard ) . Design verification is performed through analytical and code calculations, and hydrodynamic tests on a full-scale prototype. This article describes the design of the FE for RA 10 reactor, with special emphasis on those aspects that represent innovations in the traditional design (ECBE). It also presents the functional requirements, design criteria and design limits established according to the reactor operational states (author)

  4. Analysis of near-term spent fuel transportation hardware requirements and transportation costs

    International Nuclear Information System (INIS)

    Daling, P.M.; Engel, R.L.

    1983-01-01

    A computer model was developed to quantify the transportation hardware requirements and transportation costs associated with shipping spent fuel in the commercial nucler fuel cycle in the near future. Results from this study indicate that alternative spent fuel shipping systems (consolidated or disassembled fuel elements and new casks designed for older fuel) will significantly reduce the transportation hardware requirements and costs for shipping spent fuel in the commercial nuclear fuel cycle, if there is no significant change in their operating/handling characteristics. It was also found that a more modest cost reduction results from increasing the fraction of spent fuel shipped by truck from 25% to 50%. Larger transportation cost reductions could be realized with further increases in the truck shipping fraction. Using the given set of assumptions, it was found that the existing spent fuel cask fleet size is generally adequate to perform the needed transportation services until a fuel reprocessing plant (FRP) begins to receive fuel (assumed in 1987). Once the FRP opens, up to 7 additional truck systems and 16 additional rail systems are required at the reference truck shipping fraction of 25%. For the 50% truck shipping fraction, 17 additional truck systems and 9 additional rail systems are required. If consolidated fuel only is shipped (25% by truck), 5 additional rail casks are required and the current truck cask fleet is more than adequate until at least 1995. Changes in assumptions could affect the results. Transportation costs for a federal interim storage program could total about $25M if the FRP begins receiving fuel in 1987 or about $95M if the FRP is delayed until 1989. This is due to an increased utilization of federal interim storage facility from 350 MTU for the reference scenario to about 750 MTU if reprocessing is delayed by two years

  5. Safe transport of irradiated fuel by sea

    International Nuclear Information System (INIS)

    Miller, M.L.

    1997-01-01

    The development is described of a transport system dedicated to the sea transport of irradiated nuclear fuel. The background is reviewed of why shipments were required and the establishment of a specialist shipping company, Pacific Nuclear Transport Limited. A description of the ships, flasks and other equipment utilised is provided, together with details of key procedures implemented to ensure safety and customer satisfaction. (Author)

  6. The sea transport of irradiated nuclear fuel

    International Nuclear Information System (INIS)

    Miller, M.L.

    1995-01-01

    The paper describes the development of a transport system dedicated to the sea transport of irradiated nuclear fuel. It reviews the background to why shipments were required and the establishment of a specialist shipping company, Pacific Nuclear Transport Limited. A description of the ships, flasks and other equipment utilized is provided, together with details of key procedures implemented to ensure safety and customer satisfaction

  7. Transport device for nuclear fuel powder

    International Nuclear Information System (INIS)

    Adelmann, M.

    1987-01-01

    The transport device for nuclear fuel powder, which does not disintegrate during transport, has a transport pipe which starts with its entry end from the floor or a closed container and opens with its outlet end at the top into a closed separation container connect via a powder filter to a suction pump. By alternate regular opening and closing of a first control valve for transport gas fitted to a transport pipe to a supply duct and a second control valve for transport gas fitted to the container to an additional supply duct, alternating plugs of nuclear fuel powder and transport gas cushions are formed and are transported to the outlet end of the transport pipe. (orig./HP) [de

  8. Fuels Performance | Transportation Research | NREL

    Science.gov (United States)

    chemistry, combustion, and emissions to answer two primary questions: How can improved fuel chemistry lead initiative Combustion and engines Fuel chemistry Emissions and fuel economy Find out about related NREL Chemistry. (2014) Saturated Monoglyceride Effects on Low-Temperature Performance of Biodiesel Blends. G.M

  9. Transport and supply logistics of biomass fuels: Vol. 1. Supply chain options for biomass fuels

    Energy Technology Data Exchange (ETDEWEB)

    Allen, J; Browne, M; Palmer, H; Hunter, A; Boyd, J

    1996-10-01

    The study which forms part of a wider project funded by the Department of Trade and Industry, looks at the feasibility of generating electricity from biomass-fuelled power stations. Emphasis is placed on supply availabilty and transport consideration for biomass fuels such as wood wastes from forestry, short rotation coppice products, straw, miscanthus (an energy crop) and farm animal slurries. The study details the elements of the supply chain for each fuel from harvesting to delivery at the power station. The delivered cost of each fuel, the environmental impact of the biomass fuel supply and other relevant non-technical issues are addressed. (UK)

  10. A procedure validation for high conversion reactors fuel elements calculation

    International Nuclear Information System (INIS)

    Ishida, V.N.; Patino, N.E.; Abbate, M.J.; Sbaffoni, M.M.

    1990-01-01

    The present work includes procedure validation of cross sections generation starting from nuclear data and the calculation system actually used at the Bariloche Atomic Center Reactor and Neutrons Division for its application to fuel elements calculation of a high conversion reactor (HCR). To this purpose, the fuel element calculation belonging to a High Conversion Boiling water Reactor (HCBWR) was chosen as reference problem, employing the Monte Carlo method. Various cases were considered: with and without control bars, cold of hot, at different vacuum fractions. Multiplication factors, reaction rates, power maps and peak factors were compared. A sensitivity analysis of typical cells used, the approximations employed to solve the transport equation (Sn or Diffusion), the 1-D or 2-D representation and densification of the spatial network used, with the aim of evaluating their influence on the parameters studied and to come to an optimum combination to be used in future design calculations. (Author) [es

  11. MOX fuel transport: the French experience

    International Nuclear Information System (INIS)

    Sanchis, H.; Verdier, A.; Sanchis, H.

    1999-01-01

    In the back-end of the fuel cycle, several leading countries have chosen the Reprocessing, Conditioning, Recycling (RCR) option. Plutonium recycling in the form of MOX fuel is a mature industry, with successful operational experience and large-scale fabrication plants an several European countries. The COGEMA Group has developed the industrialized products to master the RCR operation including transport COGEMA subsidiary, TRANSNUCLEAIRE has been operating MOX fuel transports on an industrial scale for more than 10 years. In 1998, around 200 transports of Plutonium materials have been organised by TRANSNUCLEAIRE. These transports have been carried out by road between various facilities in Europe: reprocessing plants, manufacturing plants and power plants. The materials transported are either: PuO 2 and MOX powder; BWR and PWR MOX fuel rods; BWR and PWR MOX fuel assemblies. Because MOX fuel transport is subject to specific safety, security and fuel integrity requirements, the MOX fuel transport system implemented by TRANSNUCLEAIRE is fully dedicated. Packaging have been developed, licensed and manufactured for each kind of MOX material in compliance with relevant regulations. A fleet of vehicles qualified according to existing physical protection regulations is operated by TRANSNUCLEAIRE. TRANSNUCLEAIRE has gained a broad experience in MOX transport in 10 years. Technical and operational know-how has been developed and improved for each step: vehicles and packaging design and qualification; vehicle and packaging maintenance; transport operations. Further developments are underway to increase the payload of the packaging and to improve the transport conditions, safety and security remaining of course top priority. (authors)

  12. Automatic welding of fuel elements; Soudure automatique des elements combustibles

    Energy Technology Data Exchange (ETDEWEB)

    Briola, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1958-07-01

    The welding process depends on the type of fuel element, the can material and the number of cartridges to be welded: - inert-gas welding (used for G2 and the 1. set of EL3), - inert atmosphere arc welding (used for welding uranium and zirconium), - electronic welding (used for the 2. set of EL3 and the tank of Proserpine). (author) [French] Suivant le type d'element combustible, le materiau de gaine et l'importance de la serie a fabriquer, le soudeur dispose des differents procedes examines dans cette communication: - soudure classique a l'arc sous gaz inerte (utilisee pour G2 et le premier jeu EL3), - soudure en atmosphere complete d'argon (utilisee pour la soudure d'uranium et de zirconium), - soudure electronique (utilisee pourdeuxieme jeu EL3 et la cuve de Proserpine). (auteur)

  13. Transportation 2000. Spent fuel transportation trends in the new millenium

    International Nuclear Information System (INIS)

    Blee, David; Viebrock, James; Patterson, John

    1999-01-01

    The paper will provide a comparison of foreign research reactor spent fuel transportation today verses the assumptions used by the Department of Energy in the Environmental Impact Statement. In addition, it will suggest changes that are likely to occur in transportation logistics through the remainder of the U.S. spent fuel returns program. Cask availability, certification status, shipment strategy, cost issues, and public acceptance are among the topical areas that will be examined. Transportation requirements will be assessed in light of current participation in the returns program and the tendency for shipment plans to shift toward spent fuel return toward the end of the 13 year period of eligibility. (author)

  14. European experience with spent fuel transport

    International Nuclear Information System (INIS)

    Hunter, I.A.

    1995-01-01

    Nuclear Transport Ltd has transported 5000 tonnes of spent fuel from 35 reactors in 8 European countries since 1972. Transport management is governed by the Quality Plan for: transport administration, packaging and shipment procedures at the shipping plant, operations at the power plant, and packaging and shipment organization at the power plant. Selection of a suitable carrier device is made with regard to the shipping plant requirements, physical limitations of the reactor, fuel characteristics, and transport route constraints. The transport plan is set up taking into account exploitation of the casks, reactor shut-down requirements, fuel acceptance plans at the reprocessing plant, and cask maintenance periods. A transport cycle involving spent fuel shipment to La Hague or to Sellafield takes typically two or four weeks, respectively. Most transports through Europe are by rail. A special-design railway ferry boat serves transports to the United Kingdom. Both wet or dry casks are employed. Modern casks are designed for high burnups and for oxide fuels. (J.B.)

  15. Estimating road transport fuel consumption in Ecuador

    International Nuclear Information System (INIS)

    Sierra, Jaime Cevallos

    2016-01-01

    Road transport is one of the sectors with highest energy consumptions in the planet, with large dependence of fossil fuels, and contribution for global greenhouse gas emissions. Although, Latin America is not a high-energy consumer, its share in global consumption is expected to grow, especially in the transportation sector. This make essential for developing countries the adoption of better policies to identify the vehicle groups with largest fuel demands. The present study describes the VKT technique to disaggregate road transport energy consumption by vehicle type, applied to the road transportation system of Ecuador. It also describes the procedures performed to estimate the variables required to run the model, and some of the practical applications that be used to create public policies. Results show as the biggest fuel consumers the heavy-duty freight cargo, followed by light duty vehicles. The estimation of greenhouse gas emissions evidence that road transport released 14.3 million tons of CO_2 in 2012. When fuel consumption is compared by it costs, it can be confirmed that Ecuadorean Government covered, through subsidies, for 68% of the annual fuel costs of national road transport, demonstrating the importance of restructuring these expenditures in order to achieve an efficient road transport system. - Highlights: •The vehicle-kilometers traveled has been estimated from local info. •The fuel economy has been calculated from national and international data. •The groups with higher fuel consumption has been located. •The fuel-type dependency has been estimated for each vehicle group. •Greenhouse gas emission, and fuel costs, has been estimated for local road transport.

  16. Nuclear reactor fuel element sub-assemblies

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  17. Searching for a possible fuel element leak

    International Nuclear Information System (INIS)

    Dodd, B.; Johnson, A.G.

    1986-01-01

    A gamma spectrum analysis of a filter paper from an Oregon State University TRIGA Reactor (OSTR) continuous air monitor (CAM) which routinely monitors the air directly over the reactor tank revealed just-detectable levels of several short-lived particulate fission products typically associated with a fuel cladding failure. This prompted an intensive.search to determine the origin of these radionuclides. A number of methods were used, including a fuel element rotation program designed to ultimately remove all of the fuel elements from the core in groups of three, and a scheme to selectively sample bubbles from different parts of the core during operation. Determination of the source was made very difficult by the fact that its presence was erratic in nature and because radioactivity levels found on filter papers were on the border of detectability even when the reactor was operated at the maximum allowable power level of 1MW. The origin and source of the fission product activity was not found, no other abnormality was identified and the reactor was therefore returned to normal operation. In addition to continuing the routine operation of the reactor-top CAM, further surveillance designed to detect a positive reappearance of the source was also implemented and currently involves a complete gamma spectrum analysis of a CAM filter paper each week after a standard (controlled) 3 hour reactor run at 1 MW. (author)

  18. Fuel containing vessel for transporting nuclear fuel

    International Nuclear Information System (INIS)

    Yoshizawa, Hiroyasu; Shimizu, Fukuzo; Tanaka, Nobuyuki.

    1996-01-01

    A shock absorbing mechanism is disposed on an inner bottom of a vessel main body. The shock absorbing mechanism comprises a shock absorbing member disposed on the upper surface of a bottom wall, an annular metal plate disposed on the upper surface of the shock absorbing member and an annular spacer disposed on the upper surface of the metal plate. The shock absorbing member is made of a material such as of wood, lead, metal honeycomb or a metal mesh, which plastically deforms when applied with load higher than a predetermined level, and is formed in a square block-like form covering the upper surface of the bottom wall. The spacer is made of a thin soft material such as tetrafluoroethylene, and is formed in such a shape as capable of preventing direct contact of the lower end of the cylindrical member in a lower tie plate of nuclear fuels with the metal portion. This can ensure integrity of nuclear fuels even when they fall from a high place upon an assumed dropping accident. (I.N.)

  19. Modeling of PHWR fuel elements using FUDA code

    International Nuclear Information System (INIS)

    Tripathi, Rahul Mani; Soni, Rakesh; Prasad, P.N.; Pandarinathan, P.R.

    2008-01-01

    The computer code FUDA (Fuel Design Analysis) is used for modeling PHWR fuel bundle operation history and carry out fuel element thermo-mechanical analysis. The radial temperature profile across fuel and sheath, fission gas release, internal gas pressure, sheath stress and strains during the life of fuel bundle are estimated

  20. Subcritical Noise Analysis Measurements with Fresh and Spent Research Reactor Fuels Elements

    International Nuclear Information System (INIS)

    Valentine, T.E.; Mihalczo, J.T.; Kryter, R.C.; Miller, V.C.

    1999-01-01

    The verification of the subcriticality is of utmost importance for the safe transportation and storage of nuclear reactor fuels. Transportation containers and storage facilities are designed such that nuclear fuels remain in a subcritical state. Such designs often involve excess conservatism because of the lack of relevant experimental data to verify the accuracy of Monte Carlo codes used in nuclear criticality safety analyses. A joint experimental research program between Oak Ridge National Laboratory, Westinghouse Safety Management Solutions, Inc., and the University of Missouri was initiated to obtain measured quantities that could be directly related to the subcriticality of simple arrays of Missouri University Research Reactor (MURR) fuel elements. A series of measurement were performed to assess the reactivity of materials such as BORAL, stainless steel, aluminum, and lead that are typically used in the construction of shipping casks. These materials were positioned between the fuel elements. In addition, a limited number of measurements were performed with configurations of fresh and spent (irradiated) fuel elements to ascertain the reactivity of the spent fuel elements. In these experiments, fresh fuel elements were replaced by spent fuel elements such that the subcritical reactivity change could be measured. The results of these measurements were used by Westinghouse Safety Management Solutions to determine the subcriticality of MURR fuel elements isolated by absorbing materials. The measurements were interpreted using the MCNP-DSP Monte Carlo code to obtain the subcritical neutron multiplication factor k(sub eff), and the bias in K(sub eff) that are used in criticality safety analyses

  1. Neutron induced activity in fuel element components

    International Nuclear Information System (INIS)

    Kjellbert, N.

    1978-03-01

    A thorough investigation of the importance of various nuclides in neutron-induced radioactivity from fuel element construction materials has been carried out for both BWR and PWR fuel assemblies. The calculations were performed with the ORIGEN computer code. The investigation was directed towards the final storage of the assembly components and special emphasis was put to the examination of the sources of carbon-14, cobalt-60, nickel-59, nickel-63 and zirconium-93/niobium-93m. It is demonstrated that the nuclides nickel-59, in Inconel and stainless steel, and zirconium-93/niobium-93m, in Zircaloy, are the ones which constitute the very long term radiotoxic hazard of the irradiated materials. (author)

  2. Test of high temperature fuel element, (1)

    International Nuclear Information System (INIS)

    Akino, Norio; Shiina, Yasuaki; Nekoya, Shin-ichi; Takizuka, Takakazu; Emori, Koichi

    1980-11-01

    Heat transfer experiment to measure the characteristics of a VHTR fuel in the same condition of the reactor core was carried out using HTGL (High Temperature Helium Gas Loop) and its test section. In this report, the details of the test section, related problems of construction and some typical results are described. The newly developed heater with graphite heat transfer surface was used as a simulated fuel element to determine the heat transfer characteristics. Following conclusions were obtained; (1) Reynolds number between turbulent and transitional region is about 2600. (2) Reynolds number between transitional and laminar region is about 4800. (3) The laminarization phenomena have not been observed and are hardly occurred in annular tubes comparing with round tube. (4) Measured Nusselt numbers agree to the established correlations in turbulent and laminar regions. (author)

  3. Alternative transportation fuels: Financing issues

    International Nuclear Information System (INIS)

    Squadron, W.F.; Ward, C.O.; Brown, M.H.

    1992-06-01

    A multitude of alternative fuels could reduce air pollution and the impact of oil price shocks. Only a few of these fuels are readily available and inexpensive enough to merit serious consideration over the coming five years. In New York City, safety regulations narrow the field still further by eliminating propane. As a result, this study focuses on the three alternative fuels readily available in New York City: compressed natural gas, methanol, and electricity. Each has significant environmental benefits and each has different cost characteristics. With the Clean Air Act and the National Energy Strategy highlighting the country's need to improve urban air quality and move away from dependence on imported fuels, fleets may soon have little choice but to convert to altemative fuels. Given the potential for large infrastructure and vehicle costs, these fleets may have difficulty finding the capital to make that conversion. Ultimately, then, it will be the involvement of the private sector that will determine the success of alternative fuels. Whether it be utilities, fuel distributors or suppliers, private financing partners or others, it is critical that altemative fuels programs be structured and planned to attract their involvement. This report examines financing methods that do not involve government subsidies. It also explores financing methods that are specific to alternative fuels. Bond issues and other mechanisms that are used for conventional vehicles are not touched upon in this report. This report explores ways to spread the high cost of alternative fuels among a number of parties within the private sector. The emphasis is on structuring partnerships that suit methanol, electric, or natural gas vehicle fleets. Through these partnerships, alternative fuels may ultimately compete effectively against conventional vehicle fuels

  4. Influences of in-fuel physical-chemical processes on serviceability of energy reactor fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Bibilashvili, Yu K; Nekrasova, G A; Sukhanov, G I

    1989-01-01

    In-fuel physico-chemical processes and their effect on stress corrosion cracking of fuel element zirconium cladding are considered in the review. The mechanism of fission product release from the fuel is studied and the negative role of primarily iodine on the cladding corrosion process is demonstrated. Directions for improving the fuel element claddings and fuel to increase the fuel element serviceability are specified.

  5. Influences of in-fuel physical-chemical processes on serviceability of energy reactor fuel elements

    International Nuclear Information System (INIS)

    Bibilashvili, Yu.K.; Nekrasova, G.A.; Sukhanov, G.I.

    1989-01-01

    In-fuel physico-chemical processes and their effect on stress corrosion cracking of fuel element zirconium cladding are considered in the review. The mechanism of fission product release from the fuel is studied and the negative role of primarily iodine on the cladding corrosion process is demonstrated. Directions for improving the fuel element claddings and fuel to increase the fuel element serviceability are specified

  6. Transporting spent nuclear fuel: an overview

    International Nuclear Information System (INIS)

    1986-03-01

    Although high-level radioactive waste from both commercial and defense activities will be shipped to the repository, this booklet focuses on various aspects of transporting commercial spent fuel, which accounts for the majority of the material to be shipped. The booklet is intended to give the reader a basic understanding of the following: the reasons for transportation of spent nuclear fuel, the methods by which it is shipped, the safety and security precautions taken for its transportation, emergency response procedures in the event of an accident, and the DOE program to develop a system uniquely appropriate to NWPA transportation requirements

  7. Effects of fueling profiles on plasma transport

    International Nuclear Information System (INIS)

    Houlberg, W.A.; Mense, A.T.; Attenberger, S.E.; Milora, S.L.

    1977-01-01

    The effects of cold particle fueling profiles on particle and energy transport in an ignition sized tokamak plasma are investigated in this study with a one-dimensional, multifluid transport model. A density gradient driven trapped particle microinstability model for plasma transport is used to demonstrate potential effects of fueling profiles on ignition requirements. Important criteria for the development of improved transport models under the conditions of shallow particle fueling profiles are outlined. A discrete pellet fueling model indicates that large fluctuations in density and temperature may occur in the outer regions of the plasma with large, shallowly penetrating pellets, but fluctuations in the pressure profile are small. The hot central core of the plasma remains unaffected by the large fluctuations near the plasma edge

  8. Alternatives to traditional transportation fuels 1995

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-12-01

    This report provides information on transportation fuels other than gasoline and diesel, and the vehicles that use these fuels. The Energy Information Administration (EIA) provides this information to support the U.S. Department of Energy`s reporting obligations under Section 503 of the Energy Policy Act of 1992 (EPACT). The principal information contained in this report includes historical and year-ahead estimates of the following: (1) the number and type of alterative-fueled vehicles (AFV`s) in use; (2) the consumption of alternative transportation fuels and {open_quotes}replacement fuels{close_quotes}; and (3) the number and type of alterative-fueled vehicles made available in the current and following years. In addition, the report contains some material on special topics. The appendices include a discussion of the methodology used to develop the estimates (Appendix A), a map defining geographic regions used, and a list of AFV suppliers.

  9. Discrete elements method of neutron transport

    International Nuclear Information System (INIS)

    Mathews, K.A.

    1988-01-01

    In this paper a new neutron transport method, called discrete elements (L N ) is derived and compared to discrete ordinates methods, theoretically and by numerical experimentation. The discrete elements method is based on discretizing the Boltzmann equation over a set of elements of angle. The discrete elements method is shown to be more cost-effective than discrete ordinates, in terms of accuracy versus execution time and storage, for the cases tested. In a two-dimensional test case, a vacuum duct in a shield, the L N method is more consistently convergent toward a Monte Carlo benchmark solution

  10. Environmental economics of lignin derived transport fuels

    OpenAIRE

    Obydenkova, SV; Kouris, P Panagiotis; Hensen, EJM Emiel; Heeres, Hero J; Boot, MD Michael

    2017-01-01

    This paper explores the environmental and economic aspects of fast pyrolytic conversion of lignin, obtained from 2G ethanol plants, to transport fuels for both the marine and automotive markets. Various scenarios are explored, pertaining to aggregation of lignin from several sites, alternative energy carries to replace lignin, transport modalities, and allocation methodology. The results highlight two critical factors that ultimately determine the economic and/or environmental fuel viability....

  11. Gas transport in solid oxide fuel cells

    CERN Document Server

    He, Weidong; Dickerson, James

    2014-01-01

    This book provides a comprehensive overview of contemporary research and emerging measurement technologies associated with gas transport in solid oxide fuel cells. Within these pages, an introduction to the concept of gas diffusion in solid oxide fuel cells is presented. This book also discusses the history and underlying fundamental mechanisms of gas diffusion in solid oxide fuel cells, general theoretical mathematical models for gas diffusion, and traditional and advanced techniques for gas diffusivity measurement.

  12. Phenomena in thermal transport in fuels

    International Nuclear Information System (INIS)

    Chernatynskiy, A.; Tulenko, J.S.; Phillpot, S.R.; El-Azab, A.

    2015-01-01

    Thermal transport in nuclear fuels is a key performance metric that affects not only the power output, but is also an important consideration in potential accident situations. While the fundamental theory of the thermal transport in crystalline solids was extensively developed in the 1950's and 1960's, the pertinent analytic approaches contained significant simplifications of the physical processes. While these approaches enabled estimates of the thermal conductivity in bulk materials with microstructure, they were not comprehensive enough to provide the detailed guidance needed for the in-pile fuel performance. Rather, this guidance has come from data painfully accumulated over 50 years of experiments on irradiated uranium dioxide, the most widely used nuclear fuel. At this point, a fundamental theoretical understanding of the interplay between the microstructure and thermal conductivity of irradiated uranium dioxide fuel is still lacking. In this chapter, recent advances are summarised in the modelling approaches for thermal transport of uranium dioxide fuel. Being computational in nature, these modelling approaches can, at least in principle, describe in detail virtually all mechanisms affecting thermal transport at the atomistic level, while permitting the coupling of the atomistic-level simulations to the mesoscale continuum theory and thus enable the capture of the impact of microstructural evolution in fuel on thermal transport. While the subject of current studies is uranium dioxide, potential applications of the methods described in this chapter extend to the thermal performance of other fuel forms. (authors)

  13. Examination on the safety of handling the fuel elements in the nuclear ship 'Mutsu'

    International Nuclear Information System (INIS)

    1977-01-01

    This is the report of the Examination Committee on Total Inspection and Repair Technologies for Mutsu to the Director of Science and Technology Agency and the Minister of Transport dated July 29, 1977. The committee concluded before that the total inspection on safety and the repair of shielding can be carried out as the fuel elements are loaded, and the safety can be secured sufficiently. It was decided at the meeting of ministers concerned with Mutsu on May 17 that the safety concerning handling the fuel elements of Mutsu should be examined by the committee. Under the premise that the fuel elements are loaded again and used after the total inspection on safety and the repair of shielding, the committee examined the methods and the basic concept of safety about the taking-out, transport and preservation of the fuel elements, and the conclusions obtained are reported. The contents of the examination are the outline of the fuel elements, the present condition of the fuel elements, the safety concerning taking-out, transport and preservation of the fuel elements, and the other measures required for securing safety. The committee thinks that the safety can be secured sufficiently if the works are carried out carefully. (Kako, I.)

  14. Features of spherical uranium-graphite HTGR fuel elements control

    International Nuclear Information System (INIS)

    Kreindlin, I.I.; Oleynikov, P.P.; Shtan, A.S.

    1985-01-01

    Control features of spherical HTGR uranium-graphite fuel elements with spherical coated fuel particles are mainly determined by their specific construction and fabrication technology. The technology is chiefly based on methods of ceramic fuel (fuel microspheres fabrication) and graphite production practice it is necessary to deal with a lot of problems from determination of raw materials properties to final fuel elements testing. These procedures are described

  15. Features of spherical uranium-graphite HTGR fuel elements control

    Energy Technology Data Exchange (ETDEWEB)

    Kreindlin, I I; Oleynikov, P P; Shtan, A S

    1985-07-01

    Control features of spherical HTGR uranium-graphite fuel elements with spherical coated fuel particles are mainly determined by their specific construction and fabrication technology. The technology is chiefly based on methods of ceramic fuel (fuel microspheres fabrication) and graphite production practice it is necessary to deal with a lot of problems from determination of raw materials properties to final fuel elements testing. These procedures are described.

  16. Development of experimental methods for measuring fuel elements burnup

    International Nuclear Information System (INIS)

    PEREDA, C; HENRIQUEZ, C; NAVARRO, G; TORRES, H; KLEIN, J; CALDERON, D; MEDEL, J; MUTIS, O; DAIE, J; ITURRIETA, L; LONCOMILLA, M; ZAMBRANO, J; KESTELMAN, A

    2003-01-01

    This paper is a summary of the work carried out during the last two years in fuel burning measurements at RECH-1 for different enrichments, cooling times and burning rates. The measurements were made in two gamma-spectrometric facilities, one is installed in a hot cell and the other inside of the secondary pool of the RECH-1, where the element is under 2 meters of water. The hot cell measurements need at least 100 cooling days because of the problems generated by the transport of highly active fuel elements from the Reactor to the cell. This was the main reason for using the in-pool facility because of its capability to measure the burning of fuel elements without having to wait so long, that is with only 5 cooling days. The accumulated experience in measurements achieved in both facilities and the encouraging results show that this measuring method is reliable. The results agreed well with those obtained using the reactor's physics codes, which was the way they were obtained previously (Cw)

  17. DOE perspective on fuel cells in transportation

    Energy Technology Data Exchange (ETDEWEB)

    Kost, R.

    1996-04-01

    Fuel cells are one of the most promising technologies for meeting the rapidly growing demand for transportation services while minimizing adverse energy and environmental impacts. This paper reviews the benefits of introducing fuel cells into the transportation sector; in addition to dramatically reduced vehicle emissions, fuel cells offer the flexibility than use petroleum-based or alternative fuels, have significantly greater energy efficiency than internal combustion engines, and greatly reduce noise levels during operation. The rationale leading to the emphasis on proton-exchange-membrane fuel cells for transportation applications is reviewed as are the development issues requiring resolution to achieve adequate performance, packaging, and cost for use in automobiles. Technical targets for power density, specific power, platinum loading on the electrodes, cost, and other factors that become increasingly more demanding over time have been established. Fuel choice issues and pathways to reduced costs and to a renewable energy future are explored. One such path initially introduces fuel cell vehicles using reformed gasoline while-on-board hydrogen storage technology is developed to the point of allowing adequate range (350 miles) and refueling convenience. This scenario also allows time for renewable hydrogen production technologies and the required supply infrastructure to develop. Finally, the DOE Fuel Cells in Transportation program is described. The program, whose goal is to establish the technology for fuel cell vehicles as rapidly as possible, is being implemented by means of the United States Fuel Cell Alliance, a Government-industry alliance that includes Detroit`s Big Three automakers, fuel cell and other component suppliers, the national laboratories, and universities.

  18. Storage system and method for spent fuel elements

    International Nuclear Information System (INIS)

    Queiser, H.; Eckardt, B.

    1981-01-01

    The proposal concerns an additional protection against leakage of a FE-transport container for interim storage of spent fuel elements. The gastight container has a second cover placed at a short distance from the first cover. The intermediate hollow space can be connected with a measuring system which indicates if part of the trace gas (mostly helium) added as indicator has escaped from the container due to leakage. The description explains the method and the assembly of required lines and measuring points etc. (UWI) [de

  19. Method of manufacturing nuclear fuel elements

    International Nuclear Information System (INIS)

    Ishida, Masao; Oguma, Masaomi.

    1980-01-01

    Purpose: To effectively prevent the bending of nuclear fuel elements in the reactor by grinding the end faces of pellets due to their mutual sliding. Method: In the manufacturing process of nuclear fuel elements, a plurality of pellets whose sides have been polished are fed one by one by way of a feeding mechanism through the central aperture in an electric motor into movable arms and retained horizontally with the central axis by being held on the side. Then, the pellet held by one of the arms is urged to another pellet held by the other of the arms by way of a pressing mechanism and the mating end faces of both of the pellets are polished by mutual sliding. Thereafter, the grinding dusts resulted are eliminated by drawing pressurized air and then the pellets are enforced into a cladding tube. Thus, the pellets are charged into the cladding tube with both polished end faces being contacted to each other, whereby the axial force is uniformly transmitted within the end faces to prevent the bending of the cladding tube. (Kawakami, Y.)

  20. Transport insurance of unirradiated nuclear fuels

    International Nuclear Information System (INIS)

    Matto, H.

    1985-01-01

    Special conditions must be taken into account in transport insurance for nuclear materials even if the nuclear risk involved is negligible, as in shipments of unirradiated nuclear fuels. The shipwreck of the 'Mont Louis' has raised a number of open points which must be solved pragmatically within the framework of transport insurance. Some proposals are outlined in the article. (orig.) [de

  1. The calculation - experimental investigations of the HTGR fuel element construction

    International Nuclear Information System (INIS)

    Eremeev, V.S.; Kolesov, V.S.; Chernikov, A.S.

    1985-01-01

    One of the most important problems in the HTGR development is the creation of the fuel element gas-tight for the fission products. This problem is being solved by using fuel elements of dispersion type representing an ensemble of coated fuel particles dispersed in the graphite matrix. Gas-tightness of such fuel elements is reached at the expense of deposing a protective coating on the fuel particles. It is composed of some layers serving as diffusion barriers for fission products. It is apparent that the rate of fission products diffusion from coated fuel particles is determined by the strength and temperature of the protective coating

  2. Pre-irradiation testing of experimental fuel elements

    International Nuclear Information System (INIS)

    Basova, B.G.; Davydov, E.F.; Dvoretskij, V.G.; Ivanov, V.B.; Syuzev, V.N.; Timofeev, G.A.; Tsykanov, V.A.

    1979-01-01

    The problems of testing of experimental fuel elements of nuclear reactors on the basis of complex accountancy of the factors defining operating capacity of the fuel elements are considered. The classification of the parameters under control and the methods of initial technological testing, including testing of the fuel product, cladding and fished fuel element, is given. The requirements to the apparatus used for complex testing are formulated. One of the possible variants of representation of the information obtained in the form of the input certificate of a single fuel element under study is proposed. The processing flowsheet of the gathered information using the computer is given. The approach under consideration is a methodological basis of investigation of fuel element operating life at the testing stage of the experimental fuel elements

  3. TRIGA fuel element burnup determination by measurement and calculation

    International Nuclear Information System (INIS)

    Zagar, T.; Ravnik, M.; Persic, A.; Jeraj, R.

    2000-01-01

    To estimate the accuracy of the fuel element burnup calculation different factors influencing the calculation were studied. To cover different aspects of burnup calculations, two in-house developed computer codes were used in calculations. The first (TRIGAP) is based on a one-dimensional two-group diffusion approximation, and the second (TRIGLAV) is based on a two-dimensional four-group diffusion equation. Both codes use WIMSD program with different libraries forunit-cell cross section data calculation. The burnup accumulated during the operating history of the TRIGA reactor at Josef Stefan Institute was calculated for all fuel elements. Elements used in the core during this period were standard SS 8.5% fuel elements, standard SS 12% fuel elements and highly enriched FLIP fuel elements. During the considerable period of operational history, FLIP and standard fuel elements were used simultaneously in mixed cores. (authors)

  4. French experience in research reactor fuel transportation

    International Nuclear Information System (INIS)

    Raisonnier, Daniele

    1996-01-01

    Since 1963 Transnucleaire has safely performed a large number of national and international transports of radioactive material. Transnucleaire has also designed and supplied suitable packaging for all types of nuclear fuel cycle radioactive material from front-end and back-end products and for power or for research reactors. Transportation of spent fuel from power reactors are made on a regular and industrial basis, but this is not yet the case for the transport of spent fuel coming from research reactors. Each shipment is a permanent challenge and requires a reactive organization dealing with all the transportation issues. This presentation will explain the choices made by Transnucleaire and its associates to provide and optimize the corresponding services while remaining in full compliance with the applicable regulations and customer requirements. (author)

  5. Fabrication of the Spent Fuel Elements Rack on the ISFSF

    International Nuclear Information System (INIS)

    Slamet Wiranto; Sigit Purwanto; Safrul, H.

    2004-01-01

    The Interim Storage For Spent Fuel elements (ISFSF) was designed to be able to store the 33 spent fuel element racks with capacity of 1386 of normal spent fuel elements and 2 racks for 36 of defected ones. Until now, only 9 out of 33 racks of normal spent fuel elements and lout of 2 racks of defected fuel elements are available. Five of them have suffered from corrosion so that they are not fulfilled the requirements of the spent fuel elements storage anymore. Meanwhile, the spent fuel storage racks in the reactor are almost full. It means, the transfer of the spent fuel from reactor spent fuel storage to the ISFSF pool are compulsory needed. Therefore, it is necessary to provide the new ISFSF spent fuel storage rack with better material and fabrication method than the old one. In this design all materials consist of SS 316 L that are welded with the Argon TIG-welding. Right now there has been one new spent fuel storage rack fabricated with capacity of 42 normal spent fuel elements. (author)

  6. Fuel element concept for long life high power nuclear reactors

    Science.gov (United States)

    Mcdonald, G. E.; Rom, F. E.

    1969-01-01

    Nuclear reactor fuel elements have burnups that are an order of magnitude higher than can currently be achieved by conventional design practice. Elements have greater time integrated power producing capacity per unit volume. Element design concept capitalizes on known design principles and observed behavior of nuclear fuel.

  7. Modeling the highway transportation of spent fuel

    International Nuclear Information System (INIS)

    Harrison, I.G.

    1986-01-01

    There will be a substantial increase in the number of spent fuel shipments on the nation's highway system in the next thirty years. Most of the spent fuel will be moving from reactors to a spent fuel repository. This study develops two models that evaluate the risk and cost of moving the spent fuel. The Minimum Total Transport Risk Model (MTTRM) seeks an efficient solution for this problem by finding the minimum risk path through the network and sending all the spent fuel shipments over this one path. The Equilibrium Transport Risk Model (ETRM) finds an equitable solution by distributing the shipments over a number of paths in the network. This model decreases the risk along individual paths, but increases society's risk because the spent fuel shipments are traveling over more links in the network. The study finds that there is a trade off between path risk and societal risk. As path risk declines, societal risk rises. The cost of shipping also increases as the number of paths expand. The cost and risk of shipping spent fuel from ten reactors to four potential repository sites are evaluated using the MTTRM. The temporary monitored retrievable storage (MRS) facility in Tennessee is found to be the minimum cost and minimum risk solution. When direct shipment to the permanent sites is considered, Deaf Smith, Texas is the least cost and least incident free transport risk location. Yucca Mountain, Nevada is the least risk location when the focus is placed on the potential consequences of an accident

  8. Impact of Transmutation Scenarios on Fuel Transportation

    International Nuclear Information System (INIS)

    Saturnin, A.; Duret, B.; Allou, A.; Jasserand, F.; Fillastre, E.; Giffard, F.X.; Chabert, C.; Caron-Charles, M.; Garzenne, C.; Laugier, F.

    2015-01-01

    Minor actinides transmutation scenarios have been studied in the frame of the French Sustainable Radioactive Waste Management Act of 28 June 2006. Transmutation scenarios supposed the introduction of a sodium-cooled fast reactor fleet using homogeneous or heterogeneous recycling modes for the minor actinides. Americium, neptunium and curium (MA) or americium alone (Am) can be transmuted together in a homogeneous way embedded in FR-MOX fuel or incorporated in MA or Am-Bearing radial Blankets (MABB or AmBB). MA transmutation in Accelerator Driven System has also been studied while plutonium is being recycled in SFR. Assessments and comparisons of these advanced cycles have been performed considering technical and economic criteria. Transportation needs for fresh and used transmutation fuels is one of these criteria. Transmutation fuels have specific characteristics in terms of thermal load and neutron emissions. Thermal, radiation and criticality constraints have been taken into account in this study to suggest cask concepts for routine conditions of transport, to estimate the number of assemblies to be transported in a cask and the number of annual transports. Comparison with the no transmutation option, i.e. management of uranium and plutonium in SFRs, is also presented. Regarding these matters, no high difficulties appear for assemblies with limited content of Am (homogeneous or heterogeneous recycling modes). When fuels contain curium, technical transport uncertainties increase because of the important heat release requiring dividing fresh fuels and technological innovations development (MABB and ADS). (authors)

  9. Device for taking gaseous samples from irradiated fuel elements

    International Nuclear Information System (INIS)

    Lengacker, B.

    1983-01-01

    The described device allows to take gaseous samples from irradiated fuel elements. It is connected with a gas analyzer and a pressure gage, so that in opening the fuel can the internal pressure can be determined

  10. Safety analysis report for packaging: the ORNL HFIR unirradiated fuel element shipping container

    International Nuclear Information System (INIS)

    Evans, J.H.; Boulet, J.A.M.; Eversole, R.E.

    1977-11-01

    The ORNL HFIR unirradiated fuel element shipping container was designed and fabricated at the Oak Ridge National Laboratory for the transport of HFIR unirradiated fuel elements. The container was evaluated analytically and experimentally to determine its compliance with the applicable regulations governing containers in which radioactive and fissile materials are transported, and the evaluation is the subject of this report. Computational and test procedures were used to determine the structural integrity and thermal behavior of the cask relative to the general standards for normal conditions of transport and the standards for the hypothetical accident conditions. The results of the evaluation demonstrate that the container is in compliance with the applicable regulations

  11. Fuels Performance | Transportation Research | NREL

    Science.gov (United States)

    Biodiesel Performance in Modern Engines NREL is working cooperatively with the National Biodiesel Board on research that assesses several aspects of biodiesel compatibility with engines, vehicles, and fuel-handling infrastructure. Several cooperative studies were recently conducted that examined how biodiesel performs with new

  12. Thermal-hydraulic investigations of fuel elements

    International Nuclear Information System (INIS)

    Rehme, K.; Weinberg, D.

    1983-01-01

    Extensive fluid-dynamic examining of flow distribution and turbulent flow distribution was done to control and safeguard calculation methods allowing the determination of three-dimensional flow distribution in fuel elements. Results show that the flow distribution greatly depends on the frequency of pulse exchange between subchannels in narrow rod grids. The comparison of these measured values to VELASCO's results shows that the calculation methods need to be considerably improved. The subchannel analysis proved to be very suitable to calculate mean flow temperatures conforming with the subchannel analysis principle. However, this does not include statements on wall temperatures occurring in the structures. Mean wall temperatures can be determined by empirical interrelationships for Nusseltnumbers. On the other hand, the calculation of detailed wall temperature distributions is not possible with the subchannel analysis unless it can be further improved due to more detailed measurement results. (orig.) [de

  13. Natural uranium metallic fuel elements: fabrication and operating experience

    International Nuclear Information System (INIS)

    Hammad, F.H.; Abou-Zahra, A.A.; Sharkawy, S.W.

    1980-01-01

    The main reactor types based on natural uranium metallic fuel element, particularly the early types, are reviewed in this report. The reactor types are: graphite moderated air cooled, graphite moderated gas cooled and heavy water moderated reactors. The design features, fabrication technology of these reactor fuel elements and the operating experience gained during reactor operation are described and discussed. The interrelation between operating experience, fuel design and fabrication was also discussed with emphasis on improving fuel performance. (author)

  14. Study of fuel element characteristic of SM and SMP (SM-PRIMA) fuel assemblies

    International Nuclear Information System (INIS)

    Klinov, A.V.; Kuprienko, V.A.; Lebedev, V.A.; Makhin, V.M.; Tuchnin, L.M.; Tsykanov, V.A.

    1999-01-01

    The paper discusses the techniques and results of reactor tests and post-reactor investigations of the SM reactor fuel elements and fuel elements developed in the process of designing the specialized PRIMA test reactor with the SM reactor fuel elements used as a prototype and which are referred to as the SMP fuel elements. The behavior of fuel elements under normal operating conditions and under deviation from normal operating conditions was studied to verify the calculation techniques, to check the calculation results during preparation of the SM reactor safety substantiation report and to estimate the possibility of using such fuel elements in other projects. During tests of fuel rods under deviation from normal operating conditions their advantages were shown over fuel elements, the components of which were produced using the Al-based alloys. (author)

  15. Some properties for modeling of fuel elements

    International Nuclear Information System (INIS)

    Nichols, F.A.

    1979-01-01

    Two areas key to the materials modeling of fuel element behavior are discussed. The relative importance of atomic diffusion vs. bubble migration is first surveyed and the interplay of bubble mobility and re-solution parameter is highlighted. It is concluded that biased bubble migration at higher temperatures is required to explain available gas-release data, especially during transients. At intermediate temperatures, random bubble migration is required to explain both gas-release rates and the observation of large (approx. 700A) intragranular bubbles following in-pile and post-irradiation transients. Different fuel models employ different values of re-solution parameter, both below and above an experimentally determined value. Bubble mobilities are deduced to approach theoretical, surface diffusion-controlled values during transients, but they may be somewhat less mobile during steady-state operation. Next, the present understanding of radiation-induced hardening and creep is discussed, highlighting the interplay of these two phenomena. An overall constitutive scheme is presented and predictions of failure limits are deduced therefrom employing instability analysis

  16. Nuclear criticality safety assessment of ORR, NBS, and HFBR fuel element shipping package

    International Nuclear Information System (INIS)

    Thomas, J.T.

    1979-01-01

    A fuel element shipping package employing a borated-phenolic foam as a thermal insulating material is designed to transport as many as seven fuel elements for use in the Oak Ridge Research Reactor, the Brookhaven Fast Beam Reactor, or the National Bureau of Standards Reactor. This report presents the criticality safety evaluation and demonstrates that the requirements for a Fissile Class I package are satisfied by the design

  17. TRIGA Mark II Ljubljana - spent fuel transportation

    International Nuclear Information System (INIS)

    Ravnik, M.; Dimic, V.

    2008-01-01

    The most important activity in 1999 was shipment of the spent fuel elements back to the United States for final disposal. This activity started already in 1998 with some governmental support. In July 1999 all spent fuel elements (219 pieces) from the TRIGA research reactor in Ljubljana were shipped back to the United Stated by the ship from the port Koper in Slovenia. At the same time shipment of the spent fuel from the research reactor in Pitesti, Romania, and the research reactor in Rome, Italy, was conducted. During the loading the radiation exposure to the workers was rather low. The loading and shipment of the spent nuclear fuel went very smoothly and according the accepted time table. During the last two years the TRIGA research reactor in Ljubljana has been in operation about 1100 hours per year and without any undesired shut-down. (authors)

  18. ETR fuel element shipping container addendum to PR-T-79-011 (TR-466). Internal technical report

    International Nuclear Information System (INIS)

    Smith, M.C.

    1979-01-01

    In July, 1979, EG and G Idaho, Inc. was requested to evaluate the ETR Fuel Element Shipping Container for compliance with existing transport regulations, in order to ship GETR fuel elements from Vallecitos, California to the INEL. Technical report PR-T-79-011 (TR-466), ATR Fuel Element Shipping Container Safety Analysis, was used as a basis for this evaluation. The safety analysis contained in technical report PR-T-79-011 (TR-466) was performed utilizing the ATR, ETR, MTR, and SPERT shipping containers. The report determined the ETR Fuel Element Shipping Container does comply with the existing transport regulations for a Type A quantity, Fissile Class I shipping container. The ETR and GETR fuel elements are essentially identical in physical size, construction, and fissile material content, the analysis documented in this report has determined the shipment of GETR fuel elements in the ETR shipping container to be safe and pose no threat to the public health and safety

  19. Further developments of PWR and BWR fuel elements

    International Nuclear Information System (INIS)

    Sofer, G.A.; Busselman, G.J.; Federico, L.J.

    1988-01-01

    The performance, safety, and economy of nuclear power plants in inluenced very decisively by the quality of their fuel elements. This is why quality assurance in fuel fabrication has been a factor of great importance from the outset. Operating experince and more stringent performance requirements have resulted in a continuous process of further development of fuel elements, which has been reflected also in lower and lower failure rates and increasingly higher burn-ups. Next to further development also innovation has been an important factor contributing to the present high quality level of fuel elements, which also has allowed fuel cycle costs to be decreased quite considerably. (orig.) [de

  20. Premiering SAFE for Safety Added Fuel Element - 15020

    International Nuclear Information System (INIS)

    Bhowmik, P.K.; Shamim, J.A.; Suh, K.Y.; Suh, K.S.

    2015-01-01

    The impact of the Fukushima accident has been the willingness to implement passive safety measures in reactor design and to simplify reactor design itself. Within this framework, a new fuel element, named SAFE (Safety Added Fuel Element) based on the concept of accident tolerant fuel, is presented. SAFE is a new type of fuel element cooled internally and externally by light water and with stainless steel as the cladding material. The removal of boron may trigger a series of changes which may simplify the system greatly. A simplified thermal analysis of SAFE shows that the fuel centerline temperature is well below the maximal limit during the normal operation of the plant

  1. News from the fuel elements industry

    International Nuclear Information System (INIS)

    Racine, R.; Delannay, M.; Dehon, C.; Jouan, J.; Beuneche, M.

    1981-01-01

    This article deals successively with: the re-structuring of the PWR fuel industry in France, with the setting up of Fragema and Cogema Framatome Combustible; Fragema products, from standard fuel assembly to the development of a new advanced fuel assembly; Framatome's experience with PWR fuel; fuel performances in the light of requirements imposed by network needs follow-up; devices developed by Fragema for on-site analysis of irradiated fuel [fr

  2. The development of fuel elements for boiling water reactors

    International Nuclear Information System (INIS)

    Holzer, R.; Kilian, P.

    1984-01-01

    The longevity of today's standard fuel elements constitutes a sound basis for designing advanced fuel elements for higher discharge burnups. Operating experience as well as postirradiation examinations of discharged fuel elements indicate that the technical limits have not reached by far. However, measures to achieve an economic and reliable fuel cycle are not restricted to the design of fuel elements, but also extend into such fields as fuel management and the mode of reactor operation. Fuel elements can be grouped together in zones in the core as a function of burnup and reactivity. The loading scheme can be aligned to this approach by concentrating on typical control rod positions. Reloads can also be made up of two sublots of fuel elements with different gadolinium contents. Longer cycles, e.g., of eighteen instead of twelve months, are easy to plan reactivitywise by increasing the quantity to be replaced from at present one quarter to one third. In fuel elements designed for higher burnups, the old scheme of reloading one quarter of the fuel inventory can be retained. The measures already introduced or in the planning stage incorporate a major potential for technical and economic optimization of the fuel cycle in boiling water reactors. (orig.) [de

  3. Fuel cell assembly with electrolyte transport

    Science.gov (United States)

    Chi, Chang V.

    1983-01-01

    A fuel cell assembly wherein electrolyte for filling the fuel cell matrix is carried via a transport system comprising a first passage means for conveying electrolyte through a first plate and communicating with a groove in a second plate at a first point, the first and second plates together sandwiching the matrix, and second passage means acting to carry electrolyte exclusively through the second plate and communicating with the groove at a second point exclusive of the first point.

  4. Testing of a transport cask for research reactor spent fuel

    International Nuclear Information System (INIS)

    Mourao, Rogerio P.; Silva, Luiz Leite da; Miranda, Carlos A.; Mattar Neto, Miguel; Quintana, Jose F.A.; Saliba, Roberto O.; Novara, Oscar E.

    2011-01-01

    Since the beginning of the last decade three Latin American countries which operate research reactors - Argentina, Brazil and Chile - have been joining efforts to improve the regional capability in the management of spent fuel elements from the reactors operated in the region. As a step in this direction, a packaging for the transport of irradiated fuel from research reactors was designed by a tri-national team and a half-scale model for MTR fuel constructed in Argentina and tested in Brazil. Two test campaigns have been carried out so far, covering both normal conditions of transportation and hypothetical accident conditions. Although the specimen has not successfully performed the tests, its overall performance was considered very satisfactory, and improvements are being introduced to the design. A third test sequence is planned for 2011. (author)

  5. Fabrication technology of spherical fuel element for HTR-10

    International Nuclear Information System (INIS)

    He Jun; Zou Yanwen; Liang Tongxiang; Qiu Xueliang

    2002-01-01

    R and D on the fabrication technology of the spherical fuel elements for the 10 MW HTR Test Module (HTR-10) began from 1986. Cold quasi-isostatic molding with a silicon rubber die is used for manufacturing the spherical fuel elements.The fabrication technology and the graphite matrix materials were investigated and optimized. Twenty five batches of fuel elements, about 11000 of the fuel elements, have been produced. The cold properties of the graphite matrix materials satisfied the design specifications. The mean free uranium fraction of 25 batches was 5 x 10 -5

  6. Effects of fueling profiles on plasma transport

    International Nuclear Information System (INIS)

    Mense, A.T.; Houlberg, W.A.; Attenberger, S.E.; Milora, S.L.

    1978-04-01

    A one-dimensional (1-D), multifluid transport model is used to investigate the effects of particle fueling profiles on plasma transport in an ignition-sized tokamak (TNS). Normal diffusive properties of plasmas will likely maintain the density at the center of the discharge even if no active fueling is provided there. This significantly relaxes the requirements for fuel penetration. Not only is lower fuel penetration easier to achieve, but it may have the advantage of reducing or eliminating density gradient-driven trapped particle microinstabilities. Simulation of discrete pellet fueling indicates that relatively low velocity (approximately 10 3 m/sec) pellets may be sufficient to fuel a TNS-sized device (approximately 1.25-m minor radius), to produce a relatively broad, cool edge region of plasma which should reduce the potential for sputtering, and also to reduce the likelihood of trapped particle mode dominated transport. Low penetrating pellets containing up to 10 to 20 percent of the total plasma ions can produce fluctuations in density and temperature at the plasma edge, but the pressure profile and fusion alpha production remain almost constant

  7. Alternative transport fuels: supply, consumption and conservation

    International Nuclear Information System (INIS)

    Trindade, S.C.

    1990-01-01

    Road-based passenger and freight transport almost exclusively uses petroleum/hydrocarbon fuels in the fluid form. These fuels will probably continue to be major transport fuels well into the 21st century. As such there is need to prolong their use which can be done through: (1) conservation of fuel by increasing efficiency of internal combustion engines, and (2) conversion of natural gas, coal and peat, and biomass into alternate fuels such as ethanol, methanol, CNG, LNG, LPG, low heat-content (producer) gas and vegetable oils. Research, development and demonstration (RD and D) priorities in supply, consumption and conservation of these alternate fuels are identified and ranked in the context of situation prevailing in Brazil. Author has assigned the highest priority for research in the impact of pricing, economic, fiscal and trade policies, capital allocation criteria and institutional and legislative framework. It has also been emphasised that an integrated or systems approach is mandatory to achieve net energy gains in transport sector. (M.G.B.). 33 refs., 11 tabs., 4 figs

  8. Considerations for the transportation of spent fuel

    International Nuclear Information System (INIS)

    Jefferson, R.M.

    1984-01-01

    In our society today the transportation of radioactive materials, and most particularly spent reactor fuel, is surrounded by considerable emotion and a wealth of information, good and bad. The transportation of these materials is viewed as unique and distinct from the transportation of other hazardous materials and as a particularly vulnerable component of the nuclear power activities of this nation. Added to this is the concept, widely held, that almost everyone is an expert on the transportation of radioactive materials. One significant contribution to this level of emotion is the notion that all roads (rail and highway), on which these goods will be transported, somehow traverse everyone's backyard. The issue of the transportation of spent fuel has thus become a political battleground. Perhaps this should not be surprising since it has all of the right characteristics for such politicization in that it is pervasive, emotional, and visible. In order that those involved in the discussion of this activity might be able to reach some rational conclusions, this paper offers some background information which might be useful to a broad range of individuals in developing their own perspectives. The intent is to address the safety of transporting spent fuel from a technical standpoint without the emotional content which is frequently a part of this argument

  9. Study of fuel systems for LH2-fueled subsonic transport aircraft, volume 1

    Science.gov (United States)

    Brewer, G. D.; Morris, R. E.; Davis, G. W.; Versaw, E. F.; Cunnington, G. R., Jr.; Riple, J. C.; Baerst, C. F.; Garmong, G.

    1978-01-01

    Several engine concepts examined to determine a preferred design which most effectively exploits the characteristics of hydrogen fuel in aircraft tanks received major emphasis. Many candidate designs of tank structure and cryogenic insulation systems were evaluated. Designs of all major elements of the aircraft fuel system including pumps, lines, valves, regulators, and heat exchangers received attention. Selected designs of boost pumps to be mounted in the LH2 tanks, and of a high pressure pump to be mounted on the engine were defined. A final design of LH2-fueled transport aircraft was established which incorporates a preferred design of fuel system. That aircraft was then compared with a conventionally fueled counterpart designed to equivalent technology standards.

  10. Spent fuel storage and transportation - ANSTO experience

    International Nuclear Information System (INIS)

    Irwin, Tony

    2002-01-01

    The Australian Nuclear Science and Technology Organisation (ANSTO) has operated the 10 MW DIDO class High Flux Materials Test Reactor (HIFAR) since 1958. Refuelling the reactor produces about 38 spent fuel elements each year. Australia has no power reactors and only one operating research reactor so that a reprocessing plant in Australia is not an economic proposition. The HEU fuel for HIFAR is manufactured at Dounreay using UK or US origin enriched uranium. Spent fuel was originally sent to Dounreay, UK for reprocessing but this plant was shutdown in 1998. ANSTO participates in the US Foreign Research Reactor Spent Fuel Return program and also has a contract with COGEMA for the reprocessing of non-US origin fuel

  11. Electricity: an indigenous transport fuel

    Energy Technology Data Exchange (ETDEWEB)

    Byers, D J

    1978-12-25

    Potential reserves of hydro and geothermal power are adequate to power all road-transport vehicles should these be converted to electric drives in the future. Conversion of petrol vehicles to electric drives results in a significant increase in energy-utilization efficiency coupled with a decrease in costs, both to the country in overseas funds and to the driver in operating costs. As yet, however, New Zealand has no plan to use these resources in a transport role and is supporting no feasibility research.

  12. TAPIR, Thermal Analysis of HTGR with Graphite Sleeve Fuel Elements

    International Nuclear Information System (INIS)

    Weicht, U.; Mueller, W.

    1983-01-01

    1 - Nature of the physical problem solved: Thermal analysis of a reactor core containing internally and/or externally gas cooled prismatic fuel elements of various geometries, rating, power distribution, and material properties. 2 - Method of solution: A fuel element in this programme is regarded as a sector of a fuelled annulus with graphite sleeves of any shape on either side and optional annular gaps between fuel and graphite and/or within the graphite. It may have any centre angle and the fuelled annulus may become a solid cylindrical rod. Heat generation in the fuel is assumed to be uniform over the cross section and peripheral heat flux into adjacent sectors is ignored. Fuel elements and coolant channels are treated separately, then linked together to fit a specified pattern. 3 - Restrictions on the complexity of the problem: Maxima of: 50 fuel elements; 50 cooled channels; 25 fuel geometries; 25 coolant channel geometries; 10 axial power distributions; 10 graphite conductivities

  13. Fuel element transfer cask modelling using MCNP technique

    International Nuclear Information System (INIS)

    Rosli Darmawan

    2009-01-01

    Full text: After operating for more than 25 years, some of the Reaktor TRIGA PUSPATI (RTP) fuel elements would have been depleted. A few addition and fuel reconfiguration exercises have to be conducted in order to maintain RTP capacity. Presently, RTP spent fuels are stored at the storage area inside RTP tank. The need to transfer the fuel element outside of RTP tank may be prevalence in the near future. The preparation shall be started from now. A fuel element transfer cask has been designed according to the recommendation by the fuel manufacturer and experience of other countries. A modelling using MCNP code has been conducted to analyse the design. The result shows that the design of transfer cask fuel element is safe for handling outside the RTP tank according to recent regulatory requirement. (author)

  14. Fuel Element Transfer Cask Modelling Using MCNP Technique

    International Nuclear Information System (INIS)

    Darmawan, Rosli; Topah, Budiman Naim

    2010-01-01

    After operating for more than 25 years, some of the Reaktor TRIGA Puspati (RTP) fuel elements would have been depleted. A few addition and fuel reconfiguration exercises have to be conducted in order to maintain RTP capacity. Presently, RTP spent fuels are stored at the storage area inside RTP tank. The need to transfer the fuel element outside of RTP tank may be prevalence in the near future. The preparation shall be started from now. A fuel element transfer cask has been designed according to the recommendation by the fuel manufacturer and experience of other countries. A modelling using MCNP code has been conducted to analyse the design. The result shows that the design of transfer cask fuel element is safe for handling outside the RTP tank according to recent regulatory requirement.

  15. Methodology for substantiation of the fast reactor fuel element serviceability

    International Nuclear Information System (INIS)

    Tsykanov, V.A.; Maershin, A.A.

    1988-01-01

    Methodological aspects of fast reactor fuel element serviceability substantiation are presented. The choice of the experimental program and strategies of its realization to solve the problem set in short time, taking into account available experimental means, are substantiated. Factors determining fuel element serviceability depending on parameters and operational conditions are considered. The methodological approach recommending separate studing of the factors, which points to the possibility of data acquisition, required for the development of calculational models and substantiation of fuel element serviceability in pilot and experimental reactors, is described. It is shown that the special-purpose data are more useful for the substantiation of fuel element serviceability and analytical method development than unsubstantial and expensive complex tests of fuel elements and fuel assemblies, which should be conducted only at final stages for the improvement of the structure on the whole

  16. Reproduction of the RA reactor fuel element fabrication

    International Nuclear Information System (INIS)

    Novakovic, M.

    1961-12-01

    This document includes the following nine reports: Final report on task 08/12 - testing the Ra reactor fuel element; design concept for fabrication of RA reactor fuel element; investigation of the microstructure of the Ra reactor fuel element; Final report on task 08/13 producing binary alloys with Al, Mo, Zr, Nb and B additions; fabrication of U-Al alloy; final report on tasks 08/14 and 08/16; final report on task 08/32 diffusion bond between the fuel and the cladding of the Ra reactor fuel element; Final report on task 08/33, fabrication of the RA reactor fuel element cladding; and final report on task 08/36, diffusion of solid state metals [sr

  17. Fuel element structure - design, production and operational behaviour

    International Nuclear Information System (INIS)

    Pott, G.; Dietz, W.

    1985-01-01

    The lectures held at the meeting of the fuel element section of the Kerntechnische Gesellschaft gives a survey of developments in fuel element structure design for PWR-type, BWR-type and fast breeder reactors. For better utilization of the fuel, concepts have been developed for re-usable, removable and thus repairable fuel elements. Furthermore, the manufacturing methods for fuel element structures were refined to achieve better quality and more efficient manufacturing methods. Statements on the dimensional behaviour and on the mechanical stability of fuel element structures in normal and accident operation could be made on the basis of post-irradiation inspections. Finally, the design, manufacture and irradiation behaviour of graphite reflectors in HTGR-type reactors are described. The 12 lectures have been recorded in the data base separately. (RF) [de

  18. Determining fissile content of nuclear fuel elements

    International Nuclear Information System (INIS)

    Arya, S.P.; Grossman, L.N.; Schoenig, F.C.

    1980-01-01

    This invention relates to the determination of the fissile fuel content of fuel for nuclear reactors. A nondestructive method is described for determining rapidly, accurately and simultaneously the fissile content, enrichment and location of fuel material which may also contain amounts of burnable poison, by detecting the γ-rays emitted from the fuel material due to natural radioactive decay. (U.K.)

  19. Storage device for a long nuclear reactor fuel element and/or a long nuclear reactor fuel element part

    International Nuclear Information System (INIS)

    Vogt, M.; Schoenwitz, H.P.; Dassbach, W.

    1986-01-01

    The storage device can be erected in a dry storage room for new fuel elements and also in a storage pond for irradiated fuel elements. It consists of shells, which are arranged vertically and which have a lid. A suspension for the fuel element is provided on the underside of the lid, which acts as a support against squashing or bending in case of vertical forces acting (earthquake). (DG) [de

  20. Coherence of reactor design and fuel element design

    International Nuclear Information System (INIS)

    Vom Scheidt, S.

    1995-01-01

    Its background of more than 25 years of experience makes Framatome the world's leading company in the design and sales of fuel elements for pressurized water reactors (PWR). In 1994, the fuel fabrication units were incorporated as subsidiaries, which further strengthens the company's position. The activities in the fuel sector comprise fuel element design, selection and sourcing of materials, fuel element fabrication, and the services associated with nuclear fuel. Design responsibility lies with the Design and sales Management, which closely cooperates with the engineers of the reactor plant for which the fuel elements are being designed, for fuel elements are inseparable parts of the respective reactors. The Design and Sales Management also has developed a complete line of services associated with fuel element inspection and repair. As far as fuel element sales are concerned, Framatome delivers the first core in order to be able to assume full responsibility vis-a-vis the customer for the performance of the nuclear steam supply system. Reloads are sold through the Fragema Association established by Framatome and Cogema. (orig.) [de

  1. Spent fuel transportation regulatory and institutional issues

    International Nuclear Information System (INIS)

    Lippek, H.E.

    1978-01-01

    The problems that could result from state and local governments and other groups with relation to regulations concerning the transportation of spent nuclear fuels are discussed. The powers of the individual states as spelled out in the Clean Air Act Amendments of 1977 are set forth in some detail. The possibility of transportation employees gaining a position to demand and receive more stringent protections from hazards of radiation is pointed out

  2. Transport and storage of spent nuclear fuel

    International Nuclear Information System (INIS)

    Lung, M.; Lenail, B.

    1987-01-01

    From a safety standpoint, spent fuel is clearly not ideal for permanent disposal and reprocessing is the best method of preparing wastes for long-term storage in a repository. Furthermore, the future may demonstrate that some fission products recovered in reprocessing have economic applications. Many countries have in fact reached the point at which the recycling of plutonium and uranium from spent fuel is economical in LWR's. Even in countries where this is not yet evident, (i.e., the United States), the French example shows that the day will come when spent fuel will be retrieved for reprocessing and recycle. It is highly questionable whether spent fuel will ever be considered and treated as waste in the same sense as fission products and processed as such, i.e., packaged in a waste form for permanent disposal. Even when recycled fuel material can no longer be reused in LWR's because of poor reactivity, it will be usable in FBR's. Based on the considerable experience gained by SGN and Cogema, this paper has provided practical discussion and illustrations of spent fuel transport and storage of a very important step in the nuclear fuel management process. The best of spent fuel storage depends on technical, economic and policy considerations. Each design has a role to play and we hope that the above discussion will help clarify certain issues

  3. Drying results of K-Basin fuel element 5744U (Run 4)

    International Nuclear Information System (INIS)

    Klinger, G.S.; Oliver, B.M.; Abrefah, J.; Marschman, S.C.; MacFarlan, P.J.; Ritter, G.A.

    1998-07-01

    The water-filled K-Basins in the Hanford 100 Area have been used to store N-Reactor spent nuclear fuel (SNF) since the 1970s. Because some leaks in the basins have been detected and some of the fuel is breached due to handling damage and corrosion, efforts are underway to remove the fuel elements from wet storage. An Integrated Process Strategy (IPS) has been developed to package, dry, transport, and store these metallic uranium fuel elements in an interim storage facility on the Hanford Site. Information required to support the development of the drying processes, and the required safety analyses, is being obtained from characterization tests conducted on fuel elements removed from the K-Basins. A series of whole element drying tests (reported in separate documents, see Section 8.0) have been conducted by Pacific Northwest National Laboratory (PNNL) on several intact and damaged fuel elements recovered from both the K-East and K-West Basins. This report documents the results of the fourth of those tests, which was conducted on an N-Reactor outer fuel element removed from K-West canister 5744U. This element (referred to as Element 5744U) was stored underwater in the K-West Basin from 1983 until 1996. Element 5744U was subjected to a combination of low- and high-temperature vacuum drying treatments that were intended to mimic, wherever possible, the fuel treatment strategies of the IPS. The system used for the drying test was the Whole Element Furnace Testing System, described in Section 2.0, located in the Postirradiation Testing Laboratory (PTL, 327 Building). The test conditions and methodologies are given in Section 3.0. Inspections of the fuel element before and after the test are provided in Section 4.0. The experimental results are provided in Section 5.0, and discussed in Section 6.0

  4. Drying Results of K-Basin Fuel Element 2660M (Run 7)

    International Nuclear Information System (INIS)

    Oliver, B.M.; Klinger, G.S.; Abrefah, J.; Marschman, S.C.; MacFarlan, P.J.; Ritter, G.A.

    1999-01-01

    The water-filled K-Basins in the Hanford 100 Area have been used to store N-Reactor spent nuclear fuel (SNF) since the 1970s. Because some leaks in the basin have been detected and some of the fuel is breached due to handling damage and corrosion, efforts are underway to remove the fuel elements from wet storage. An Integrated Process Strategy (IPS) has been developed to package, dry, transport, and store these metallic uranium fuel elements in an interim storage facility on the Hanford Site (WHC 1995). Information required to support the development of the drying processes, and the required safety analyses, is being obtained from characterization tests conducted on fuel elements removed from the K-Basins. A series of whole element drying tests (reported in separate documents, see Section 8.0) have been conducted by Pacific Northwest National Laboratory (PNNL) on several intact and damaged fuel elements recovered from both the K-East and K-West Basins. This report documents the results of the seventh of those tests, which was conducted on an N-Reactor outer fuel element removed from K-West canister 2660M. This element (referred to as Element 2660M) was stored underwater in the K-West Basin from 1983 until 1996. Element 2660M was subjected to a combination of low- and high-temperature vacuum drying treatments that were intended to mimic, wherever possible, the fuel treatment strategies of the IPS. The system used for the drying test was the Whole Element Furnace Testing System, described in Section 2.0, located in the Postirradiation Testing Laboratory (PTL, 327 Building). The test conditions and methodologies are given in Section 3.0. Inspections of the fuel element before and after the test are provided in Section 4.0. The experimental results are provided in Section 5.0, and discussed in Section 6.0

  5. Drying Results of K-Basin Fuel Element 6513U (Run 8)

    International Nuclear Information System (INIS)

    Oliver, B.M.; Klinger, G.S.; Abrefah, J.; Marschman, S.C.; MacFarlan, P.J.; Ritter, G.A.

    1999-01-01

    The water-filled K-Basins in the Hanford 100 Area have been used to store N-Reactor spent nuclear fuel (SNF) since the 1970s. Because some leaks in the basin have been detected and some of the fuel is breached due to handling damage and corrosion, efforts are underway to remove the fuel elements from wet storage. An Integrated Process Strategy (IPS) has been developed to package, dry, transport, and store these metallic uranium fuel elements in an interim storage facility on the Hanford Site (WHC 1995). Information required to support the development of the drying processes, and the required safety analyses, is being obtained from characterization tests conducted on fuel elements removed from the K-Basins. A series of whole element drying tests (reported in separate documents, see Section 8.0) have been conducted by Pacific Northwest National Laboratory (PNNL)on several intact and damaged fuel elements recovered from both the K-East and K-West Basins. This report documents the results of the eighth of those tests, which was conducted on an N-Reactor outer fuel element removed from K-West canister 6513U. This element (referred to as Element 6513U) was stored underwater in the K-West Basin from 1983 until 1996. Element 6513U was subjected to a combination of low- and high-temperature vacuum drying treatments that were intended to mimic, wherever possible, the fuel treatment strategies of the IPS. The system used for the drying test was the Whole Element Furnace Testing System, described in Section 2.0, located in the Postirradiation Testing Laboratory (PTL, 327 Building). The test conditions and methodologies are given in Section 3.0. Inspections of the fuel element before and after the test are provided in Section 4.0. The experimental results are provided in Section 5.0 and discussed in Section 6.0

  6. Drying results of K-Basin fuel element 1164M (run 6)

    International Nuclear Information System (INIS)

    Oliver, B.M.; Klinger, G.S.; Abrefah, J.; Marschman, S.C.; MacFarlan, P.J.; Ritter, G.A.

    1998-08-01

    The water-filled K-Basins in the Hanford 100 Area have been used to store N-Reactor spent nuclear fuel (SNF) since the 1970s. Because some leaks in the basin have been detected and some of the fuel is breached due to handling damage and corrosion, efforts are underway to remove the fuel elements from wet storage. An Integrated Process Strategy (IPS) has been developed to package, dry, transport, and store these metallic uranium fuel elements in an interim storage facility on the Hanford site (WHC 1995). Information required to support the development of the drying processes, and the required safety analyses, is being obtained from characterization tests conducted on fuel elements removed from the K-Basins. A series of whole element drying tests (reported in separate documents, see Section 8.0) have been conducted by Pacific Northwest National Laboratory (PNNL) on several intact and damaged fuel elements recovered from both the K-East and K-West Basins. This report documents the results of the sixth of those tests, which was conducted on an N-Reactor outer fuel element removed from K-West canister 1164 M. This element (referred to as Element 1164M) was stored underwater in the K-West Basin from 1983 until 1996. Element 1164M was subjected to a combination of low- and high-temperature vacuum drying treatments that were intended to mimic, wherever possible, the fuel treatment strategies of the IPS. The system used for the drying test was the Whole Element Furnace Testing System, described in Section 2.0, located in the Postirradiation Testing laboratory (PTL, 327 Building). The test conditions and methodologies are given in Section 3.0. Inspections of the fuel element before and after the test are provided in Section 4.0. The experimental results are provided in Section 5.0, and discussed in Section 6.0

  7. Discrete elements method of neutral particle transport

    International Nuclear Information System (INIS)

    Mathews, K.A.

    1983-01-01

    A new discrete elements (L/sub N/) transport method is derived and compared to the discrete ordinates S/sub N/ method, theoretically and by numerical experimentation. The discrete elements method is more accurate than discrete ordinates and strongly ameliorates ray effects for the practical problems studied. The discrete elements method is shown to be more cost effective, in terms of execution time with comparable storage to attain the same accuracy, for a one-dimensional test case using linear characteristic spatial quadrature. In a two-dimensional test case, a vacuum duct in a shield, L/sub N/ is more consistently convergent toward a Monte Carlo benchmark solution than S/sub N/, using step characteristic spatial quadrature. An analysis of the interaction of angular and spatial quadrature in xy-geometry indicates the desirability of using linear characteristic spatial quadrature with the L/sub N/ method

  8. Environmental economics of lignin derived transport fuels

    NARCIS (Netherlands)

    Obydenkova, Svetlana V.; Kouris, Panos D.; Hensen, Emiel J. M.; Heeres, Hero J.; Boot, Michael D.

    2017-01-01

    This paper explores the environmental and economic aspects of fast pyrolytic conversion of lignin, obtained from 2G ethanol plants, to transport fuels for both the marine and automotive markets. Various scenarios are explored, pertaining to aggregation of lignin from several sites, alternative

  9. Control system of fuel transporting device

    International Nuclear Information System (INIS)

    Yokota, Minoru.

    1981-01-01

    Purpose: To effectively avoid an obstacle in a fuel transporting device by reading the outputs of absolute position detectors mounted on movable trucks, controlling the movements of the trucks, and thereby smoothly and accurately positioning the fuel transporting device at predetermined position and providing a contact detector thereat. Method: The outputs from absolute position detectors which are mounted on a longitudinally movable truck and a laterally movable truck are input to an input/output control circuit. The input/output control circuit serves to compare, the position a fuel transporting device is to be moved to, with the present position on the basis of said input detection signal and a command signal from an operator console, to calculate the amount of movement to be driven, to produce an operation signal therefor to a control panel, and to drive and control the drive motors which are respectively mounted on the trucks for the fuel transfer device. On the other hand, in case that the transfer device comes into contact with an obstacle, the contact detector will immediately operate to produce a stop command through the control panel to the transporting device, and avoid a collision with the obstacle. (Yoshino, Y.)

  10. Experience related to the safety of advanced LMFBR fuel elements

    International Nuclear Information System (INIS)

    Kerrisk, J.F.

    1975-07-01

    Experiments and experience relative to the safety of advanced fuel elements for the liquid metal fast breeder reactor are reviewed. The design and operating parameters and some of the unique features of advanced fuel elements are discussed breifly. Transient and steady state overpower operation and loss of sodium bond tests and experience are discussed in detail. Areas where information is lacking are also mentioned

  11. Design and main characteristics of HTGR fuel elements

    International Nuclear Information System (INIS)

    Chernikov, A.S.; Kolesov, V.S.; Permyakov, L.N.; Koshelev, Yu.V.; Mikhajlichenko, L.I.

    1983-01-01

    Two types of spherical fuel elements and coated particles were investigated under the operating conditions of the high temperature reactors in the Soviet Union (VGR-50 and VG-400). This paper gives the main characteristics of spherical fuel elements (thermal conductivity, static and dynamic strength, wear resistance, release of gaseous fission products, etc.) as determined in test facilities. (author)

  12. Hastelloy X fuel element creep relaxation and residual effects

    International Nuclear Information System (INIS)

    Castle, R.A.

    1971-01-01

    A worst case, seven element, asymmetric fuel, thermal environment was assumed and a creep relaxation analysis generated. The fuel element clad is .020 inch Hastelloy X. The contact load decreased from 11.6 pounds to 5.87 pounds in 100,000 hours. The residual stresses were then computed for various shutdown times. (U.S.)

  13. Legal questions concerning the termination of spent fuel element reprocessing

    International Nuclear Information System (INIS)

    John, Michele

    2005-01-01

    The thesis on legal aspects of the terminated spent fuel reprocessing in Germany is based on the legislation, jurisdiction and literature until January 2004. The five chapters cover the following topics: description of the problem; reprocessing of spent fuel elements in foreign countries - practical and legal aspects; operators' responsibilities according to the atomic law with respect to the reprocessing of Geman spent fuel elements in foreign countries; compatibility of the prohibition of Geman spent fuel element reprocessing in foreign countries with international law, European law and German constitutional law; results of the evaluation

  14. Attempt to produce silicide fuel elements in Indonesia

    International Nuclear Information System (INIS)

    Soentono, S.; Suripto, A.

    1991-01-01

    After the successful experiment to produce U 3 Si 2 powder and U 3 Si 2 -Al fuel plates using depleted U and Si of semiconductor quality, silicide fuel was synthesized using x -Al available at the Fuel Element Production Installation (FEPI) at Serpong, Indonesia. Two full-size U 3 Si 2 -Al fuel elements, having similar specifications to the ones of U 3 O 8 -Al for the RSG-GAS (formerly known as MPR-30), have been produced at the FEPI. All quality controls required have been imposed to the feeds, intermediate, as well as final products throughout the production processes of the two fuel elements. The current results show that these fuel elements are qualified from fabrication point of view, therefore it is expected that they will be permitted to be tested in the RSG-GAS, sometime by the end of 1989, for normal (∝50%) and above normal burn-up. (orig.)

  15. Improved techniques for appendage attachment to PHWR fuel elements

    International Nuclear Information System (INIS)

    Raj, R.N.J.; Laxminarayana, B.; Narayanan, P.S.A.; Gupta, U.C.; Varma, B.P.; Sinha, K.K.

    1995-01-01

    Nuclear Fuel Complex, India switched-over to split-wart type PHWR fuel bundles in mid-80s. Since then over 60,000 bundles of this type have been fabricated for Indian PHWRs. After considering various technical aspects, resistance welding was chosen for appendage attachment to the fuel elements. The paper describes experiences in scaling up of the technique to industrial production of PHWR fuel bundles, design and development of special-purpose equipment for this purpose, and the QA procedures employed for regular production. It also deals with appendage welding of 37 Element fuel bundles and improvements planned in the appendage welding process. (author)

  16. Fuel element clusters for nuclear reactors

    International Nuclear Information System (INIS)

    Anthony, A.J.; Hutchinson, J.J.

    1975-01-01

    In the fuel element assembly for nuclear reactors the influence of temperature cycles upon the stability of the joints between the individual components, especially between the control rod guide tubes and the connecting rods and end plates, respectively, is reduced. For this purpose, the connection is designed as a bolted connection connecting, on the one hand, the guide tubes and guide bolts and, on the other hand, these two components and the end plates. Moreover, the materials of the guide tubes, bolts and end plates are selected so that their respective thermal expansion coefficients differ. The material which can be used for the end plates and the guide bolts is stainless steel and stainless steel plus inconel (nickel-chrome-iron alloy), respectively; for the guide tubes it is a zirconium alloy (zircaloy). In addition to some technical designs of the bolted connections the materials and lengths of the components are selected in such a way that the expansion path of the components held by a bolted connection is equal to that of the stressing part. (DG/RF) [de

  17. Multi-fuel reformers for fuel cells used in transportation. Phase 1: Multi-fuel reformers

    Science.gov (United States)

    1994-05-01

    DOE has established the goal, through the Fuel Cells in Transportation Program, of fostering the rapid development and commercialization of fuel cells as economic competitors for the internal combustion engine. Central to this goal is a safe feasible means of supplying hydrogen of the required purity to the vehicular fuel cell system. Two basic strategies are being considered: (1) on-board fuel processing whereby alternative fuels such as methanol, ethanol or natural gas stored on the vehicle undergo reformation and subsequent processing to produce hydrogen, and (2) on-board storage of pure hydrogen provided by stationary fuel processing plants. This report analyzes fuel processor technologies, types of fuel and fuel cell options for on-board reformation. As the Phase 1 of a multi-phased program to develop a prototype multi-fuel reformer system for a fuel cell powered vehicle, the objective of this program was to evaluate the feasibility of a multi-fuel reformer concept and to select a reforming technology for further development in the Phase 2 program, with the ultimate goal of integration with a DOE-designated fuel cell and vehicle configuration. The basic reformer processes examined in this study included catalytic steam reforming (SR), non-catalytic partial oxidation (POX) and catalytic partial oxidation (also known as Autothermal Reforming, or ATR). Fuels under consideration in this study included methanol, ethanol, and natural gas. A systematic evaluation of reforming technologies, fuels, and transportation fuel cell applications was conducted for the purpose of selecting a suitable multi-fuel processor for further development and demonstration in a transportation application.

  18. Safety analysis report for packaging: the ORNL HFIR spent-fuel-element shipping cask

    International Nuclear Information System (INIS)

    Evans, J.H.; Chipley, K.K.; Eversole, R.E.; Just, R.A.; Llewellyn, G.H.

    1977-11-01

    The Oak Ridge National Laboratory High Flux Isotope Reactor (HFIR) spent-fuel-element shipping cask is used to transport HFIR, Oak Ridge Research Reactor (ORR), and other reactor fuel elements. The cask was analytically evaluated to determine its compliance with the applicable regulations governing containers in which radioactive materials are transported. Computational procedures and tests were used to determine behavior of the cask relative to the general standards for the hypothetical accident conditions. The results of the evaluation show that the cask is in compliance with the applicable regulations

  19. The transportation of PuO2 and MOX fuel and management of irradiated MOX fuel

    International Nuclear Information System (INIS)

    Dyck, H.P.; Rawl, R.; Durpel, L. van den

    2000-01-01

    Information is given on the transportation of PuO 2 and mixed-oxide (MOX) fuel, the regulatory requirements for transportation, the packages used and the security provisions for transports. The experience with and management of irradiated MOX fuel and the reprocessing of MOX fuel are described. Information on the amount of MOX fuel irradiated is provided. (author)

  20. Device for a nuclear reactor. [Fuel element spacers

    Energy Technology Data Exchange (ETDEWEB)

    Foulds, R B; Kasberg, A H; Puechl, K H; Bleiberg, M L

    1972-03-08

    A spacer design for fuel element clusters for PWR type reactors is described. It consists of a frame supporting an egg-carton like grid each sector of which is provided with springs which grip the fuel pins. The spring design is such as to prevent fuel pin vibrations and at same time accommodate fuel pin deformations. Formulae for the calculation of natural frequencies, spring stiffness and friction loads are presented.

  1. CARA, new concept of advanced fuel element for HWR

    International Nuclear Information System (INIS)

    Florido, P.C.; Crimello, R.O.; Bergallo, J.E.; Marino, A.C.; Delmastro, D.F.; Brasnarof, D.O.; Gonzalez, J.H.

    1999-01-01

    All Argentinean NPPs (2 in operation, 1 under construction), use heavy water as coolant and moderator. With very different reactor concepts (pressure Vessel and CANDU type designs), the fuel elements are completely different in its concepts too. Argentina produces both types of fuel elements at a manufacturing fuel element company, called CONUAR. The very different fuel element's designs produce a very complex economical behavior in this company, due to the low production scale. The competitiveness of the Argentinean electric system (Argentina has a market driven electric system) put another push towards to increase the economical competitiveness of the nuclear fuel cycle. At present, Argentina has a very active Slightly Enriched Uranium (SEU) Program for the pressure vessel HWR type, but without strong changes in the fuel concept itself. Then, the Atomic Energy Commission in Argentina (CNEA) has developed a new concept of fuel element, named CARA, trying to achieve very ambitious goals, and substantially improved the competitiveness of the nuclear option. The ambitious targets for CARA fuel element are compatibility (a single fuel element for all Argentinean's HWR) using a single diameter fuel rod, improve the security margins, increase the burnup and do not exceed the CANDU fabrication costs. In this paper, the CARA concept will be presented, in order to explained how to achieve all together these goals. The design attracted the interest of the nuclear power operator utility (NASA), and the fuel manufacturing company (CONUAR). Then a new Project is right now under planning with the cooperation of three parts (CNEA - NASA - CONUAR) in order to complete the whole development program in the shortest time, finishing in the commercial production of CARA fuel bundle. At the end of the this paper, future CARA development program will be described. (author)

  2. Studies of Lanthanide Transport in Metallic Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Jinsuo; Taylor, Christopher

    2018-04-02

    Metallic nuclear fuels were tested in fast reactor programs and performed well. However, metallic fuels have shown the phenomenon of FCCI that are due to deleterious reactions between lanthanide fission products and cladding material. As the burnup is increased, lanthanide fission products that contact with the cladding could react with cladding constituents such as iron and chrome. These reactions produce higher-melting intermetallic compounds and low-melting alloys, and weaken the mechanical integrity. The lanthanide interaction with clad in metallic fuels is recognized as a long-term, high-burnup cause of the clad failures. Therefore, one of the key concerns of using metallic fuels is the redistribution of lanthanide fission products and migration to the fuel surface. It is believed that lanthanide migration is in part due to the thermal gradient between the center and the fuel-cladding interface, but also largely in part due to the low solubility of lanthanides within the uranium-based metal fuel. PIE of EBR-II fuels shows that lanthanides precipitate directly and do not dissolve to an appreciable extent in the fuel matrix. Based on the PIE data from EBR-II, a recent study recommended a so-called “liquid-like” transport mechanism for lanthanides and certain other species. The liquid-like transport model readily accounts for redistribution of Ln, noble metal fission products, and cladding components in the fuel matrix. According to the novel mechanism, fission products can transport as solutes in liquid metals, such as liquid cesium or liquid cesium–sodium, and on pore surfaces and fracture surfaces for metals near their melting temperatures. Transport in such solutions is expected to be much more rapid than solid-state diffusion. The mechanism could explain the Ln migration to the fuel slug peripheral surface and their deposition with a sludge-like form. Lanthanides have high solubility in liquid cesium but have low solubility in liquid sodium. As a

  3. Statistical estimation of fast-reactor fuel-element lifetime

    International Nuclear Information System (INIS)

    Proshkin, A.A.; Likhachev, Yu.I.; Tuzov, A.N.; Zabud'ko, L.M.

    1980-01-01

    On the basis of a statistical analysis, the main parameters having a significant influence on the theoretical determination of fuel-element lifetimes in the operation of power fast reactors in steady power conditions are isolated. These include the creep and swelling of the fuel and shell materials, prolonged-plasticity lag, shell-material corrosion, gap contact conductivity, and the strain diagrams of the shell and fuel materials obtained for irradiated materials at the corresponding strain rates. By means of deeper investigation of these properties of the materials, it is possible to increase significantly the reliability of fuel-element lifetime predictions in designing fast reactors and to optimize the structure of fuel elements more correctly. The results of such calculations must obviously be taken into account in the cost-benefit analysis of projected new reactors and in choosing the optimal fuel burnup. 9 refs

  4. Model studying the processes arising during fuel element overheating

    International Nuclear Information System (INIS)

    Usynin, G.B.; Anoshkin, Yu.I.; Vlasichev, G.N.; Galitskikh, Yu.N.; Semenychev, M.A.

    1986-01-01

    A calculational technique for studying heating and melting of fuel elements in the BN type reactors during an accident with heat release failure and a simulator with central rod heater intended for out-of-pile experiments is developed. The time rangeof the characteristic melting steps for the most thermally stressed fuel element at the reactor nominal power is calculated. The experimental study of the fuel element melting using a simulator with a tungsten heater has proved that the technique for the simultor and fuel can melting, respectively, is correct. The developed technique is used for determining the geometrical values and operational conditions for experiments with simulators, when quantitative and qualitative characteristics of the process under study are rather close to those natural for fuel elements

  5. Elements of nuclear reactor fueling theory

    International Nuclear Information System (INIS)

    Egan, M.R.

    1984-01-01

    Starting with a review of the simple batch size effect, a more general theory of nuclear fueling is derived to describe the behavior and physical requirements of operating cycle sequences and fueling strategies having practical use in the management of nuclear fuel. The generalized theory, based on linear reactivity modeling, is analytical and represents the effects of multiple-stream, multiple-depletion-batch fueling configurations in systems employing arbitrary, non-integer batch size strategies, and containing fuel with variable energy generation rates. Reactor operating cycles and cycle sequences are represented with realistic structure that includes the effects of variable cycle energy production, cycle lengths, end-of-cycle operating extensions and maneuvering allowances. Results of the analytical theory are first applied to the special case of degenerate equilibrium cycle sequences, yielding several fundamental principles related to the selection of refueling strategy, and which govern fueling decisions normally made by the fuel manager. It is also demonstrated in this application that the simple batch size effect is not valid for non-integer fueling strategies, even in the simplest sequence configurations, and that it systematically underestimates the fueling requirements of degenerate sequences in general

  6. Operational requirements of spherical HTR fuel elements and their performance

    International Nuclear Information System (INIS)

    Roellig, K.; Theymann, W.

    1985-01-01

    The German development of spherical fuel elements with coated fuel particles led to a product design which fulfils the operational requirements for all HTR applications with mean gas exit temperatures from 700 deg C (electricity and steam generation) up to 950 deg C (supply of nuclear process heat). In spite of this relatively wide span for a parameter with strong impact on fuel element behaviour, almost identical fuel specifications can be used for the different reactor purposes. For pebble bed reactors with relatively low gas exit temperatures of 700 deg C, the ample design margins of the fuel elements offer the possibility to enlarge the scope of their in-service duties and, simultaneously, to improve fuel cycle economics. This is demonstrated for the HTR-500, an electricity and steam generating 500 MWel eq plant presently proposed as follow-up project to the THTR-300. Due to the low operating temperatures of the HTR-500 core, the fuel can be concentrated in about 70% of the pebbles of the core thus saving fuel cycle costs. Under all design accident conditions fuel temperatures are maintained below 1250 deg C. This allows a significant reduction in the engineered activity barriers outside the primary circuit, in particular for the loss of coolant accident. Furthermore, access to major primary circuit components and the reuse of the fuel elements after any design accident are possible. (author)

  7. The advanced neutron source three-element-core fuel grading

    International Nuclear Information System (INIS)

    Gehin, J.C.

    1995-01-01

    The proposed Advanced Neutron Source (ANS) pre-conceptual design consists of a two-element 330 MW f nuclear reactor fueled with highly-enriched uranium and is cooled, moderated, and reflected with heavy water. Recently, the ANS design has been changed to a three-element configuration in order to permit a reduction of the enrichment, if required, while maintaining or improving the thermal-hydraulic margins. The core consists of three annular fuel elements composed of involute-shaped fuel plates. Each fuel plate has a thickness of 1.27 mm and consists of a fuel meat region Of U 3 Si 2 -Al (50% enriched in one case that was proposed) and an aluminum filler region between aluminum cladding. The individual plates are separated by a 1.27 mm coolant channel. The three element core has a fuel loading of 31 kg of 235 U which is sufficient for a 17-day fuel cycle. The goal in obtaining a new fuel grading is to maximize important temperature margins. The limits imposed axe: (1) Limit the temperature drop over the cladding oxide layer to less than 119 degrees C to avoid oxide spallation. (2) Limit the fuel centerline temperature to less than 400 degrees C to avoid fuel damage. (3) Limit the cladding wall temperature to less than the coolant. incipient-boiling temperature to avoid coolant boiling. Other thermal hydraulic conditions, such as critical heat flux, are also considered

  8. Hot fuel examination facility element spacer wire-wrap machine

    International Nuclear Information System (INIS)

    Tobias, D.A.; Sherman, E.K.

    1989-01-01

    Nondestructive examinations of irradiated experimental fuel elements conducted in the Argonne National Laboratory Hot Fuel Examination Facility/North (HFEF/N) at the Idaho National Engineering Laboratory include laser and contact profilometry (element diameter measurements), electrical eddy-current testing for cladding and thermal bond defects, bow and length measurements, neutron radiography, gamma scanning, remote visual exam, and photography. Profilometry was previously restricted to spiral profilometry of the element to prevent interference with the element spacer wire wrapped in a helix about the Experimental Breeder Reactor II (EBR-II)-type fuel element from end to end. By removing the spacer wire prior to conducting profilometry examination, axial profilometry techniques may be used, which are considerably faster than spiral techniques and often result in data acquisition more important to experiment sponsors. Because the element must often be reinserted into the nuclear reactor (EBR-II) for additional irradiation, however, the spacer wire must be reinstalled on the highly irradiated fuel element by remote means after profilometry of the wireless elements. The element spacer wire-wrap machine developed at HFEF is capable of helically wrapping fuel elements with diameters up to 1.68 cm (0.660 in.) and 2.44-m (96-in.) lengths. The machine can accommodate almost any desired wire pitch length by simply inserting a new wrapper gear module

  9. Nuclear criticality assessment of Oak Ridge research fuel element storage

    International Nuclear Information System (INIS)

    Thomas, J.T.

    1978-06-01

    Spent and partially spent Oak Ridge Research Reactor (ORR) fuel elements are retained in the storage section of the ORR pool facility. Determination of a maximum expected neutron multiplication factor for the storage area is accomplished by a validated calculational method. The KENO Monte Carlo code and the Hansen-Roach 16-group neutron cross section sets were validated by calculations of critical experiments performed with early ORR fuel elements and with SPERT-D fuel elements. Calculations of various fuel element arrangements are presented which confirm the subcriticality previously inferred from critical experiments and indicate the k/sub eff/ would not exceed 0.85, were the storage area to be filled to capacity with storage racks containing elements with the fissionable material loading increased to 350 g of 235 U

  10. Experimental Study of Elements Promoting Mixing in Fuel Elements

    International Nuclear Information System (INIS)

    Silin, Nicolas; Juanico, Luis; Delmastro, Dario

    2003-01-01

    In the present work a thermal tracing technique is used to measure the increase of the mixing between subchannels in the presence of different mixing elements.As representative elements a spacer, a spacer with mixing vanes and turbulence promoter buttons were considered.The performance of these elements was evaluated by studying the behavior of a thermal trace in each case.Also the pressure drop for each case is presented.The results present a qualitative and quantitative guide for the application of each one of these appendages in future nuclear elements

  11. L. Transportation of fuel and wastes

    International Nuclear Information System (INIS)

    1976-01-01

    The principles applied to the transport of nuclear fuels and wastes have been founded on the more general provisions governing the transport of radioactive materials. Safe shipment of radioactive materials has historically been sought by specifying required characteristics in the shipping packages and establishing minimum acceptable levels of package integrity. The reason for this is that in the course of transport by road, rail, sea, or air, consignments of radioactive material are in close proximity to members of the public, and in many cases they are loaded or unloaded by transport workers who have had no special training or experience in the handling of such substances. The procedures adopted to ensure transport safety have worked satisfactorily. Both in the USA and the UK, the industry and regulatory authorities have established outstanding safety records in shipping radioactive materials over a period of thirty years. It is claimed that there have been no injuries due to the radioactive nature of the shipments, nor has there been a release of nuclear materials serious enough to be a threat of death or injury. Admittedly, about 95% of the 800,000 shipments estimated in the USA each year involve small quantities for use in industry, medicine, agriculture and education. However the principals underlying the safe packaging of these and reactor fuels are the same, and there is little reason to doubt that a similar safety record can be maintained

  12. Fuel transporting device in nuclear reactor

    International Nuclear Information System (INIS)

    Inoue, Tatsumi.

    1975-01-01

    Object: To obtain a support structure of an excellent quakeproof property for a fuel transporting device provided for the transportation of fuel between a reactor building and an auxiliary building in a pressure tube reactor or the like. Structure: The structure comprises an oblique transfer chute loosely penetrating the reactor building, reactor container and auxiliary building, a transfer chute support outer cylinder surrounding the transfer chute and having one end coupled to the transfer chute and other end coupled to the container, flexible seal members respectively provided on the reactor building side and on the auxiliary building side and surrounding the transfer chute and a slidable support supported on the side of the auxiliary building such that it can be in frictional contact with the outer periphery of the transfer chute. With this construction, the relative displacements of various parts caused by an earthquake or the like can be absorbed by the support outer cylinder, flexible seals and slidable support. (Ikeda, J.)

  13. Nuclear fuel element, and method of producing same

    International Nuclear Information System (INIS)

    Armijo, J.S.; Esch, E.L.

    1986-01-01

    This invention relates to an improvement in nuclear fuel elements having a composite container comprising a cladding sheath provided with a protective barrier of zirconium metal covering the inner surface of the sheath, rendering such fuel elements more resistant to hydrogen accumulation in service. The invention specifically comprises removing substantially all zirconium metal of the barrier layer from the part of the sheath surrounding and defining the plenum region. Thus the protective barrier of zirconium metal covers only the inner surface of the fuel container in the area immediately embracing the fissionable fuel material

  14. ELOCA: fuel element behaviour during high temperature transients

    International Nuclear Information System (INIS)

    Sills, H.E.

    1979-03-01

    The ELOCA computer code was developed to simulate the uniform thermal-mechanical behaviour of a fuel element during high-temperature transients such as a loss-of-coolant accident (LOCA). Primary emphasis is on the diametral expansion of the fuel sheath. The model assumed is a single UO2/zircaloy-clad element with axisymmetric properties. Physical effects considered by the code are fuel expansion, cracking and melting; variation, during the transient, of internal gas pressure; changing fuel/sheath heat transfer; thermal, elastic and plastic sheath deformation (anisotropic); Zr/H 2 O chemical reaction effects; and beryllium-assisted crack penetration of the sheath. (author)

  15. WELWING, Material Buckling for HWR with Annular Fuel Elements

    International Nuclear Information System (INIS)

    Grosskopf, O.G.P.

    1973-01-01

    1 - Nature of the physical problem solved: WELWING was developed to calculate the material buckling of reactor systems consisting of annular fuel elements in heavy water as moderator for various moderator to fuel ratios. The moderator to fuel ratio for the maximum material buckling for the particular system is selected automatically and the corresponding material buckling is calculated. 2 - Method of solution: The method used is an analytical solution of the one-group diffusion equations with various corrections and approximations. 3 - Restrictions on the complexity of the problem: Up to 32 different materials in the fuel element may be used

  16. Method of dismantling nuclear fuel elements

    International Nuclear Information System (INIS)

    Adams, G.J.

    1983-01-01

    Nuclear fuel assemblies of the kind comprising fuel pins in dimpled cellular grids are freed from the grids to aid dismantling of the assemblies by causing a rotary sleeve to pass concentrically over the pins to remove the dimples in the grids and thereby increase the freedom of the pins in the cells of the grids. (author)

  17. Design of experiments for test of fuel element reliability

    International Nuclear Information System (INIS)

    Boehmert, J.; Juettner, C.; Linek, J.

    1989-01-01

    Changes of fuel element design and modifications of the operational conditions have to be tested in experiments and pilot projects for nuclear safety. Experimental design is an useful statistical method minimizing costs and risks for this procedure. The main problem of our work was to investigate the connection between failure rate of fuel elements, sample size, confidence interval, and error probability. Using the statistic model of the binomial distribution appropriate relations were derived and discussed. A stepwise procedure based on a modified sequential analysis according to Wald was developed as a strategy of introduction for modifications of the fuel element design and of the operational conditions. (author)

  18. Method of removing crud deposited on fuel element clusters

    International Nuclear Information System (INIS)

    Yokota, Tokunobu; Yashima, Akira; Tajima, Jun-ichiro.

    1982-01-01

    Purpose: To enable easy elimination of claddings deposited on the surface of fuel element. Method: An operator manipulates a pole from above a platform, engages the longitudinal flange of the cover to the opening at the upper end of a channel box and starts up a suction pump. The suction amount of the pump is set such that water flow becomes within the channel box at greater flow rate than the operational flow rate in the channel box of the fuel element clusters during reactor operation. This enables to remove crud deposited on the surface of individual fuel elements with ease and rapidly without detaching the channel box. (Moriyama, K.)

  19. Sodium removal of fuel elements by vacuum distillation

    International Nuclear Information System (INIS)

    Buescher, E.; Haubold, W.; Jansing, W.; Kirchner, G.

    1978-01-01

    Cleaning of sodium-wetted core components can be performed by using either lead, moist nitrogen, or alcohol. The advantages of these methods for cleaning fuel elements without causing damage are well known. The disadvantage is that large amounts of radioactive liquids are formed during handling in the latter two cases. In this paper a new method to clean components is described. The main idea is to remove all liquid metal from the core components within a comparatively short period of time. Fuel elements removed from the reactor must be cooled because of high decay heat release. To date, vacuum distillation of fuel elements has not yet been applied

  20. Environmental economics of lignin derived transport fuels.

    Science.gov (United States)

    Obydenkova, Svetlana V; Kouris, Panos D; Hensen, Emiel J M; Heeres, Hero J; Boot, Michael D

    2017-11-01

    This paper explores the environmental and economic aspects of fast pyrolytic conversion of lignin, obtained from 2G ethanol plants, to transport fuels for both the marine and automotive markets. Various scenarios are explored, pertaining to aggregation of lignin from several sites, alternative energy carries to replace lignin, transport modalities, and allocation methodology. The results highlight two critical factors that ultimately determine the economic and/or environmental fuel viability. The first factor, the logistics scheme, exhibited the disadvantage of the centralized approach, owing to prohibitively expensive transportation costs of the low energy-dense lignin. Life cycle analysis (LCA) displayed the second critical factor related to alternative energy carrier selection. Natural gas (NG) chosen over additional biomass boosts well-to-wheel greenhouse gas emissions (WTW GHG) to a level incompatible with the reduction targets set by the U.S. renewable fuel standard (RFS). Adversely, the process' economics revealed higher profits vs. fossil energy carrier. Copyright © 2017 The Author(s). Published by Elsevier Ltd.. All rights reserved.

  1. Spent-fuel transportation - a success story

    International Nuclear Information System (INIS)

    Gertz, C.P.; Schoonen, D.H.; Wakeman, B.H.

    1986-01-01

    Spent nuclear fuel research and development (R and D) demonstrations and associated transportation activities are being performed as a part of the storage cask performance testing programs at the Idaho National Engineering Laboratory (INEL). These spent-fuel programs support the Nuclear Waste Policy Act (NWPA) and US Department of Energy (DOE) objectives for cooperative demonstrations with the utilities, testing at federal sites, and alternatives for viable transportation systems. A cooperative demonstration program with the private sector to develop dry storage technologies that the US Nuclear Regulatory Commission (NRC) can generically approve is in place as well as cost-shared dry storage R and D program at a federal facility to collect the necessary licensing data. In addition to the accomplishments in the cask performance and testing demonstrations, the long-distance transportation of a large number of spent-fuel assemblies is considered a success story. The evaluation and implementation of applicable requirements, industry perspective, and extensive planning all contributed to this achievement

  2. The nuclear fuel cycle: (2) fuel element manufacture

    International Nuclear Information System (INIS)

    Doran, J.

    1976-01-01

    Large-scale production of nuclear fuel in the United Kingdom is carried out at Springfields Works of British Nuclear Fuels Ltd., a company formed from the United Kingdom Atomic Energy Authority in 1971. The paper describes in some detail the Springfields Works processes for the conversion of uranium ore concentrate to uranium tetrafluoride, then conversion of the tetrafluoride to either uranium metal for cladding in Magnox to form fuel for the British Mk I gas-cooled reactors, or to uranium hexafluoride for enrichment of the fissile 235 U isotope content at the Capenhurst Works of BNFL. Details are given of the reconversion at Springfields Works of this enriched uranium hexafluoride to uranium dioxide, which is pelleted and then clad in either stainless steel or zircaloy containers to form the fuel assemblies for the British Mk II AGR or advanced gas-cooled reactors or for the water reactor fuels. (author)

  3. Thermally-induced bowing of CANDU fuel elements

    International Nuclear Information System (INIS)

    Suk, H.C.; Sim, K.S.; Park, J.H.; Park, G.S.

    1995-01-01

    Considering only the thermally-induced bending moments which are generated both within the sheath and between the fuel and sheath by an asymmetric temperature distribution with respect to the axis of an element, a generalized and explicit analytical formula for the thermally-induced bending is developed in this paper, based on the cases of 1) the bending of an empty tube treated by neglecting of the fuel/sheath mechanical interaction and 2) the fuel/sheath interaction due to the pellet and sheath temperature variations. In each of the cases, the temperature asymmetries in sheath are modelled to be caused by the combined effects of (i) non-uniform coolant temperature due to imperfect coolant mixing, (ii) variable sheath/coolant heat transfer coefficient, (iii) asymmetric heat generation due to neutron flux gradients across an element and so as to inclusively cover the uniform temperature distributions within the fuel and sheath with respect to the axial centerline. Investigating the relative importance of the various parameters affecting fuel element bowing, the element bowing is found to be greatly affected with the variations of element length, sheath diameter, pellet/sheath mechanical interaction and neutron flux depression factors, pellet thermal expansion coefficient, pellet/sheath heat transfer coefficient in comparison with those of other parameters such as sheath thickness, film heat transfer coefficient, sheath thermal expansion coefficient, and sheath and pellet thermal conductivities. Also, the element bowing of the standard 37-element bundle and CANFLEX 43-element bundle for the use in CANDU-6 reactors was analyzed with the formula, which could help to demonstrate the integrity of the fuel. All the required input data for the analyses were generated in terms of the reactor operation conditions on the reactor physics, thermal hydraulics and fuel performance by using various CANDU computer codes. The analysis results indicate that the CANFLEX 43-element

  4. A thermodynamic assessment of the behavior of cesium and rubidium in reactor fuel elements

    International Nuclear Information System (INIS)

    Kohli, R.

    1981-01-01

    A comprehensive thermodynamic model is developed to assess the reaction and transport behavior of fission products in LWR fuel elements. The emphasis is on the chemistry of cesium and rubidium and their reactions with the fuel, other fission products, and the zircaloy cladding. Equilibrium thermodynamic calculations have been performed on the most plausible reactions to predict the chemical state of the fission products. The relevance of the predictions to pellet-clad interaction failures is discussed in detail. (orig.)

  5. Repurposing an irradiated instrumented TRIGA fuel element for regular use

    Energy Technology Data Exchange (ETDEWEB)

    Oliveira, Paulo F.; Souza, Luiz C.A., E-mail: pfo@cdtn.br, E-mail: lcas@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2015-07-01

    TRIGA IPR-R1 is a research reactor also used for training and radioisotope production, located at the Centro de Desenvolvimento da Tecnologia Nuclear da Comissao Nacional de Energia Nuclear (Nuclear Technology Development Centre, Brazilian National Nuclear Energy Commission - CDTN/CNEN). Its first criticality occurred in November 1960. All original fuel elements were aluminum-clad. In 1971 nine new fuel elements, stainless steel-clad were acquired. One of them was an instrumented fuel element (IFE), equipped with 3 thermocouples. The IFE was introduced into the core only on August 2004, and remained there until July 2007. It was removed from the core after the severing of contacts between the thermocouples and their extension cables. After an unsuccessful attempt to recover electrical access to the thermocouples the IFE was transferred from the reactor pool to an auxiliary spent fuel storage well, with water, in the reactor room. In December 2011 the IFE was transferred to an identical well, dry, where it remains so far. This work is a proposal for recovery of this instrumented fuel element, by removing the cable guide rod and adaptation of a superior terminal plug similar to conventional fuel elements. This will enable its handling through the same tool used for regular fuel elements and its return to the reactor core. This is a delicate intervention in terms of radiological protection, and will require special care to minimize the exposure of operators. (author)

  6. Nuclear Energy and Synthetic Liquid Transportation Fuels

    Science.gov (United States)

    McDonald, Richard

    2012-10-01

    This talk will propose a plan to combine nuclear reactors with the Fischer-Tropsch (F-T) process to produce synthetic carbon-neutral liquid transportation fuels from sea water. These fuels can be formed from the hydrogen and carbon dioxide in sea water and will burn to water and carbon dioxide in a cycle powered by nuclear reactors. The F-T process was developed nearly 100 years ago as a method of synthesizing liquid fuels from coal. This process presently provides commercial liquid fuels in South Africa, Malaysia, and Qatar, mainly using natural gas as a feedstock. Nuclear energy can be used to separate water into hydrogen and oxygen as well as to extract carbon dioxide from sea water using ion exchange technology. The carbon dioxide and hydrogen react to form synthesis gas, the mixture needed at the beginning of the F-T process. Following further refining, the products, typically diesel and Jet-A, can use existing infrastructure and can power conventional engines with little or no modification. We can then use these carbon-neutral liquid fuels conveniently long into the future with few adverse environmental impacts.

  7. MTR fuel element supply by CERCA through CECCN after the production transfer from NUKEM

    International Nuclear Information System (INIS)

    Hassel, H.W.

    1991-01-01

    The transfer of fuel element supply contracts, the corresponding Al-materials, structure parts, documents, uranium metal, customers related know-how, tools and equipment from NUKEM to CERCA has been completed, thus now giving a high flexibility for CERCA's workshop to fabricate and inspect large quantities of several types of fuel elements simultaneously. Based on this fact, on strategic planning for the next couple of years and on the fact that after 10 years of RERTR program the necessary high density fuel has been successfully developed and implemented, 'business as usual' in the field of fabrication has well become possible. The RERTR community should now use the great chance to concentrate all its efforts on problems which still strongly influence the fabrication and the use of MTR fuel elements: supply of enriched uranium,reprocessing capabilities and politics, transports of nuclear materials. (author)

  8. Burnable poison fuel element and its fabrication

    International Nuclear Information System (INIS)

    Zukeran, Atsushi; Inoue, Kotaro; Aizawa, Hiroko.

    1985-01-01

    Purpose: To enable to optionally vary the excess reactivity and fuel reactivity. Method: Burnable poisons with a large neutron absorption cross section are contained in fuel material, by which the excess reactivity at the initial stage in the reactor is suppressed by the burnable poisons and the excess reactivity is released due to the reduction in the atomic number density of the burnable poisons accompanying the burning. The burnable poison comprises spherical or rod-like body made of a single material or spherical or rod-like member made of a plurality kind of materials laminated in a layer. These spheres or rods are dispersed in the fuel material. By adequately selecting the shape, combination and the arrangement of the burnable poisons, the axial power distribution of the fuel rods are flattened. (Moriyama, K.)

  9. Sustainable fuel for the transportation sector.

    Science.gov (United States)

    Agrawal, Rakesh; Singh, Navneet R; Ribeiro, Fabio H; Delgass, W Nicholas

    2007-03-20

    A hybrid hydrogen-carbon (H(2)CAR) process for the production of liquid hydrocarbon fuels is proposed wherein biomass is the carbon source and hydrogen is supplied from carbon-free energy. To implement this concept, a process has been designed to co-feed a biomass gasifier with H(2) and CO(2) recycled from the H(2)-CO to liquid conversion reactor. Modeling of this biomass to liquids process has identified several major advantages of the H(2)CAR process. (i) The land area needed to grow the biomass is transportation sector. (ii) Whereas the literature estimates known processes to be able to produce approximately 30% of the United States transportation fuel from the annual biomass of 1.366 billion tons, the H(2)CAR process shows the potential to supply the entire United States transportation sector from that quantity of biomass. (iii) The synthesized liquid provides H(2) storage in an open loop system. (iv) Reduction to practice of the H(2)CAR route has the potential to provide the transportation sector for the foreseeable future, using the existing infrastructure. The rationale of using H(2) in the H(2)CAR process is explained by the significantly higher annualized average solar energy conversion efficiency for hydrogen generation versus that for biomass growth. For coal to liquids, the advantage of H(2)CAR is that there is no additional CO(2) release to the atmosphere due to the replacement of petroleum with coal, thus eliminating the need to sequester CO(2).

  10. Thermomechanical analysis of nuclear fuel elements

    International Nuclear Information System (INIS)

    Hernandez L, H.

    1997-01-01

    This work presents development of a code to obtain the thermomechanical analysis of fuel rods in the fuel assemblies inserted in the core of BWR reactors. The code uses experimental correlations developed in several laboratories. The development of the code is divided in two parts: a) the thermal part and b) the mechanical part, extending both the fuel and the cladding materials. The thermal part consists of finding the radial distribution of temperatures in the pellet, from the fuel centerline up to the coolant, along the total active length, considering one and two phase flow in the coolant, as a result of the pressure drop in the system. The mechanical part analyzes the effects of temperature gradients, pressure and irradiation, to which the fuel rod is subjected. The strains produced by swelling, creep and thermal stress in the fuel material are analyzed. In the same way the strains in the cladding are analyzed, considering the effects produced by the pressure exerted on the cladding by pellet swelling, by the pressure caused by fission gas release toward the cavities, and by the strain produced on the cladding by the pressure changes of the system. (Author)

  11. Examination of irradiated fuel elements using gamma scanning technique

    International Nuclear Information System (INIS)

    Ichim, O.; Mincu, M.; Man, I.; Stanica, M.

    2016-01-01

    The purpose of this paper is to validate the gamma scanning technique used to calculate the activity of gamma fission products from CANDU/TRIGA irradiated fuel elements. After a short presentation of the equipments used and their characteristics, the paper describes the calibration technique for the devices and how computed tomography reconstruction is done. Following the previously mentioned steps is possible to obtain the axial and radial profiles and the computed tomography reconstruction for calibration sources and for the irradiated fuel elements. The results are used to validate the gamma scanning techniques as a non-destructive examination method. The gamma scanning techniques will be used to: identify the fission products in the irradiated CANDU/TRIGA fuel elements, construct the axial and radial distributions of fission products, get the distribution in cross section through computed tomography reconstruction, and determine the nuclei number and the fission products activity of the irradiated CANDU/TRIGA fuel elements. (authors)

  12. Optimization of FBR fuel element for high burnup

    International Nuclear Information System (INIS)

    Marbach, G.; Millet, P.

    1985-03-01

    After a brief historical background showing evolution of the French fast reactor fuel element from RAPSODIE to PHENIX and SUPER PHENIX we have examined the main points which have permitted to increase irradiation performance of the subassembly

  13. Alternative Fuel Transportation Optimization Tool : Description, Methodology, and Demonstration Scenarios.

    Science.gov (United States)

    2015-09-01

    This report describes an Alternative Fuel Transportation Optimization Tool (AFTOT), developed by the U.S. Department of Transportation (DOT) Volpe National Transportation Systems Center (Volpe) in support of the Federal Aviation Administration (FAA)....

  14. Drying results of K-Basin fuel element 1990 (Run 1)

    International Nuclear Information System (INIS)

    Marschman, S.C.; Abrefah, J.; Klinger, G.S.; Oliver, B.M.; MacFarlan, P.J.; Ritter, G.A.

    1998-06-01

    The water-filled K-Basins in the Hanford 100-Area have been used to store N-Reactor spent nuclear fuel (SNF) since the 1970s. Because some leaks in the basins have been detected and some of the fuel is breached due to handling damage and corrosion, efforts are underway to remove the fuel elements from wet storage. An Integrated Process Strategy (IPS) has been developed to package, dry, transport, and store these metallic uranium fuels in an interim storage facility on the Hanford Site (WHC 1995). Information required to support the development of the drying processes, and the required safety analyses, is being obtained from characterization tests conducted on fuel elements removed from the K-Basins. A series of whole element drying tests (reported in separate documents, see Section 8.0) have been conducted by Pacific Northwest National Laboratory (PNNL) on several intact and damaged fuel elements recovered from both the K-East and K-West Basins. This report documents the results of the first of those tests (Run 1), which was conducted on an N-Reactor inner fuel element (1990) that had been stored underwater in the K-West Basin (see Section 2.0). This fuel element was subjected to a combination of low- and high-temperature vacuum drying treatments that were intended to mimic, wherever possible, the fuel treatment strategies of the IPS. The testing was conducted in the Whole Element Furnace Testing System, described in Section 3.0, located in the Postirradiation Testing Laboratory (PTL, 327 Building). The test conditions and methodology are given in Section 4.0, and the experimental results provided in Section 5.0. These results are further discussed in Section 6.0

  15. Dynamic characterization of the CAREM fuel element prototype

    International Nuclear Information System (INIS)

    Ghiselli, Alberto M.; Fiori, Jose M.; Ibanez, Luis A.

    2004-01-01

    As a previous step to make a complete test plan to evaluate the hydrodynamic behavior of the present configuration of the CAREM type fuel element, a dynamic characterization analysis is required, without the dynamic response induced by the flowing fluid. This paper presents the tests made, the methods and instrumentation used, and the results obtained in order to obtain a complete dynamic characterization of the CAREM type fuel element. (author)

  16. Method for fuel element leak detection in pressurized water reactors

    International Nuclear Information System (INIS)

    Kunze, U.

    1983-01-01

    The method is aimed at detecting fuel element leaks during reactor operation. It is based on neutron flux measurements at many points in the core, using at least two detectors at a time. The detectors must be arranged in the direction of the coolant flow. Values obtained from periodic measurements are compared with threshold values. The location of fuel element leaks is determined from those values exceeding the threshold of individual detectors

  17. A computer program for structural analysis of fuel elements

    International Nuclear Information System (INIS)

    Hayashi, I.M.V.; Perrotta, J.A.

    1988-01-01

    It's presented the code ELCOM for the matrix analysis of tubular structures coupled by rigid spacers, typical of PWR's fuel elements. The code ELCOM makes a static structural analysis, where the displacements and internal forces are obtained for each structure at the joints with the spacers, and also, the natural frequencies and vibrational modes of an equivalent integrated structure are obtained. The ELCOM result is compared to a PWR fuel element structural analysis obtained in published paper. (author) [pt

  18. Determining reactor fuel elements broken by Cerenkov counting

    International Nuclear Information System (INIS)

    Guo Juhao; Dong Shiyuan; Feng Yuying

    1996-01-01

    The basis and method of determining fuel elements broken in a reactor by Cerenkov counting measured with liquid scintillation spectrometer are introduced. The radioactive characteristic of the radiation nuclides generating Cherenkov radiation in the primary water of 200 MW nuclear district heating reactor is analyzed. The activity of the activation products in the primary water and the fission products in the fuel elements are calculated. A feasibility of Cerenkov counting measure was analyzed. This method is simple and quick

  19. Elements of nuclear reactor fueling theory

    International Nuclear Information System (INIS)

    Egan, M.R.

    1984-01-01

    Starting with a review of the simple batch size effect, a more general theory of nuclear fueling is derived to describe the behaviour and physical requirements of operating cycle sequences and fueling strategies having practical use in fuel management. The generalized theory, based on linear reactivity modeling, is analytical and represents the effects of multiple-stream, multiple-depletion-batch fueling configurations in systems employing arbitrary, non-integer batch size strategies, and containing fuel with variable energy generation rates. Reactor operating cycles and cycle sequences are represented with realistic structure that includes the effects of variable cycle energy production, cycle lengths, end-of-cycle operating extensions and manoeuvering allowances. Results of the analytical theory are first applied to the special case of degenerate equilibrium cycle sequences, yielding several fundamental principles related to the selection of refueling strategy. Numerical evaluations of degenerate equilibrium cycle sequences are then performed for a typical PWR core, and accompanying fuel cycle costs are calculated. The impact of design and operational limits as constraints on the performance mappings for this reactor are also studied with respect to achieving improved cost performance from the once-through fuel cycle. The dynamics of transition cycle sequences are then examined using the generalized theory. Proof of the existence of non-degenerate equilibrium cycle sequences is presented when the mechanics of the fixed reload batch size strategy are developed analytically for transition sequences. Finally, an analysis of the fixed reload enrichment strategy demonstrates the potential for convergence of the transition sequence to a fully degenerate equilibrium sequence. (author)

  20. Nuclear fuel element recovery using PEDSCO RMI Unit

    International Nuclear Information System (INIS)

    Martin, D.G.; Pedersen, B.V.

    1984-01-01

    In September 1982, a PEDSCO Remote Mobile Investigation Unit was used to recover damaged irradiated fuel elements from a fueling machine and trolley deck at Bruce Nuclear Generating Station 'A'. This Canadian-made remote controlled vehicle was originally designed for explosive ordinance disposal by law enforcement agencies. This paper describes its adaptation to nuclear service and its first mission, within a nuclear facility

  1. The Calculation Of Total Radioactivity Of Kartini Reactor Fuel Element

    International Nuclear Information System (INIS)

    Budisantoso, Edi Trijono; Sardjono, Y.

    1996-01-01

    The total radioactivity of Kartini reactor fuel element has been calculated by using ORIGEN2. In this case, the total radioactivity is the sum of alpha, beta, and gamma radioactivity from activation products nuclides, actinide nuclides and fission products nuclides in the fuel element. The calculation was based on irradiation history of fuel in the reactor core. The fuel element no 3203 has location history at D, E, and F core zone. The result is expressed in graphics form of total radioactivity and photon radiations as function of irradiation time and decay time. It can be concluded that the Kartini reactor fuel element in zone D, E, and F has total radioactivity range from 10 Curie to 3000 Curie. This range is for radioactivity after decaying for 84 days and that after reactor shut down. This radioactivity is happened in the fuel element for every reactor operation and decayed until the fuel burn up reach 39.31 MWh. The total radioactivity emitted photon at the power of 0.02 Watt until 10 Watt

  2. Determination of a geometry-dependent parameter and development of a calculation model for describing the fission products transport from spherical fuel elements of graphite moderated gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Weissfloch, R

    1973-07-15

    The fuel elements of high-temperature reactors, coated with pyrolitic carbon and covered with graphite, release fission products like all other fuel elements. Because of safety reasons, the rate of this release has to be kept low and has also to be predictable. Measured values from irradiation tests and from post-irradiation tests about the actual release of different fission products are presented. The physical and chemical mechanism, which determines the release, is extraordinarily complex and in particular not clearly defined. Because of the mentioned reasons, a simplified calculation model was developed, which only considers the release-mechanisms phenomenologically. This calculation model coincides very well in its results with values received in experiments until now. It can be held as an interim state on the way to a complete theory.

  3. Determination of a geometry-dependent parameter and development of a calculation model for describing the fission products transport from spherical fuel elements of graphite moderated gas-cooled reactors

    International Nuclear Information System (INIS)

    Weissfloch, R.

    The fuel elements of High-Temperature Reactors, coated with pyrolitic carbon and covered with graphite, release fission products like all other fuel elements. Because of safety reasons the rate of this release has to be kept low and has also to be predictable. Measured values from irradiation tests and from post-irradiation tests about the actual release of different fission products are present. The physical and chemical mechanism, which determines the release, is extraordinarily complex and in particular not clearly defined. Because of the mentioned reasons a simplified calculation model was developed, which only considers the release-mechanisms phenomenologically. This calculation model coincides very well in its results with values received in experiments until now. It can serve as an interim state on the way to a complete theory. (U.S.)

  4. Life cycle analysis of transportation fuel pathways

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2012-02-24

    The purpose of this work is to improve the understanding of the concept of life cycle analysis (LCA) of transportation fuels and some of its pertinent issues among non-technical people, senior managers, and policy makers. This work should provide some guidance to nations considering LCA-based policies and to people who are affected by existing policies or those being developed. While the concept of employing LCA to evaluate fuel options is simple and straightforward, the act of putting the concept into practice is complex and fraught with issues. Policy makers need to understand the limitations inherent in carrying out LCA work for transportation fuel systems. For many systems, even those that have been employed for a 100 years, there is a lack of sound data on the performance of those systems. Comparisons between systems should ideally be made using the same tool, so that differences caused by system boundaries, allocation processes, and temporal issues can be minimized (although probably not eliminated). Comparing the results for fuel pathway 1 from tool A to those of fuel system 2 from tool B introduces significant uncertainty into the results. There is also the question of the scale of system changes. LCA will give more reliable estimates when it is used to examine small changes in transportation fuel pathways than when used to estimate large scale changes that replace current pathways with completely new pathways. Some LCA tools have been developed recently primarily for regulatory purposes. These tools may deviate from ISO principles in order to facilitate simplicity and ease of use. In a regulatory environment, simplicity and ease of use are worthy objectives and in most cases there is nothing inherently wrong with this approach, particularly for assessing relative performance. However, the results of these tools should not be confused with, or compared to, the results that are obtained from a more complex and rigorous ISO compliant LCA. It should be

  5. LEU fuel element produced by the Egyptian fuel manufacturing pilot plant

    International Nuclear Information System (INIS)

    Zidan, W.I.

    2000-01-01

    The Egyptian Fuel Manufacturing Pilot Plant, FMPP, is a Material Testing Reactor type (MTR) fuel element facility, for producing the specified fuel elements required for the Egyptian Second Research Reactor, ETRR-2. The plant uses uranium hexafluoride (UF 6 , 19.75% U 235 by wt) as a raw material which is processed through a series of the manufacturing, inspection and test plan to produce the final specified fuel elements. Radiological safety aspects during design, construction, operation, and all reasonably accepted steps should be taken to prevent or reduce the chance of accidents occurrence. (author)

  6. Detection and location of leaking TRIGA fuel elements

    International Nuclear Information System (INIS)

    Bouchey, G.D.; Gage, S.J.

    1970-01-01

    Several TRIGA facilities have experienced difficulty resulting from cladding failures of aluminum clad TRIGA fuel elements. Recently, at the University of Texas at Austin reactor facility, fission product releases were observed during 250 kW operation and were attributed to a leaking fuel element. A rather extensive testing program has been undertaken to locate the faulty element. The used sniffer device is described, which provides a quick, easily constructed, and extremely sensitive means of locating leaking fuel elements. The difficulty at The University of Texas was compounded by extremely low levels and the sporadic nature of the releases. However, in the more typical situation, in which a faulty element consistently releases relatively large quantities of fission gas, such a device should locate the leak with little difficulty

  7. New, innovative and sustainable transportation fuels

    Energy Technology Data Exchange (ETDEWEB)

    Lassi, U. (Univ. of Oulu, Dept. of Chemistry (Finland)). email: ulla.lassi@oulu.fi; Keiski, R. (Univ. of Oulu, Dept. of Process and Environmental Engineering (Finland)); Kordas, K. (Univ. of Oulu, Microelectronics and Materials Physics Laboratories (Finland)); Mikkola, J.-P. (Aabo Akademi Univ., Lab. of Industrial Chemistry and Reaction Engineering, Turku (Finland))

    2009-07-01

    Secondary products from the industry - e.g. by-products of food and paper/pulp industry - can be used to manufacture new liquid biofuels or fuel components. A particularly interesting alternative is provided by butanol, which can be produced from biomass, since it seems to be most suitable for replacing petrol as fuel in gasoline engines. Besides, it is very energy efficient and also suitable to be produced on an industrial scale. Production of biobutanol and other higher alcohols is studied in the research project 'New, innovative sustainable transportation fuels for mobile applications; from biocomponents to flexible liquid fuels (SusFuFlex)'. The project is carried out as a joint project between the University of Oulu and Aabo Akademi University. It is financied by the Academy of Finland in 2008-2011, within the research programme for Sustainable Energy. Research focuses on the production of higher bioalcohols and other compounds suitable as oxygenates (e.g. butanol, pentanol, mixed alcohols; e.g. glycerine ethers, glycerol carbonate). The objectives of the research are (1) to evaluate the old and novel procedures for microbiological production of butanol, higher alcohols and oxygenates as fossil fuel substitutes, (2) to develop and optimize catalytic materials and chemical reaction routes for the production of higher alcohols and other bio-derived compounds applicable as gasoline fuel and its additives, (3) to conduct a sustainability analysis of the processes to be developed, to analyze the atom economy of the new processes and to make a preliminary economical analysis, and (4) to integrate the processes and know-how developed by the research groups

  8. Element bow profiles from new and irradiated CANDU fuel bundles

    International Nuclear Information System (INIS)

    Dennier, D.; Manzer, A.M.; Ryz, M.A.

    1996-01-01

    Improved methods of measuring element profiles on new CANDU fuel bundles were developed at the Sheridan Park Engineering Laboratory, and have now been applied in the hot cells at Whiteshell Laboratories. For the first time, the outer element profiles have been compared between new, out-reactor tested, and irradiated fuel elements. The comparison shows that irradiated element deformation is similar to that observed on elements in out-reactor tested bundles. In addition to the restraints applied to the element via appendages, the element profile appears to be strongly influenced by gravity and the end loads applied by local deformation of the endplate. Irradiation creep in the direction of gravity also tends to be a dominant factor. (author)

  9. Push piece for spent fuel elements magazine

    International Nuclear Information System (INIS)

    Griveau, R.; Kerlau, D.; Tucoulat, D.; Colas, J.; Pellier, R.

    1989-01-01

    The push piece permits the displacement of little section elements in a magazine of high section. At the end of cut, the push piece leans its flank against an auxiliary blank holder and the element is pushed by a paddle, the push piece being immobilized [fr

  10. European experience in the transport of irradiated light-water reactor fuel

    International Nuclear Information System (INIS)

    Curtis, H.W.

    1979-01-01

    Various methods of transport of irradiated fuel flasks in Europe are described. While many problems in the transport of heavy flasks have been solved some remain and new ones have appeared. Some of these problems are the accumulation of crud on the surface of fuel elements, the problems of failed fuel, stringent criticality criteria, the ''sweating out'' of contaminated flasks, the access, road or rail, to reactor sites, and the maintenance of the transport vehicles. Some future trends in the direction of heavy flasks in the range of 75 to 100 tonnes are indicated

  11. Validation of computer codes and modelling methods for giving proof of nuclear safety of transport and storage of spent VVER-type nuclear fuels. Pt. 2. Criticality safety during transport and storage of spent VVER fuel elements. Final report; Einschaetzung von Rechenprogrammen und Methoden zum Nachweis der nuklearen Sicherheit bei Transport und Lagerung von WWER-Kernbrennstoffen. T. 2. Kritikalitaetssicherheit bei Transport und Lagerung von WWER-Brennelementen. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Buechse, H.; Langowski, A.; Lein, M.; Nagel, R.; Schmidt, H.; Stammel, M.

    1995-03-15

    The report gives the results of investigations on the validation of computer codes used to prove nuclear safety during transport and storage of spent VVER - fuel of NPP Greifswald and Rheinsberg. Characteristics of typical spent fuel (nuclide concentration, neutron source strength, gamma spectrum, decay heat) - calculated with several codes - and dose rates (e.g. in the surrounding of a loaded spent fuel cask) - based on the different source terms - are presented. Differences and their possible reasons are discussed. The results show that despite the differences in the source terms all relevant health physics requirements are met for all cases of source term. The validation of the criticality code OMEGA was established by calculation of appr. 200 critical experiments of LWR fuel, including VVER fuel rod arrangements. The mean error of the effective multiplication factor k{sub eff} is -0,01 compared to the experiment for this area of applicability. Thus, the OMEGA error of 2% assumed in earlier works has turned out to be sufficiently conservative. (orig.) [Deutsch] Der Bericht enthaelt die Ergebnisse von Untersuchungen zur Validierung von Rechenprogrammen, welche zum Nachweis der nukelaren Sicherheit bei Transport und Lagerung von WWER-Kernbrennstoff der KKW Greifswald und Rheinsberg eingesetzt wurden. Es werden eine Charakteristik des abgebrannten Brennstoffs (Nuklidkonzentrationen, Neutronenquellstaerke, Gammaspektrum, Nachzerfallsleistung) - berechnet mit verschiedenen Programmen - und Ortsdosisleistungen (z.B. in der Umgebung eines Transportbehaelters) - basierend auf den verschiedenen Quelltermen - angegeben. Differenzen und Ursachen werden diskutiert. Die Ergebnisse zeigen, dass trotz der Differenzen in den Quelltermen alle strahlenschutztechnisch relevanten Aussagen unbeeinflusst bleiben. Fuer die Einschaetzung des Gueltigkeitsbereiches des Monte-Carlo-Programms OMEGA wurden ca. 200 kritische Experimente mit LWR-Brennstoff unter besonderer Beruecksichtigung

  12. A finite element method for neutron transport

    International Nuclear Information System (INIS)

    Ackroyd, R.T.

    1978-01-01

    A variational treatment of the finite element method for neutron transport is given based on a version of the even-parity Boltzmann equation which does not assume that the differential scattering cross-section has a spherical harmonic expansion. The theory of minimum and maximum principles is based on the Cauchy-Schwartz equality and the properties of a leakage operator G and a removal operator C. For systems with extraneous sources, two maximum and one minimum principles are given in boundary free form, to ease finite element computations. The global error of an approximate variational solution is given, the relationship of one the maximum principles to the method of least squares is shown, and the way in which approximate solutions converge locally to the exact solution is established. A method for constructing local error bounds is given, based on the connection between the variational method and the method of the hypercircle. The source iteration technique and a maximum principle for a system with extraneous sources suggests a functional for a variational principle for a self-sustaining system. The principle gives, as a consequence of the properties of G and C, an upper bound to the lowest eigenvalue. A related functional can be used to determine both upper and lower bounds for the lowest eigenvalue from an inspection of any approximate solution for the lowest eigenfunction. The basis for the finite element is presented in a general form so that two modes of exploitation can be undertaken readily. The model can be in phase space, with positional and directional co-ordinates defining points of the model, or it can be restricted to the positional co-ordinates and an expansion in orthogonal functions used for the directional co-ordinates. Suitable sets of functions are spherical harmonics and Walsh functions. The latter set is appropriate if a discrete direction representation of the angular flux is required. (author)

  13. Results of fuel elements fabrication on the basis of increased concentration dioxide fuel for research reactors

    International Nuclear Information System (INIS)

    Alexandrov, A.B.; Afanasiev, V.L.; Enin, A.A.; Suprun, V.B.

    1996-01-01

    According to the Russian Reduced Enrichment for Research and Test Reactors (RERTR) program, that were constructed under the Russian projects, at the Novosibirsk Chemical Concentrates Plant the pilot series of different configuration (WR-M2, MR, IRT-4M) fuel elements, based on increased concentration uranium dioxide fuel, have been fabricated for reactor tests. Comprehensive fabricated fuel elements quality estimation has been carried out. (author)

  14. Transport of MOX fuel from Europe to Japan

    International Nuclear Information System (INIS)

    2002-01-01

    The MOX fuel transports from Europe to Japan represent a main part in the implementing of the Japan nuclear program. They complement the 160 transports of spent fuels realized from Japan to Europe and the vitrified residues return from France to Japan. In this framework the document presents the MOX fuel, the use of the MOX fuel in reactor, the proliferation risks, the MOX fuel transport to Japan, the public health, the transport regulations, the safety and the civil liability. (A.L.B.)

  15. Ethanol as a Fuel for Road Transportation

    Energy Technology Data Exchange (ETDEWEB)

    Larsen, U.; Johansen, T.; Schramm, J.

    2009-05-15

    Bioethanol as a motor fuel in the transportation sector, mainly for road transportation, has been subject to many studies and much discussion. Furthermore, the topic involves not only the application and engine technical aspects, but also the understanding of the entire life cycle of the fuel, well-to-wheels, including economical, environmental, and social aspects. It is not, however, the aim of this report to assess every single one of these aspects. The present report aims to address the technical potential and problems as well as the central issues related to the general application of bioethanol as an energy carrier in the near future. A suitable place to start studying a fuel is at the production stage, and bioethanol has been found to have a potential to mitigate greenhouse gases, depending on the production method. This and a potential for replacing fossil fuel-based oil (and being renewable) are the main reasons why ethanol is considered and implemented. Therefore, we must focus on two central questions related to ethanol implementation: how much carbon dioxide (CO2) can be mitigated and how much fossil fuel can be replaced? A number of life cycle assessments have been performed in order to provide estimates. These assessments have generally shown that bioethanol has very good potential and can mitigate CO2 emissions very effectively, but It has also been shown that the potential for both fossil fuel replacement and CO2 mitigation is totally dependent on the method used to produce the fuel. Bioethanol can be made from a wide range of biomass resources, not all equally effective at mitigating CO2 emissions and replacing fossil fuel. The Brazilian ethanol experience has in many ways shown the way for the rest of the world, not least in the production stage. Brazil was the first and biggest producer of bioethanol, but the United States, China, India, and European Union have since then increased their production dramatically. Overall, bioethanol represents the

  16. Administrative mechanics of research fuel transportation

    International Nuclear Information System (INIS)

    Harmon, Diane W.

    1983-01-01

    This presentation contains the discussion on the multitude of administrative mechanics that have to be meshed for the successful completion of a shipment of spent fuel, HEU or LEU in the research reactors fuel cycle. The costs associated with transportation may be the equivalent of 'a black hole', so an overview of cost factors is given. At the end one could find that this black hole factor in the budget is actually a bargain. The first step is the quotation phase. The cost variables in the quotation contain the cost of packaging i.e. containers; the complete routing of the packages and the materials. Factors that are of outmost importance are the routing restrictions and regulations, physical security regulations. All of this effort is just to provide a valid quotation not to accomplish the goal of completing a shipment. Public relations cannot be omitted either

  17. Administrative mechanics of research fuel transportation

    Energy Technology Data Exchange (ETDEWEB)

    Harmon, Diane W [Edlow International Company, Washington, DC (United States)

    1983-09-01

    This presentation contains the discussion on the multitude of administrative mechanics that have to be meshed for the successful completion of a shipment of spent fuel, HEU or LEU in the research reactors fuel cycle. The costs associated with transportation may be the equivalent of 'a black hole', so an overview of cost factors is given. At the end one could find that this black hole factor in the budget is actually a bargain. The first step is the quotation phase. The cost variables in the quotation contain the cost of packaging i.e. containers; the complete routing of the packages and the materials. Factors that are of outmost importance are the routing restrictions and regulations, physical security regulations. All of this effort is just to provide a valid quotation not to accomplish the goal of completing a shipment. Public relations cannot be omitted either.

  18. Sipping test on a failed MTR fuel element

    International Nuclear Information System (INIS)

    Terremoto, Luis Antonio Albiac; Zeituni, Carlos Alberto; Silva, Antonio Teixeira e; Perrotta, Jose Augusto; Silva, Jose Eduardo Rosa da

    2002-01-01

    This work describes sipping tests performed on MTR fuel elements of the IEA-R1 research reactor, in order to determinate which one failed in the core during a routine operation of the reactor. radioactive iodine isotopes 131 I and 133 I, employed as failure indicators, were detected in samples corresponding to the fuel element IEA-156. The specific activity of each sample, as well as the average leaking rate, were measured for 137 Cs. The nuclear fuels U 3 O 8 - Al dispersion and U - Al alloy were compared concerning their measured average leaking rates of 137 Cs. (author)

  19. 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

  20. Comparison of the Transportation Risks Resulting from Accidents during the Transportation of the Spent Fuel

    International Nuclear Information System (INIS)

    Jeong Jong Tae; Cho, Dong Kuen; Choi, Heui Joo; Choi, Jong Won

    2007-01-01

    The safe, environmentally sound and publicly acceptable disposal of high level wastes and spent fuels is becoming a very important issue. The operational safety assessment of a repository including a transportation safety assessment is a fundamental part in order to achieve this goal. According to the long term management strategy for spent fuels in Korea, they will be transported from the spent fuel pools in each nuclear power plant to the central interim storage facility (CISF) which is to start operation in 2016. Therefore, we have to determine the safe and economical logistics for the transportation of these spent fuels by considering their transportation risks and costs. In this study, we developed four transportation scenarios by considering the type of transportation casks and transport means in order to suggest safe and economical transportation logistics for spent fuels. Also, we estimated and compared the transportation risks resulting from the accidents during the transportation of spent fuels for these four transportation scenarios

  1. Numerical simulation of ion transport membrane reactors: Oxygen permeation and transport and fuel conversion

    KAUST Repository

    Hong, Jongsup; Kirchen, Patrick; Ghoniem, Ahmed F.

    2012-01-01

    Ion transport membrane (ITM) based reactors have been suggested as a novel technology for several applications including fuel reforming and oxy-fuel combustion, which integrates air separation and fuel conversion while reducing complexity

  2. Fabrication of fuel elements on the basis of increased concentration fuel composition

    International Nuclear Information System (INIS)

    Alexandrov, A.B.; Afanasiev, V.L.; Enin, A.A.; Suprun, V.B.

    2004-01-01

    As a part of Russian Program RERTR Reduced Enrichment for Research and Test Reactors), at NCCP, Inc. jointly with the State Scientific Centre VNIINM the mastering in industrial environment of design and fabrication process of fuel elements (FE) with increased concentration fuel compositions is performed. Fuel elements with fuel composition on the basis of dioxide uranium with nearly 4 g/cm 3 fuel concentration have been produced thus confirming the principal possibility of fuel enrichment reduction down to 20% for research reactors which were built up according to the projects of the former USSR, by increasing the oxide fuel concentration in fuel assemblies (FAs). The form and geometrical dimensions of FEs and FAs shall remain unchanged, only uranium mass in FA shall be increased. (author)

  3. Process for assembling a nuclear fuel element

    International Nuclear Information System (INIS)

    Wachtendonk, H.J. von.

    1984-01-01

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

  4. Handling system for nuclear reactor fuel and reflector elements

    International Nuclear Information System (INIS)

    Hawke, B.C.; Goldman, L.A.

    1980-01-01

    A system for canning, inspecting and transferring to a storage area fuel and reflector elements from a nuclear reactor is described. The canning mechanism operates in a sealed gaseous environment and visual and mechanical inspection of the elements is possible by an operator from a remote shielded area. (UK)

  5. Applications and experience with a new instrumented fuel element

    International Nuclear Information System (INIS)

    Morris, F.M.

    1972-01-01

    Previously reported information to TRIGA Reactor Conference I concerning the development of a new concept in an instrumented fuel element is updated and expanded. The evaluation of these new instrumented elements is discussed and some areas of application to reactor behavior are described. Experiments concerning temperature and flux mapping under varying conditions are investigated and conclusions are given. (author)

  6. End plug welding of nuclear fuel elements-AFFF experience

    International Nuclear Information System (INIS)

    Bhatt, R.B.; Singh, S.; Aniruddha Kumar; Amit; Arun Kumar; Panakkal, J.P.; Kamath, H.S.

    2004-01-01

    Advanced Fuel Fabrication Facility is engaged in the fabrication of mixed oxide (U,Pu)O 2 fuel elements of various types of nuclear reactors. Fabrication of fuel elements involves pellet fabrication, stack making, stack loading and end plug welding. The requirement of helium bonding gas inside the fuel elements necessitates the top end plug welding to be carried out with helium as the shielding gas. The severity of the service conditions inside a nuclear reactor imposes strict quality control criteria, which demands for almost defect free welds. The top end plug welding being the last process step in fuel element fabrication, any rejection at this stage would lead to loss of effort prior to this step. Moreover, the job becomes all the more difficult with mixed oxide (MOX) as the entire fabrication work has to be carried out in glove box trains. In the case of weld rejection, accepted pellets are salvaged by cutting the clad tube. This is a difficult task and recovery of pellets is low (requiring scrap recovery operation) and also leads to active metallic waste generation. This paper discusses the experience gained at AFFF, in the past 12 years in the area of end plug welding for different types of MOX fuel elements

  7. Gap's dimensions in fuel elements from neutron radiography

    International Nuclear Information System (INIS)

    Notea, A.; Segal, Y.; Trichter, F.

    1985-01-01

    Quantitative Nondestructive evaluation (QNDE) is of upmost importance in the design and manufacture of nuclear fuel elements. Accurate non-destructive measurements of gaps, cracks, displacements, etc. supply vital information for optimizing fuel manufacturing. Neutron radiography is a powerful NDT method for examining spent fuel elements. However, it turned out that the extraction of dimensions, especially in the submillimetric range is questionable. In this paper neutron radiography of pellet-to-pellet gaps in fuel elements is modelled and two procedures for dimension extraction are presented. It is shown that for a wide gap the dimension is preferable, extracted from the width of the film profile, while for narrow gaps it is preferable to extract it from the maximum of the density profile

  8. RA-3 core with uranium silicide fuel elements

    International Nuclear Information System (INIS)

    Abbate, Maximo J.; Sbaffoni, Maria M.

    2000-01-01

    Following on with studies on uranium silicide fuel elements, this paper reports some comparisons between the use of standard ECN [U 3 O 8 ] fuel elements and type P-06 [from U 3 Si 2 ] fuel elements in the RA-3 core.The first results showed that the calculated overall mean burn up is in agreement with that reported for the facility, which gives more confidence to the successive ones. Comparing the mentioned cores, the silicide one presents several advantages such as: -) a mean burn up increase of 18 %; -) an extraction burn up increase of 20 %; -) 37.4 % increase in full power days, for mean burn up. All this is meritorious for this fuel. Moreover, grouped and homogenized libraries were prepared for CITVAP code that will be used for planning experiments and other bidimensional studies. Preliminary calculations were also performed. (author)

  9. Quality control in the fuel elements production process

    International Nuclear Information System (INIS)

    Katanic-Popovic, J.; Spasic, Z.; Djuricis, Lj.

    1977-01-01

    Recently great attention has been paid at the international level to the analysis of production processes and quality control of fuel and fuel elements with the aim to speed up activity of proposing and accepting standards and measurement methods. IAEA also devoted great interest to these problems appealing to more active participation of all users and producers fuel elements in a general effort to secure successful work of nuclear plants. For adequate and timely participation in future in the establishment and analysis of general requirements and documentation for the control of purchased or self produced fuel elements in out country it is necessary to be well informed and to follow this activity at the international level. (author)

  10. Fuel Retrieval and Management of Fuel Element Debris

    International Nuclear Information System (INIS)

    Chande, Shridhar; Lachaume, J. L.

    2013-01-01

    Nuclear accidents involving core meltdown have not been so rare. While the first occurred in early fifties, it is reported that about 20 have occurred worldwide in military and commercial reactors. The more recent and major accidents are 1. Three Mile Island, USA in 1979: Approximately half the core was melted, and flowed to the bottom of the reactor pressure vessel however the pressure vessel remained intact and contained the damaged fuel. 2. Chernobyl, former USSR in 1984: Explosive release of radioactive material occurred. About 6 tons of fuel was dispersed as air-borne particles. Most of the core was damaged or melted. 3. Fukushima, Japan 2011: Three units suffered melt down. In unit 1 almost all the fuel assemblies melted and accumulated at the bottom of the vessel. It is reported that the vessel failed and the molten corium has penetrated the concrete. In the units 2 and 3, partial melting of cores has occurred. In several of these cases, fuel retrieval and management activities have been carried out. The experience and insights gained from these activities will be extremely useful for planning and execution of similar activities in future if ever they are needed. The purpose of this session was to exchange this experience and also to share the lessons learned. This is of particularly important, at this juncture, when planning and preparation for retrieval of damaged cores in Fukushima NPP is in progress. (author)

  11. Discrete element modelling of bedload transport

    Science.gov (United States)

    Loyer, A.; Frey, P.

    2011-12-01

    Discrete element modelling (DEM) has been widely used in solid mechanics and in granular physics. In this type of modelling, each individual particle is taken into account and intergranular interactions are modelled with simple laws (e.g. Coulomb friction). Gravity and contact forces permit to solve the dynamical behaviour of the system. DEM is interesting to model configurations and access to parameters not directly available in laboratory experimentation, hence the term "numerical experimentations" sometimes used to describe DEM. DEM was used to model bedload transport experiments performed at the particle scale with spherical glass beads in a steep and narrow flume. Bedload is the larger material that is transported on the bed on stream channels. It has a great geomorphic impact. Physical processes ruling bedload transport and more generally coarse-particle/fluid systems are poorly known, arguably because granular interactions have been somewhat neglected. An existing DEM code (PFC3D) already computing granular interactions was used. We implemented basic hydrodynamic forces to model the fluid interactions (buoyancy, drag, lift). The idea was to use the minimum number of ingredients to match the experimental results. Experiments were performed with one-size and two-size mixtures of coarse spherical glass beads entrained by a shallow turbulent and supercritical water flow down a steep channel with a mobile bed. The particle diameters were 4 and 6mm, the channel width 6.5mm (about the same width as the coarser particles) and the channel inclination was typically 10%. The water flow rate and the particle rate were kept constant at the upstream entrance and adjusted to obtain bedload transport equilibrium. Flows were filmed from the side by a high-speed camera. Using image processing algorithms made it possible to determine the position, velocity and trajectory of both smaller and coarser particles. Modelled and experimental particle velocity and concentration depth

  12. Endplug Welding Techniques developed for SFR Metallic Fuel Elements

    International Nuclear Information System (INIS)

    Lee, Jung Won; Kim, Soo Sung; Woo, Yoon Myeng; Kim, Hyung Tae; Lee, Ho Jin; Kim, Ki Hwan

    2013-01-01

    In Korea, the R and D on SFR has been begun since 1997, as one of the national long-term nuclear R and D programs. The international collaborative research is under way on fuel developments within Advanced Fuel Project for Gen-IV SFR with the closed fuel cycle of full actinide recycling, while TRU bearing metallic fuel, U-TRU-Zr alloy fuel, was selected and is being developed. For the fabrication of SFR metallic fuel elements, the endplug welding is a crucial process. The sealing of endplug to cladding tube should be hermetically perfect to prevent a leakage of fission gases and to maintain a good reactor performance. In this study, the welding technique, welding equipment, welding conditions and parameters were developed to make SFR metallic fuel elements. The TIG welding technique was adopted and the welding joint design was developed. And the optimal welding conditions and parameters were also established. In order to make SFR metallic fuel elements, the welding technique, welding equipment, welding conditions and parameters were developed. The TIG welding technique was adopted and the welding joint design was developed. And the optimal welding conditions and parameters were also established

  13. Neutron physics computation of CERCA fuel elements for Maria Reactor

    International Nuclear Information System (INIS)

    Andrzejewski, K.J.; Kulikowska, T.; Marcinkowska, Z.

    2008-01-01

    Neutron physics parameters of CERCA design fuel elements were calculated in the framework of the RERTR (Reduced Enrichment for Research and Test Reactors) program for Maria reactor. The analysis comprises burnup of experimental CERCA design fuel elements for 4 cycles in Maria Reactor To predict the behavior of the mixed core the differences between the CERCA fuel (485 g U-235 as U 3 Si 2 , 5 fuel tubes, low enrichment 19.75 % - LEU) and the presently used MR-6 fuel (430 g as UO 2 , 6 fuel tubes, high enrichment 36 % - HEU) had to be taken into account. The basic tool used in neutron-physics analysis of Maria reactor is program REBUS using in its dedicated libraries of effective microscopic cross sections. The cross sections were prepared using WIMS-ANL code, taking into account the actual structure, temperature and material composition of the fuel elements required preparation of new libraries.The problem is described in the first part of the present paper. In the second part the applicability of the new library is shown on the basis of the fuel core computational analysis. (author)

  14. Weld Joint Design for SFR Metallic Fuel Element Closures

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jung Won; Kim, Soo Sung; Woo, Yoon Myeng; Kim, Hyung Tae; Kim, Ki Hwan; Yoon, Kyung Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    The sodium-cooled fast reactor (SFR) system is among the six systems selected for Gen-IV promising systems and expected to become available for commercial introduction around 2030. In Korea, the R and D on SFR has been begun since 1997, as one of the national long-term nuclear R and D programs. The international collaborative research is under way on fuel developments within Advanced Fuel Project for Gen-IV SFR with the closed fuel cycle of full actinide recycling, while TRU bearing metallic fuel, U-TRU-Zr alloy fuel, was selected and is being developed. For the fabrication of SFR metallic fuel elements, the endplug welding is a crucial process. The sealing of endplug to cladding tube should be hermetically perfect to prevent a leakage of fission gases and to maintain a good reactor performance. In this study, the joint designs for endplug welding were investigated. For the irradiation test of SFR metallic fuel element, the TIG welding technique was adopted and the welding joint design was developed based on the welding conditions and parameters established. In order to make SFR metallic fuel elements, the weld joint design was developed based on the TIG welding technique.

  15. Endplug Welding Techniques developed for SFR Metallic Fuel Elements

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jung Won; Kim, Soo Sung; Woo, Yoon Myeng; Kim, Hyung Tae; Lee, Ho Jin; Kim, Ki Hwan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    In Korea, the R and D on SFR has been begun since 1997, as one of the national long-term nuclear R and D programs. The international collaborative research is under way on fuel developments within Advanced Fuel Project for Gen-IV SFR with the closed fuel cycle of full actinide recycling, while TRU bearing metallic fuel, U-TRU-Zr alloy fuel, was selected and is being developed. For the fabrication of SFR metallic fuel elements, the endplug welding is a crucial process. The sealing of endplug to cladding tube should be hermetically perfect to prevent a leakage of fission gases and to maintain a good reactor performance. In this study, the welding technique, welding equipment, welding conditions and parameters were developed to make SFR metallic fuel elements. The TIG welding technique was adopted and the welding joint design was developed. And the optimal welding conditions and parameters were also established. In order to make SFR metallic fuel elements, the welding technique, welding equipment, welding conditions and parameters were developed. The TIG welding technique was adopted and the welding joint design was developed. And the optimal welding conditions and parameters were also established.

  16. DART model for irradiation-induced swelling of dispersion fuel elements including aluminum-fuel interaction

    International Nuclear Information System (INIS)

    Rest, J.; Hofman, G.L.

    1997-01-01

    The Dispersion Analysis Research Tool (DART) contains models for fission-gas-induced fuel swelling, interaction of fuel with the matrix aluminum, for the resultant reaction-product swelling, and for the calculation of the stress gradient within the fuel particle. The effects of an aluminide shell on fuel particle swelling are evaluated. Validation of the model is demonstrated by a comparison of DART calculations of fuel swelling of U 3 SiAl-Al and U 3 Si 2 -Al for various dispersion fuel element designs with the data

  17. Transuranium element recovering method for spent nuclear fuel

    International Nuclear Information System (INIS)

    Todokoro, Akio; Kihara, Yoshiyuki; Okada, Hisashi

    1998-01-01

    Spent fuels are dissolved in nitric acid, the obtained dissolution liquid is oxidized by electrolysis, and nitric acid of transuranium elements are precipitated together with nitric acid of uranium elements from the dissolution solution and recovered. Namely, the transuranium elements are oxidized to an atomic value level at which nitric acid can be precipitated by an oxidizing catalyst, and cooled to precipitate nitric acid of transuranium elements together with nitric acid of transuranium elements, accordingly, it is not necessary to use a solvent which has been used so far upon recovering transuranium elements. Since no solvent waste is generated, a recovery method taking the circumstance into consideration can be provided. Further, nitric acid of uranium elements and nitric acid of transuranium elements precipitated and recovered together are dissolved in nitric acid again, cooled and only uranium elements are precipitated selectively, and recovered by filtration. The amount of wastes can be reduced to thereby enabling to mitigate control for processing. (N.H.)

  18. Uranium density reduction on fuel element side plates assessment

    International Nuclear Information System (INIS)

    Rios, Ilka A.; Andrade, Delvonei A.; Domingos, Douglas B.; Umbehaun, Pedro E.

    2011-01-01

    During operation of IEA-R1 research reactor, located at Instituto de Pesquisas Energeticas e Nucleares, IPEN - CNEN/SP, an abnormal oxidation on some fuel elements was noted. It was also verified, among the possible causes of the problem, that the most likely one was insufficient cooling of the elements in the core. One of the propositions to solve or minimize the problem is to reduce uranium density on fuel elements side plates. In this paper, the influence of this change on neutronic and thermal hydraulic parameters for IEA-R1 reactor is verified by simulations with the codes HAMMER and CITATION. Results are presented and discussed. (author)

  19. Uranium density reduction on fuel element side plates assessment

    Energy Technology Data Exchange (ETDEWEB)

    Rios, Ilka A. [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), Sao Paulo, SP (Brazil); Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Andrade, Delvonei A.; Domingos, Douglas B.; Umbehaun, Pedro E. [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    During operation of IEA-R1 research reactor, located at Instituto de Pesquisas Energeticas e Nucleares, IPEN - CNEN/SP, an abnormal oxidation on some fuel elements was noted. It was also verified, among the possible causes of the problem, that the most likely one was insufficient cooling of the elements in the core. One of the propositions to solve or minimize the problem is to reduce uranium density on fuel elements side plates. In this paper, the influence of this change on neutronic and thermal hydraulic parameters for IEA-R1 reactor is verified by simulations with the codes HAMMER and CITATION. Results are presented and discussed. (author)

  20. Nevada commercial spent nuclear fuel transportation experience

    International Nuclear Information System (INIS)

    1991-09-01

    The purpose of this report is to present an historic overview of commercial reactor spent nuclear fuel (SNF) shipments that have occurred in the state of Nevada, and to review the accident and incident experience for this type of shipments. Results show that between 1964 and 1990, 309 truck shipments covering approximately 40,000 miles moved through Nevada; this level of activity places Nevada tenth among the states in the number of truck shipments of SNF. For the same period, 15 rail shipments moving through the State covered approximately 6,500 miles, making Nevada 20th among the states in terms of number of rail shipments. None of these shipments had an accident or an incident associated with them. Because the data for Nevada are so limited, national data on SNF transportation and the safety of truck and rail transportation in general were also assessed

  1. Cost effectiveness of transportation fuels from biomass

    International Nuclear Information System (INIS)

    De Jager, D.; Faaij, A.P.C.; Troelstra, W.P.

    1998-06-01

    The aim of the study on the title subject was to investigate whether stimulation of the production and use of biofuels for transportation is worthwhile compared to the production of electricity from biomass. Several options are compared to each other and with reference technologies on the basis of the consumption or the avoided input of fossil fuels, emissions of greenhouse gases, specific costs and cost effectiveness. For each phase in the biomass conversion process (cultivation, pretreatment, transportation, conversion, distribution and final consumption) indicators were collected from the literature. Next to costs of the bioconversion routes attention is paid to other relevant aspects that are important for the introduction of the technological options in the Netherlands. 41 refs

  2. INTERACTION OF AIR TRANSPORTATION AND FUEL-SUPPLY COMPANIES

    Directory of Open Access Journals (Sweden)

    I. P. Zheleznaya

    2014-01-01

    Full Text Available The article describes the role of aviation fuel in the life of air transport. Fueling industry worldwide solves two main tasks - ensuring the safety and economy of air traffic. In Russia, there is one more task of airlines fuel supply. The article deals with fuel pricing taking into consideration today's realities.

  3. Validation of structural design of JHR fuel element

    International Nuclear Information System (INIS)

    Brisson, S.; Miras, G.; Le Bourdonnec, L.; Lemoine, P.; Anselmet, M.C.; Marelle, V.

    2010-01-01

    The validation of the structural design of the Jules Horowitz Reactor fuel element was made by the Finite Element Method, starting from the Computer Aided Design. The JHR fuel element is a cylindrical assembly of three sectors composed of eight rolled fuel plates. A roll-swaging process is used to join the fuel plates to three aluminium stiffeners. The hydraulic gap between each plate is 1.95 mm. The JHR fuel assembly is fastened at both ends to the upper and lower endfittings by riveting. The main stresses are essentially thermal loads, imposed on the fuel zone of the plates. These thermal loads result from the nuclear heat flux (W/cm 2 ). The mechanical loads are mainly hydraulic thrust forces. The average coolant velocity is 15 m/s. Seismic effects are also studied. The fuel assembly is entirely modelled by thin shells. The model takes into account asymmetric thermal loads which often appear in Research Reactors. The mechanics of the fuel plates vary in function of the burn up. These mechanical properties are derived from the data sets used in the MAIA code, and the validity of the structure is demonstrable at throughout the life of the fuel. Results concerning displacement are compared to functional criteria, while results concerning stress are compared to RCC-MX criteria. The results of this analysis show that the mechanical and geometrical integrity of the JHR fuel elements is respected for Operating Categories 1 and 2. This paper presents the methodology of this demonstration for the results obtained. (author)

  4. Space reactor fuel element testing in upgraded TREAT

    International Nuclear Information System (INIS)

    Todosow, M.; Bezler, P.; Ludewig, H.; Kato, W.Y.

    1993-01-01

    The testing of candidate fuel elements at prototypic operating conditions with respect to temperature, power density, hydrogen coolant flow rate, etc., is a crucial component in the development and qualification of nuclear rocket engines based on the Particle Bed Reactor (PBR), NERVA-derivative, and other concepts. Such testing may be performed at existing reactors, or at new facilities. A scoping study has been performed to assess the feasibility of testing PBR based fuel elements at the TREAT reactor. Initial results suggests that full-scale PBR elements could be tested at an average energy deposition of ∼60--80 MW-s/L in the current TREAT reactor. If the TREAT reactor was upgraded to include fuel elements with a higher temperture limit, average energy deposition of ∼100 MW/L may be achievable

  5. Space reactor fuel element testing in upgraded TREAT

    Science.gov (United States)

    Todosow, Michael; Bezler, Paul; Ludewig, Hans; Kato, Walter Y.

    1993-01-01

    The testing of candidate fuel elements at prototypic operating conditions with respect to temperature, power density, hydrogen coolant flow rate, etc., is a crucial component in the development and qualification of nuclear rocket engines based on the Particle Bed Reactor (PBR), NERVA-derivative, and other concepts. Such testing may be performed at existing reactors, or at new facilities. A scoping study has been performed to assess the feasibility of testing PBR based fuel elements at the TREAT reactor. Initial results suggests that full-scale PBR elements could be tested at an average energy deposition of ˜60-80 MW-s/L in the current TREAT reactor. If the TREAT reactor was upgraded to include fuel elements with a higher temperture limit, average energy deposition of ˜100 MW/L may be achievable.

  6. Brazing process in nuclear fuel element fabrication

    International Nuclear Information System (INIS)

    Katam, K.; Sudarsono

    1982-01-01

    The purpose of the brazing process is to join the spacers and pads of fuel pins, so that the process is meant as a soldering technique and not only as a hardening or reinforcing process such as in common brazing purposes. There are some preliminary processes before executing the brazing process such as: materials preparation, sand blasting, brazing metal coating tack welding the spacers and pads on the fuel cladding. The metal brazing used is beryllium in strip form which will be evaporated in vacuum condition to coat the spacers and pads. The beryllium vapor and dust is very hazardous to the workers, so all the line process of brazing needs specials safety protection and equipment to protect the workers and the processing area. Coating process temperature is 2470 deg C with a vacuum pressure of 10 -5 mmHg. Brazing process temperature process is 1060 deg C with a vacuum pressure of 10 -6 mmHg. The brazing process with beryllium coating probably will give metallurgical structural change in the fuel cladding metal at the locations of spacers and pads. The quality of brazing is highly influenced by and is depending on the chemical composition of the metal and the brazing metal, materials preparations, temperature, vacuum pressure, time of coating and brazing process. The quality control of brazing could be performed with methods of visuality geometry, radiography and metallography. (author)

  7. Alternative Fuels Data Center: Biodiesel Truck Transports Capitol Christmas

    Science.gov (United States)

    Tree Biodiesel Truck Transports Capitol Christmas Tree to someone by E-mail Share Alternative Fuels Data Center: Biodiesel Truck Transports Capitol Christmas Tree on Facebook Tweet about Alternative Fuels Data Center: Biodiesel Truck Transports Capitol Christmas Tree on Twitter Bookmark Alternative

  8. Contributions to LWR spent fuel storage and transport

    International Nuclear Information System (INIS)

    The papers included in this document describe the aspects of spent LWR fuel storage and transport-behaviour of spent fuel during storage; use of compact storage packs; safety of storage; design of storage facilities AR and AFR; description of transport casks and transport procedures

  9. Simulation on reactor TRIGA Puspati core kinetics fueled with thorium (Th) based fuel element

    Energy Technology Data Exchange (ETDEWEB)

    Mohammed, Abdul Aziz, E-mail: azizM@uniten.edu.my; Rahman, Shaik Mohmmed Haikhal Abdul [Universiti Tenaga Nasional. Jalan Ikram-UNITEN, 43000 Kajang, Selangor (Malaysia); Pauzi, Anas Muhamad, E-mail: anas@uniten.edu.my; Zin, Muhamad Rawi Muhammad; Jamro, Rafhayudi; Idris, Faridah Mohamad [Malaysian Nuclear Agency, Bangi, 43000 Kajang, Selangor (Malaysia)

    2016-01-22

    In confronting global energy requirement and the search for better technologies, there is a real case for widening the range of potential variations in the design of nuclear power plants. Smaller and simpler reactors are attractive, provided they can meet safety and security standards and non-proliferation issues. On fuel cycle aspect, thorium fuel cycles produce much less plutonium and other radioactive transuranic elements than uranium fuel cycles. Although not fissile itself, Th-232 will absorb slow neutrons to produce uranium-233 ({sup 233}U), which is fissile. By introducing Thorium, the numbers of highly enriched uranium fuel element can be reduced while maintaining the core neutronic performance. This paper describes the core kinetic of a small research reactor core like TRIGA fueled with a Th filled fuel element matrix using a general purpose Monte Carlo N-Particle (MCNP) code.

  10. Criticality safety requirements for transporting EBR-II fuel bottles stored at INTEC

    International Nuclear Information System (INIS)

    Lell, R. M.; Pope, C. L.

    2000-01-01

    Two carrier/shipping cask options are being developed to transport bottles of EBR-II fuel elements stored at INTEC. Some fuel bottles are intact, but some have developed leaks. Reactivity control requirements to maintain subcriticality during the hypothetical transport accident have been examined for both transport options for intact and leaking bottles. Poison rods, poison sleeves, and dummy filler bottles were considered; several possible poison materials and several possible dummy filler materials were studied. The minimum number of poison rods or dummy filler bottles has been determined for each carrier for transport of intact and leaking bottles

  11. A Study of Transport Airplane Crash-Resistant Fuel Systems

    National Research Council Canada - National Science Library

    Robertson, S

    2002-01-01

    ...), of transport airplane crash-resistant fuel system (CRFS). The report covers the historical studies related to aircraft crash fires and fuel containment concepts undertaken by the FAA, NASA, and the U.S...

  12. Analysis of the Range of Applicability of Thermodynamic Calculations in the Engineering of Nitride Fuel Elements

    Science.gov (United States)

    Ivanov, A. S.; Rusinkevich, A. A.; Belov, G. V.; Ivanov, Yu. A.

    2017-12-01

    The domains of applicability of thermodynamic calculations in the engineering of nitride fuel are analyzed. Characteristic values of the following parameters, which affect directly the concentration equilibration time, are estimated: nuclide production rate; characteristic times to local equilibrium in the considered temperature range; characteristic time needed for a stationary temperature profile to be established; characteristic time needed for a quasi-stationary concentration field to be established on a scale comparable to the size of a fuel pellet. It is demonstrated that equilibrium thermodynamic calculations are suitable for estimating the chemical and phase composition of fuel. However, a two-layer kinetic model should be developed in order to characterize the transport processes in condensed and gaseous phases. The process of diffusive transport needs to be taken into account in order to determine the composition in the hot region at the center of a fuel element.

  13. Nuclear fuel element and a method of manufacture thereof

    International Nuclear Information System (INIS)

    Wood, J.C.

    1975-01-01

    A nuclear fuel element having a sheath of zirconium or a zirconium alloy and a cross-linked siloxane lacquer coating on the inner surface of the sheath and separating the nuclear fuel material from the sheath is described. The siloxane lacquer coating retards cracking of the sheath by iodine vapor emitted by the fuel during burn-up, and acts as a lubricant for the fuel to prevent rupture of the sheath by thermal ratchetting of the fuel against the sheath and caused by differential thermal expansion between the fuel and the sheath. A silicone grease is applied as a thin layer in the sheath and then baked so that oxidative cleavage of the side chains of the grease occurs to form a cross-linked siloxane lacquer coating bonded to the sheath

  14. Regulation on the transport of nuclear fuel materials by vehicles

    International Nuclear Information System (INIS)

    1984-01-01

    The regulations applying to the transport of nuclear fuel materials by vehicles, mentioned in the law for the regulations of nuclear source materials, nuclear fuel materials and reactors. The transport is for outside of the factories and the site of enterprises by such modes of transport as rail, trucks, etc. Covered are the following: definitions of terms, places of fuel materials handling, loading methods, limitations on mix loading with other cargo, radiation dose rates concerning the containers and the vehicles, transport indexes, signs and indications, limitations on train linkage during transport by rail, security guards, transport of empty containers, etc. together with ordinary rail cargo and so on. (Mori, K.)

  15. Aircraft transporting container for nuclear fuel

    International Nuclear Information System (INIS)

    Kurakami, Jun-ichi; Kubo, Minoru.

    1991-01-01

    The present invention concerns an air craft transporting container for nuclear fuels. A sealing container that seals a nuclear fuel container and constitutes a sealed boundary for the transporting container is incorporated in an inner container. Shock absorbers are filled for absorbing impact shock energy in the gap between the inner container and the sealing container. The inner container is incorporated with wooden impact shock absorbers being filled so that it is situated in a substantially central portion of an external container. Partitioning cylinders are disposed coaxially in the cylindrical layer filled with wooden impact shock absorbers at an intermediate portion between the outer and the inner containers. Further, a plurality of longitudinally intersecting partitioning disks are disposed each at a predetermined distance in right and left cylindrical wooden impact shock absorbing layers which are in contact with the end face of the inner container. Accordingly, the impact shock energy can be absorbed by the wooden impact shock absorbers efficiently by a plurality of the partitioning disks and the partitioning cylinders. (I.N.)

  16. Memory list for the ordering of nuclear fuel elements with UO2 fuel

    International Nuclear Information System (INIS)

    1977-01-01

    The memory list will help to simplify and speed up the technical procedure of fuel element supply for nuclear reactors. Operators of nuclear power plants take great interest in the latest state of thechnology, if sufficiently tested, being applied with regard to material, manufacturing and testing methods. In order to obtain an unlimited availability of the nuclear plant in the future, this application of technology should be taken care of when designing and producing fuel elements. When ordering fuel elements special attention should be drawn to the interdependence of reactor and fuel element with reqard to design and construction, about which, howevers, no further details are given. When ordering fuel elements the operator give the producer all design data of the reactor core and the fuel elements as well as the planned operation mode. He also hands in the respective graphs and the required conditions for design so that a correct and detailed offer can be supplied. An exemplary extent of supply is shown in the given memory list. The regulations required herefore on passing technical material to the fuel element producers have to be established by agreements made by the customer. The order to be given should be itemized as follows: requirements, quality controland quality assurance, warranties and conditions, limits and extent of supply, terms of delivery. (orig./HP) [de

  17. Fuel Combustion Laboratory | Transportation Research | NREL

    Science.gov (United States)

    Fuel Combustion Laboratory Fuel Combustion Laboratory NREL's Fuel Combustion Laboratory focuses on designs, using both today's technology and future advanced combustion concepts. This lab supports the combustion chamber platform for fuel ignition kinetics research, was acquired to expand the lab's

  18. Drying Results of K-Basin Fuel Element 6603M (Rune 5)

    Energy Technology Data Exchange (ETDEWEB)

    B.M. Oliver; G.A. Ritter; G.S. Klinger; J. Abrefah; L.R. Greenwood; P.J. MacFarlan; S.C. Marschman

    1999-09-24

    The water-filled K-Basins in the Hanford 100 Area have been used to store N-Reactor spent nuclear fuel (SNF) since the 1970s. An Integrated Process Strategy (IPS) has been developed to package, dry, transport, and store these metallic uranium spent nuclear fuels in an interim storage facility on the Hanford Site (WHC 1995). Information required to support the development of the drying processes, and the required safety analyses, is being obtained from characterization tests conducted on fuel elements removed from the K-Basins. A series of drying tests (reported in separate documents, see Section 8.0) have been conducted by Pacific Northwest National Laboratory (PNNL) on several intact and damaged fuel elements recovered from both the K-East and K-West Basins. This report documents the results of the fifth of those tests conducted on an N-Reactor outer fuel element (6603M) which had been stored underwater in the Hanford 100 Area K-West basin from 1983 until 1996. This fuel element was subjected to a combination of low- and high-temperature vacuum drying treatments which were intended to mimic, wherever possible, the fuel treatment strategies of the IPS. The system used for the drying test was the Whole Element Furnace Testing System, described in Section 2.0. The test conditions and methodologies are given in Section 3.0. Inspections on the fuel element before and after the test are provided in Section 4.0. The experimental results are provided in Section 5.0. Discussion of the results is given in Section 6.0.

  19. Modelling of fission product release behavior from HTR spherical fuel elements under accident conditions

    International Nuclear Information System (INIS)

    Verfondern, K.; Mueller, D.

    1991-01-01

    Computer codes for modelling the fission product release behavior of spherical fuel elements for High Temperature Reactors (HTR) have been developed for the purpose of being used in risk analyses for HTRs. An important part of the validation and verification procedure for these calculation models is the theoretical investigation of accident simulation experiments which have been conducted in the KueFA test facility in the Hot Cells at KFA. The paper gives a presentation of the basic modeling and the calculational results of fission product release from modern German HTR fuel elements in the temperature range 1600-1800 deg. C using the TRISO coated particle failure model PANAMA and the diffusion model FRESCO. Measurements of the transient release behavior for cesium and strontium and of their concentration profiles after heating have provided informations about diffusion data in the important retention barriers of the fuel: silicon carbide and matrix graphite. It could be shown that the diffusion coefficients of both cesium and strontium in silicon carbide can significantly be reduced using a factor in the range of 0.02 - 0.15 compared to older HTR fuel. Also in the development of fuel element graphite, a tendency towards lower diffusion coefficients for both nuclides can be derived. Special heating tests focussing on the fission gases and iodine release from the matrix contamination have been evaluated to derive corresponding effective diffusion data for iodine in fuel element graphite which are more realistic than the iodine transport data used so far. Finally, a prediction of krypton and cesium release from spherical fuel elements under heating conditions will be given for fuel elements which at present are irradiated in the FRJ2, Juelich, and which are intended to be heated at 1600/1800 deg. C in the KueFA furnace in near future. (author). 7 refs, 11 figs

  20. Drying Results of K-Basin Fuel Element 6603M (Rune 5)

    International Nuclear Information System (INIS)

    Oliver, B.M.; Ritter, G.A.; Klinger, G.S.; Abrefah, J.; Greenwood, L.R.; MacFarlan, P.J.; Marschman, S.C.

    1999-01-01

    The water-filled K-Basins in the Hanford 100 Area have been used to store N-Reactor spent nuclear fuel (SNF) since the 1970s. An Integrated Process Strategy (IPS) has been developed to package, dry, transport, and store these metallic uranium spent nuclear fuels in an interim storage facility on the Hanford Site (WHC 1995). Information required to support the development of the drying processes, and the required safety analyses, is being obtained from characterization tests conducted on fuel elements removed from the K-Basins. A series of drying tests (reported in separate documents, see Section 8.0) have been conducted by Pacific Northwest National Laboratory (PNNL) on several intact and damaged fuel elements recovered from both the K-East and K-West Basins. This report documents the results of the fifth of those tests conducted on an N-Reactor outer fuel element (6603M) which had been stored underwater in the Hanford 100 Area K-West basin from 1983 until 1996. This fuel element was subjected to a combination of low- and high-temperature vacuum drying treatments which were intended to mimic, wherever possible, the fuel treatment strategies of the IPS. The system used for the drying test was the Whole Element Furnace Testing System, described in Section 2.0. The test conditions and methodologies are given in Section 3.0. Inspections on the fuel element before and after the test are provided in Section 4.0. The experimental results are provided in Section 5.0. Discussion of the results is given in Section 6.0

  1. Memento. Maritime transport of MOX fuels from Europe to Japan

    International Nuclear Information System (INIS)

    1999-07-01

    The maritime transport of MOX fuels from Europe to Japan represents the last of the 3 steps of transport of the nuclear fuel reprocessing-recycling program settled between ORC (Japan), BNFL (UK) and Cogema (France). This document summarizes the different aspects of this program: the companies concerned, the physical protection measures, the US-Japan agreements (accompanying warship), the in-depth safety, the handling of MOX fuels (containers and ships), and the Japan MOX fuel needs. (J.S.)

  2. Pyrochemical recovery of actinide elements from spent light water reactor fuel

    International Nuclear Information System (INIS)

    Johnson, G.K.; Pierce, R.D.; Poa, D.S.; McPheeters, C.C.

    1994-01-01

    Argonne National Laboratory is investigating salt transport and lithium pyrochemical processes for recovery of transuranic (TRU) elements from spent light water reactor fuel. The two processes are designed to recover the TRU elements in a form compatible with the Integral Fast Reactor (IFR) fuel cycle. The IFR is uniquely effective in consuming these long-lived TRU elements. The salt transport process uses calcium dissolved in Cu-35 wt % Mg in the presence of a CaCl 2 salt to reduce the oxide fuel. The reduced TRU elements are separated from uranium and most of the fission products by using a MgCl 2 transport salt. The lithium process, which does not employ a solvent metal, uses lithium in the presence of a LiCl salt as the reductant. After separation from the salt, the reduced metal is introduced into an electrorefiner, which separates the TRU elements from the uranium and fission products. In both processes, reductant and reduction salt are recovered by electrochemical decomposition of the oxide reaction product

  3. Fuel element production at BWX technologies

    International Nuclear Information System (INIS)

    Pace, Brett

    1997-01-01

    Effective July 1, 1997, the Government Group portion of the Babcock and Wilcox company was incorporated separately to become BWX Technologies, Inc. (BWXT) a wholly-owned subsidiary of the Babcock and Wilcox Company. The names of the divisions and other business units of the former Babcock and Wilcox Government Group (Advanced Systems Operations, Naval Nuclear Fuel Division, and Nuclear Equipment Division) will remain unchanged, but they are now known as divisions or business units of BWXT. The management of all units and their reporting relationships will likewise remain unchanged. (author)

  4. Production of pellets for nuclear fuel elements

    International Nuclear Information System (INIS)

    Butler, G.G.

    1982-01-01

    A method for producing nuclear fuel pellets each made up of a central portion and an outer annular portion surrounding the central portion, the two portions differing in composition. Such pellets are termed annular-layered pellets. The method comprises the steps of pressing powdered refractory material which has been granulated to form separately a central portion and an outer annular portion, assembling the portions together, compacting the assembly and sintering the compact. The portions are bonded together during sintering. The difference in composition may include a difference in density or isotopic enrichment as well as a chemical difference. (author)

  5. Fire and blast safety manual for fuel element manufacture

    International Nuclear Information System (INIS)

    Ensinger, U.; Koehler, B.; Mester, W.; Riotte, H.G.; Sehrbrock, H.W.

    1988-01-01

    The manual aims to enable people involved in the planning, operation, supervision, licensing or appraisal of fuel element factories to make a quick and accurate assessment of blast safety. In Part A, technical plant principles are shown, and a summary lists the flammable materials and ignition sources to be found in fuel element factories, together with theoretical details of what happens during a fire or a blast. Part B comprises a list of possible fires and explosions in fuel element factories and ways of preventing them. Typical fire and explosion scenarios are analysed more closely on the basis of experiments. Part B also contains a list and an assessment of actual fires and explosions which have occurred in fuel element factories. Part C contains safety measures to protect against fire and explosion, in-built fire safety, fire safety in plant design, explosion protection and measures to protect people from radiation and other hazards when fighting fires. A distinction is drawn between UO 2 , MOX and HTR fuel elements. (orig./DG) [de

  6. Commercial Aspect of Research Reactor Fuel Element Production

    International Nuclear Information System (INIS)

    Susanto, B.G; Suripto, A

    1998-01-01

    Several aspects affecting the commercialization of the Research Reactor Fuel Element Production Installation (RR FEPI) under a BUMN (state-owned company)have been studied. The break event point (BEP) value based on total production cost used is greatly depending upon the unit selling price of the fuel element. At a selling price of USD 43,500/fuel element, the results of analysis shows that the BEP will be reached at 51% of minimum available capacity. At a selling price of US$ 43.500/fuel element the total income (after tax) for 7 years ahead is US $ 4.620.191,- The net present value in this study has a positive value is equal to US $ 2.827.527,- the internal rate of return will be 18% which is higher than normal the bank interest rare (in US dollar) at this time. It is concluded therefore that the nuclear research reactor fuel element produced by state-owned company BUMN has a good prospect to be sold commercially

  7. Finite element simulation of thermal, elastic and plastic phenomena in fuel elements

    International Nuclear Information System (INIS)

    Soba, Alejandro; Denis, Alicia C.

    1999-01-01

    Taking as starting point an irradiation experiment of the first Argentine MOX fuel prototype, performed at the HFR reactor of Petten, Holland, the deformation suffered by the fuel element materials during burning has been numerically studied. Analysis of the pellet-cladding interaction is made by the finite element method. The code determines the temperature distribution and analyzes elastic and creep deformations, taking into account the dependency of the physical parameters of the problem on temperature. (author)

  8. In-pile tests of HTGR fuel particles and fuel elements

    International Nuclear Information System (INIS)

    Chernikov, A.S.; Kolesov, V.S.; Deryugin, A.I.

    1985-01-01

    Main types of in-pile tests for specimen tightness control at the initial step, research of fuel particle radiation stability and also study of fission product release from fuel elements during irradiation are described in this paper. Schemes and main characteristics of devices used for these tests are also given. Principal results of fission gas product release measurements satisfying HTGR demands are illustrated on the example of fuel elements, manufactured by powder metallurgy methods and having TRISO fuel particles on high temperature pyrocarbon and silicon carbide base. (author)

  9. Criticality safety evaluation for TWR-S fuel assembly transportation using TK-S16 containers

    International Nuclear Information System (INIS)

    Pesic, M.P.; Steljic, M.M.; Antic, D.P.

    2002-01-01

    Criticality safety issues, concerning transportation of fresh high-enriched uranium fuel elements (TWR-S fuel assembly type) with Russian containers TK-S16, are objects of study in this paper. Three-dimensional (3D) models of fuel element and container were made, based upon their well-known geometry and material structure. The way to pack fuel elements in a bundle inside of the container is proposed. Calculations were done by MCNP4B2 computer code. This Monte Carlo criticality code determined the effective multiplication factor from the cross-section data and specific geometry data. This evaluation demonstrated the subcriticality of a single package and an array of packages during normal conditions of transport and various hypothetical accident conditions. (author)

  10. Alternative transportation fuels in the USA: government hydrogen vehicle programs

    International Nuclear Information System (INIS)

    Cannon, J.S.

    1993-01-01

    The linkage between natural gas-based transportation and hydrogen-based transportation strategies, two clean burning gaseous fuels, provides a strong policy rationale for increased government sponsorship of hydrogen vehicle research and demonstration programs. Existing federal and state government hydrogen vehicle projects are discussed in this paper: research at the NREL, alternate-fueled buses, Renewable Hydrogen for the State of Hawaii program, New York state alternative transportation fuels program, Colorado program. 9 refs

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  12. Transportation of radioactive wastes from nuclear fuel cycles

    International Nuclear Information System (INIS)

    1979-09-01

    This paper discusses current and foreseen radioactive waste transportation systems as they apply to the INFCE Working Group 7 study. The types of wastes considered include spent fuel, which is treated as a waste in once-through fuel cycles; high-, medium-, and low-level waste; and gaseous waste. Regulatory classification of waste quantities and containers applicable to these classifications are discussed. Radioactive wastes are presently being transported in a safe and satisfactory manner. None of the INFCE candidate fuel cycles pose any extraordinary problems to future radioactive waste transportation and such transportation will not constitute a decisive factor in the choice of a preferred fuel cycle

  13. Postulated accident scenarios for the on-site transport of spent nuclear fuel

    International Nuclear Information System (INIS)

    Morandin, G.; Sauve, R.

    2004-01-01

    Once a spent fuel container is loaded with spent fuel it typically travels on-site to a processing building for permanent lid attachment. During on-site transport a lid clamp is utilized to ensure the container lid remains in place. The safe on-site transport of spent nuclear fuel must rely on the structural integrity of the transport container and system of transport. Regard for on-site traffic and safe, efficient travel routes are important and manageable with well thought-out planning. Non-manageable incidences, such as flying debris from tornado force winds or postulated blasts in proximity to the transport container, that may result in high velocity impact and shock loading on the transport system must be considered. This paper consists of simulations that consider these types of postulated accident scenarios using detailed nonlinear finite element techniques

  14. Performance and management of IPR-R1 fuel elements

    International Nuclear Information System (INIS)

    Stasiulevicius, R.; Maretti Junior, F.

    1983-01-01

    The performance of fuel elements during the 23 years of the reactor operation, is presented aiming to introduce improvements in the fuel load distribution and consequent increase of the reactivity. A computer code CORE was developed aiming to calculate the individual burnup of the fuel elements and the value of the reactivity for several core configurations, establishing a routine to control the nuclear material in the IPR-R1. The values calculated were compared with the experimental results. Some alternatives to augment the reactivity of the present core are presented foreseeing the fuel load availability for operation with 100Km and, for angmenting the power reaction in a next stage. (E.G.) [pt

  15. Fission product release from HTGR coated microparticles and fuel elements

    International Nuclear Information System (INIS)

    Gusev, A.A.; Deryugin, A.I.; Lyutikov, R.A.; Chernikov, A.S.

    1991-01-01

    The article presents the results of the investigation of fission products release from microparticles with UO 2 core and five-layer HII PyC- and SiC base protection layers of TRICO type as well as from spherical fuel elements based thereon. It is shown that relative release of short-lived xenon and crypton from microparticles does not exceed (2-3) 10 -7 . The release of gaseous fission products from fuel elements containing no damaged coated microparticles, is primarily determined by the contamination of matrix graphite with fuel. An analytical dependence is derived, the dependence described the relation between structural parameters of coated microparticles, irradiation conditions and fuel burnup at which depressurization of coated microparticles starts

  16. Development and testing of the EDF-2 reactor fuel element

    International Nuclear Information System (INIS)

    Delpeyroux, P.

    1964-01-01

    This technical report reviews the work which has been necessary for defining the EDF-2 fuel element. After giving briefly the EDF-2 reactor characteristics and the preliminary choice of parameters which made it possible to draw up a draft plan for the fuel element, the authors consider the research proper: - Uranium studies: tests on the passage into the β phase of an internal crown of a tube, bending of the tube under the effect of a localized force, welding of the end-pellets and testing for leaks. The resistance of the tube to crushing and of the pellets to yielding under the external pressure have been studied in detail in another CEA report. - Can studies: conditions of production and leak proof testing of the can, resistance of the fins to creep due to the effect of the gas flow. - Studies of the extremities of the element: creep under compression and welding of the plugs to the can. - Cartridge studies: determination of the characteristics of the can fuel fixing grooves and of the canning conditions, verification of the resistance of the fuel element to thermal cycling, determination of the temperature drop at the can-fuel interface dealt with in more detail in another CEA report. - Studies of the whole assembly: this work which concerns the graphite jacket, the support and the cartridge vibrations has been carried out by the Mechanical and Thermal Study Service (Mechanics Section). In this field the Fuel Element Study Section has investigated the behaviour of the centering devices in a gas current. The outcome of this research is the defining of the plan of the element the production process and the production specifications. The validity of ail these out-of-pile tests will be confirmed by the in-pile tests already under way and by irradiation of the elements in the EDF-2 reactor itself. In conclusion the programme is given for improving the fuel element and for defining the fuel element for the second charge. (authors) [fr

  17. Calculating the plutonium in spent fuel elements

    International Nuclear Information System (INIS)

    Barnham, Keith

    1992-01-01

    Many members of the public are concerned about plutonium. They are worried about its environmental, health and proliferation risks. Fundamental to all such considerations are two related questions: how much plutonium do nuclear reactors produce ? and how accurately do the relevant authorities know these production figures ? These two questions have been studied with particular reference to the UK civil Magnox reactors. In 1990 these were still the only UK civil reactors whose spent fuel had been reprocessed to extract plutonium in routine production. It has not been possible to conclude that the relevant government industry and safeguard authorities are aware of how much plutonium these reactors produce and that the figures are known to the highest achievable accuracy. To understand why, this chapter will outline some of the history of the attempts to get answers to these two questions. (author)

  18. Method of making a graphite fuel element having carbonaceous fuel bodies

    International Nuclear Information System (INIS)

    Miertschin, G.N.; Leary, D.F.

    1977-01-01

    Particulate nuclear fuel material, particulate carbon and pitch are combined with an additive which is effective to reduce the coke yield upon carbonization to mold a green fuel body. The additive may be polystyrene, a styrene-butadiene copolymer, an aromatic hydrocarbon having a molecular weight between about 75 and 300 or a saturated hydrocarbon polymer. The green fuel body is inserted in a complementary cavity within a porous nuclear fuel element body and heated in situ to decompose the pitch and additive, leaving a relatively close-fitting fuel body in the cavity

  19. The manufacture of LEU fuel elements at Dounreay

    Energy Technology Data Exchange (ETDEWEB)

    Gibson, J.

    1997-08-01

    Two LEU test elements are being manufactured at Dounreay for test irradiation in the HFR at Petten, The Netherlands. This paper describes the installation of equipment and the development of the fabrication and inspection techniques necessary for the manufacture of LEU fuel plates. The author`s experience in overcoming the technical problems of stray fuel particles, dog-boning, uranium homogeneity and the measurement of uranium distribution is also described.

  20. 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

  1. Analysis of the ATR fuel element swaging process

    International Nuclear Information System (INIS)

    Richins, W.D.; Miller, G.K.

    1995-12-01

    This report documents a detailed evaluation of the swaging process used to connect fuel plates to side plates in Advanced Test Reactor (ATR) fuel elements. The swaging is a mechanical process that begins with fitting a fuel plate into grooves in the side plates. Once a fuel plate is positioned, a lip on each of two side plate grooves is pressed into the fuel plate using swaging wheels to form the joints. Each connection must have a specified strength (measured in terms, of a pullout force capacity) to assure that these joints do not fail during reactor operation. The purpose of this study is to analyze the swaging process and associated procedural controls, and to provide recommendations to assure that the manufacturing process produces swaged connections that meet the minimum strength requirement. The current fuel element manufacturer, Babcock and Wilcox (B ampersand W) of Lynchburg, Virginia, follows established procedures that include quality inspections and process controls in swaging these connections. The procedures have been approved by Lockheed Martin Idaho Technologies and are designed to assure repeatability of the process and structural integrity of each joint. Prior to July 1994, ATR fuel elements were placed in the Hydraulic Test Facility (HTF) at the Idaho National Engineering Laboratory (AGNAIL), Test Reactor Area (TRA) for application of Boehmite (an aluminum oxide) film and for checking structural integrity before placement of the elements into the ATR. The results presented in this report demonstrate that the pullout strength of the swaged connections is assured by the current manufacturing process (with several recommended enhancements) without the need for- testing each element in the HTF

  2. Ecological aspects of water coal fuel transportation and application

    Directory of Open Access Journals (Sweden)

    Anna SHVORNIKOVA

    2010-01-01

    Full Text Available This paper deals with the aspects of influence of transportation process and burning of water coal fuel on an ecological condition of environment. Also mathematical dependences between coal ash level and power consumption for transportation are presented.

  3. BNFL's new spent fuel transport flask - Excellox 8

    International Nuclear Information System (INIS)

    McWilliam, D.S.

    2002-01-01

    Since British Nuclear Fuels plc (BNFL) was formed in 1971 its transport service has safely moved spent light water reactor fuel from many locations abroad to its fuel handling plants at Sellafield in the UK. To support this business a number of types of flasks have been designed and used. One of the types used has been the Excellox family of water-filled flasks. To support future business opportunities a new flask, designed to meet the requirements of the new IAEA transport regulations TS-R-1 (ST-1, Revised), has been developed. The flask will be a type B(U)F. This new flask design will maximise fuel carrying capacity to minimise transport costs. The design capacity of the new Excellox 8 flask is to be 12 pressurised water reactor or 32 boiling water reactor fuel assemblies. The objective of this BNFL project is to provide another economic spent nuclear fuel transport system, in support of BNFL transport business. (author)

  4. Irradiation program of slightly enriched fuel elements at the Atucha I nuclear power plant

    International Nuclear Information System (INIS)

    Casario, J.A.; Cesario, R.H.; Perez, R.A.; Sidelnik, J.I.

    1987-01-01

    An irradiation program of fuel elements with slightly enriched uranium is implemented, tending to the homogenization of core at Atucha I nuclear power plant. The main benefits of the enrichment program are: a) to extend the average discharge burnup of fuel elements, reducing the number of elements used to generate the same amount of energy. This implies a smaller annual consumption of elements and consequently the reduction of transport and replacement operations and of the storage pool systems as well as that of radioactive wastes; b) the saving of uranium and structural materials (Zircaloy and others). In the initial stage of program an homogeneous core enrichment of 0.85% by weight of U-235 is anticipated. The average discharge burnup of fuel elements, as estimated by previous studies, is approximately 11.6 MW d/kg U. The annual consumption of fuel elements is reduced from 396 of natural uranium to 205, with a load factor of 0.85. It is intended to reach the next equilibrium steps with an enrichment of 1.00 and 1.20% in U-235. (Author)

  5. Low Cost Nuclear Thermal Rocket Cermet Fuel Element Environment Testing

    Science.gov (United States)

    Bradley, David E.; Mireles, Omar R.; Hickman, Robert R.

    2011-01-01

    Deep space missions with large payloads require high specific impulse (Isp) and relatively high thrust in order to achieve mission goals in reasonable time frames. Conventional, storable propellants produce average Isp. Nuclear thermal rockets (NTR) capable of high Isp thrust have been proposed. NTR employs heat produced by fission reaction to heat and therefore accelerate hydrogen which is then forced through a rocket nozzle providing thrust. Fuel element temperatures are very high (up to 3000K) and hydrogen is highly reactive with most materials at high temperatures. Data covering the effects of high temperature hydrogen exposure on fuel elements is limited. The primary concern is the mechanical failure of fuel elements which employ high-melting-point metals, ceramics or a combination (cermet) as a structural matrix into which the nuclear fuel is distributed. It is not necessary to include fissile material in test samples intended to explore high temperature hydrogen exposure of the structural support matrices. A small-scale test bed designed to heat fuel element samples via non-contact RF heating and expose samples to hydrogen is being developed to assist in optimal material and manufacturing process selection without employing fissile material. This paper details the test bed design and results of testing conducted to date.

  6. Properties of U3Si2-Al dispersion fuel element and its application

    International Nuclear Information System (INIS)

    Yin Changgeng

    2001-01-01

    The properties of U 3 Si 2 fuel and U 3 Si 2 -Al dispersion fuel element are introduced, which include U-loading; the banding quality, U-homogeneity and 'dog-bone' phenomenon, the minimum thickness of cladding and the corrosion performances. The fabrication technique of fuel elements, NDT for fuel plates, assemble technique of fuel elements and the application of U 3 Si 2 -Al dispersion fuel elements in the world are introduced

  7. Expert system for surveillance and diagnosis of breach fuel elements

    Science.gov (United States)

    Gross, Kenny C.

    1989-01-01

    An apparatus and method are disclosed for surveillance and diagnosis of breached fuel elements in a nuclear reactor. A delayed neutron monitoring system provides output signals indicating the delayed neutron activity and age and the equivalent recoil areas of a breached fuel element. Sensors are used to provide outputs indicating the status of each component of the delayed neutron monitoring system. Detectors also generate output signals indicating the reactor power level and the primary coolant flow rate of the reactor. The outputs from the detectors and sensors are interfaced with an artificial intelligence-based knowledge system which implements predetermined logic and generates output signals indicating the operability of the reactor.

  8. Expert system for surveillance and diagnosis of breach fuel elements

    International Nuclear Information System (INIS)

    Gross, K.C.

    1989-01-01

    An apparatus and method are disclosed for surveillance and diagnosis of breached fuel elements in a nuclear reactor. A delayed neutron monitoring system provides output signals indicating the delayed neutron activity and age and the equivalent recoil areas of a breached fuel element. Sensors are used to provide outputs indicating the status of each component of the delayed neutron monitoring system. Detectors also generate output signals indicating the reactor power level and the primary coolant flow rate of the reactor. The outputs from the detectors and sensors are interfaced with an artificial intelligence-based knowledge system which implements predetermined logic and generates output signals indicating the operability of the reactor

  9. Design and research of fuel element for pulsed reactor

    International Nuclear Information System (INIS)

    Tian Sheng

    1994-05-01

    The fuel element is the key component for pulsed reactor and its design is one of kernel techniques for pulsed reactor. Following the GA Company of US the NPIC (Nuclear Power Institute of China) has mastered this technique. Up to now, the first pulsed reactor in China (PRC-1) has been safely operated for about 3 years. The design and research of fuel element undertaken by NPIC is summarized. The verification and evaluation of this design has been carried out by using the results of measured parameters during operation and test of PRC-1 as well as comparing the design parameters published by others

  10. Reactor fuel element heat conduction via numerical Laplace transform inversion

    International Nuclear Information System (INIS)

    Ganapol, Barry D.; Furfaro, Roberto

    2001-01-01

    A newly developed numerical Laplace transform inversion (NLTI) will be presented to determine the transient temperature distribution within a nuclear reactor fuel element. The NLTI considered in this presentation has evolved to its present state over the past 10 years of application. The methodology adopted is one that relies on acceleration of the convergence of an infinite series towards its limit. The inversion will be applied to the prediction of the transient temperature distribution within an MTR type nuclear fuel element through a novel formulation of the solution to the transformed heat conduction equation. (author)

  11. Reactor fuel element heat conduction via numerical Laplace transform inversion

    Energy Technology Data Exchange (ETDEWEB)

    Ganapol, Barry D.; Furfaro, Roberto [University of Arizona, Tucson, AZ (United States). Dept. of Aerospace and Mechanical Engineering], e-mail: ganapol@cowboy.ame.arizona.edu

    2001-07-01

    A newly developed numerical Laplace transform inversion (NLTI) will be presented to determine the transient temperature distribution within a nuclear reactor fuel element. The NLTI considered in this presentation has evolved to its present state over the past 10 years of application. The methodology adopted is one that relies on acceleration of the convergence of an infinite series towards its limit. The inversion will be applied to the prediction of the transient temperature distribution within an MTR type nuclear fuel element through a novel formulation of the solution to the transformed heat conduction equation. (author)

  12. Experimental study of some mounting brackets to support fuel elements

    International Nuclear Information System (INIS)

    Aubert, M.; Poglia, S.; Roche, R.

    1958-09-01

    In an atomic pile with vertical channels, fuel elements are stacked on one another. According to a possible assembly, fuel element can be contained by a graphite sleeve and be supported by a mounting bracket in this sleeve. Sleeves are then stacked on one another. The authors report the investigation of different designs for these mounting brackets. They describe their mechanical role and their mechanical, aerodynamic, neutronic and test conditions. They report tests performed on brackets made in graphite and on brackets made in stainless steel and graphite, and discuss the obtained results

  13. Store for burnt-up fuel elements of nuclear reactors

    International Nuclear Information System (INIS)

    Kumpf, H.

    1981-01-01

    Burnt-up fuel elements of nuclear reactors have to be cooled during storage. For this reason the boxes which surround the fuel elements can have cooling air flowing round them in natural flow. This air is taken through the walls of a storage building through zones of parallel pipes, whose diameter and spacing are in the ratio of 1 : 0.5 to 1 : 2. The pipes have dust filters. Prefilters with fan drive are situated in parallel with the inlet pipe zones. (orig.) [de

  14. Fuel element load/unload machine for the PEC reactor

    International Nuclear Information System (INIS)

    Clayton, K.F.

    1984-01-01

    GEC Energy Systems Limited are providing two fuel element load/unload machines for use in the Italian fast reactor programme. One will be used in the mechanism test facility (IPM) at Casaccia, to check the salient features of the machine operating in a sodium environment prior to the second machine being installed in the PEC Brasimone Reactor. The machine is used to handle fuel elements, control rods and other reactor components in the sodium-immersed core of the reactor. (U.K.)

  15. FABRICATION OF TUBE TYPE FUEL ELEMENT FOR NUCLEAR REACTORS

    Science.gov (United States)

    Loeb, E.; Nicklas, J.H.

    1959-02-01

    A method of fabricating a nuclear reactor fuel element is given. It consists essentially of fixing two tubes in concentric relationship with respect to one another to provide an annulus therebetween, filling the annulus with a fissionablematerial-containing powder, compacting the powder material within the annulus and closing the ends thereof. The powder material is further compacted by swaging the inner surface of the inner tube to increase its diameter while maintaining the original size of the outer tube. This process results in reduced fabrication costs of powdered fissionable material type fuel elements and a substantial reduction in the peak core temperatures while materially enhancing the heat removal characteristics.

  16. The source regime for irradiation plant operated with fuel elements

    International Nuclear Information System (INIS)

    Suckow, W.

    1976-11-01

    The rapid and irregular decay of the gamma radiation from reactor fuel elements requires the establishment of an optimal source regime in order to utilise reactor fuel elements as radiation sources on a technological basis. Critical values have been derived which enable the determination of optimal conditions. In this context all technologically interesting types of source regimes have been examined. Methods to achieve a high gamma yield and a satisfactory dose consistency with time have been developed and important values for these two aspects have been derived. The conditions for optimal radiation source regimes are described in the final conclusions. (author)

  17. Thermal analysis on NAC-STC spent fuel transport cask under different transport conditions

    Energy Technology Data Exchange (ETDEWEB)

    Xu, Yumei [Institute of Process Equipment, Zhejiang University, Hangzhou (China); Yang, Jian, E-mail: zdhjkz@zju.edu.cn [Institute of Process Equipment, Zhejiang University, Hangzhou (China); Xu, Chao; Wang, Weiping [Institute of Process Equipment, Zhejiang University, Hangzhou (China); Ma, Zhijun [Department of Material Engineering, South China University of Technology, Guangzhou (China)

    2013-12-15

    Highlights: • Spent fuel cask was investigated as a whole instead of fuel assembly alone. • The cask was successfully modeled and meshed after several simplifications. • Equivalence method was used to calculate the properties of parts. • Both the integral thermal field and peak values are captured to verify safety. • The temperature variations of key parts were also plotted. - Abstract: Transport casks used for conveying spent nuclear fuel are inseparably related to the safety of the whole reprocessing system for spent nuclear fuel. Thus they must be designed according to rigorous safety standards including thermal analysis. In this paper, for NAC-STC cask, a finite element model is established based on some proper simplifications on configurations and the heat transfer mechanisms. Considering the complex components and gaps, the equivalence method is presented to define their material properties. Then an equivalent convection coefficient is introduced to define boundary conditions. Finally, the temperature field is captured and analyzed under both normal and accident transport conditions by using ANSYS software. The validity of numerical calculation is given by comparing its results with theoretical calculation. Obtaining the integral distribution laws of temperature and peak temperature values of all vital components, the security of the cask can be evaluated and verified.

  18. Design and fabrication procedures of Super-Phenix fuel elements

    International Nuclear Information System (INIS)

    Leclere, J.; Vialard, J.-L.; Delpeyroux, P.

    1975-01-01

    For Super-Phenix fuel assemblies, Phenix technological arrangements will be used again, but they will be simplified as far as possible. The maximum fuel can temperature has been lowered in order to obtain a good behavior of hexagonal tubes and cans at high irradiation levels. An important experimental programme and the experience gained from Phenix operation will confirm the merits of the options retained. The fuel element fabrication is envisaged to take place in the plutonium workshop at Cadarache. Usual procedures will be employed and both reliability and automation will be increased [fr

  19. Pyrochemical head-end treatment for fast reactor fuel elements

    International Nuclear Information System (INIS)

    Avogadro, A.

    1978-01-01

    The paper presents the R and D work performed at Ispra and Mol during the period 1965-1975 in order to find a way to overcome technical and economical difficulties arising when the conventional reprocessing is applied to fast reactor fuel elements. The work had been directed towards 3 specific topics: a) liquid-metal decladding of spent stainless steel - clad fuels (solinox process). b) oxidative pulverisation by fused salts and extraction of volatile fission products (satex process). c) Pyrochemical separation of plutonium from the bulk of the fuel

  20. Finite element analysis of advanced neutron source fuel plates

    International Nuclear Information System (INIS)

    Luttrell, C.R.

    1995-08-01

    The proposed design for the Advanced Neutron Source reactor core consists of closely spaced involute fuel plates. Coolant flows between the plates at high velocities. It is vital that adjacent plates do not come in contact and that the coolant channels between the plates remain open. Several scenarios that could result in problems with the fuel plates are studied. Finite element analyses are performed on fuel plates under pressure from the coolant flowing between the plates at a high velocity, under pressure because of a partial flow blockage in one of the channels, and with different temperature profiles

  1. Drying damaged K West fuel elements (Summary of whole element furnace runs 1 through 8); TOPICAL

    International Nuclear Information System (INIS)

    LAWRENCE, L.A.

    1998-01-01

    N Reactor fuel elements stored in the Hanford K Basins were subjected to high temperatures and vacuum conditions to remove water. Results of the first series of whole element furnace tests i.e., Runs 1 through 8 were collected in this summary report. The report focuses on the six tests with breached fuel from the K West Basin which ranged from a simple fracture at the approximate mid-point to severe damage with cladding breaches at the top and bottom ends with axial breaches and fuel loss. Results of the tests are summarized and compared for moisture released during cold vacuum drying, moisture remaining after drying, effects of drying on the fuel element condition, and hydrogen and fission product release

  2. Neutronic analysis of a fuel element with variations in fuel enrichment and burnable poison

    Energy Technology Data Exchange (ETDEWEB)

    Faria, Rochkhudson B. de; Martins, Felipe; Velasquez, Carlos E.; Cardoso, Fabiano; Fortini, Angela; Pereira, Claubia, E-mail: rochkdefaria@yahoo.com.br, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear

    2017-07-01

    In this work, the goal was to evaluate the neutronic behavior during the fuel burnup changing the amount of burnable poison and fuel enrichment. For these analyses, it was used a 17 x 17 PWR fuel element, simulated using the 238 groups library cross-section collapsed from ENDF/BVII.0 and TRITON module of SCALE 6.0 code system. The results confirmed the effective action of the burnable poison in the criticality control, especially at Beginning Of Cycle (BOC) and in the burnup kinetics, because at the end of the fuel cycle there was a minimal residual amount of neutron absorbers ({sup 155}Gd and {sup 157}Gd), as expected. At the end of the cycle, the fuel element was still critical in all simulated situations, indicating the possibility of extending the fuel burn. (author)

  3. LMFBR fuel-design environment for endurance testing, primarily of oxide fuel elements with local faults

    International Nuclear Information System (INIS)

    Warinner, D.K.

    1980-01-01

    The US Department of Energy LMFBR Lines-of-Assurance are briefly stated and local faults are given perspective with an historical review and definition to help define the constraints of LMFBR fuel-element designs. Local-fault-propagation (fuel-element failure-propagation and blockage propagation) perceptions are reviewed. Fuel pin designs and major LMFBR parameters affecting pin performance are summarized. The interpretation of failed-fuel data is aided by a discussion of the effects of nonprototypicalities. The fuel-pin endurance expected in the US, USSR, France, UK, Japan, and West Germany is outlined. Finally, fuel-failure detection and location by delayed-neutron and gaseous-fission-product monitors are briefly discussed to better realize the operational limits

  4. Improved lumped parameter for annular fuel element thermohydraulic analysis

    International Nuclear Information System (INIS)

    Duarte, Juliana Pacheco; Su, Jian; Alvim, Antonio Carlos Marques

    2011-01-01

    Annular fuel elements have been intensively studied for the purpose of increasing power density in light water reactors (LWR). This paper presents an improved lumped parameter model for the dynamics of a LWR core with annular fuel elements, composed of three sub-models: the fuel dynamics model, the neutronics model, and the coolant energy balance model. The transient heat conduction in radial direction is analyzed through an improved lumped parameter formulation. The Hermite approximation for integration is used to obtain the average temperature of the fuel and cladding and also to obtain the average heat flux. The volumetric heat generation in fuel rods was obtained with the point kinetics equations with six delayed neutron groups. The equations for average temperature of fuel and cladding are solved along with the point kinetic equations, assuming linear reactivity and coolant temperature in cases of reactivity insertion. The analytical development of the model and the numeric solution of the ordinary differential equation system were obtained by using Mathematica 7.0. The dynamic behaviors for average temperatures of fuel, cladding and coolant in transient events as well as the reactor power were analyzed. (author)

  5. Nature of the elements transporting long-chain fatty acids through the red cell membrane

    DEFF Research Database (Denmark)

    Bojesen, Inge Norby; Bojesen, Eigil

    1998-01-01

    Docosahexaenoic acid, linoleic acid, red cell membrane, transporting elements, transport kinetics, fatty acid transport......Docosahexaenoic acid, linoleic acid, red cell membrane, transporting elements, transport kinetics, fatty acid transport...

  6. Comparison of the transportation risks for the spent fuel in Korea for different transportation scenarios

    International Nuclear Information System (INIS)

    Jeong, Jongtae; Cho, D.K.; Choi, H.J.; Choi, J.W.

    2011-01-01

    According to the long term management strategy for spent fuels in Korea, they will be transported from the spent fuel pools in each nuclear power plant to the central interim storage facility (CISF) which is to start operation in 2016. At the start of the operation of the final repository (FR), by the year 2065, transport will then take place between the CISF and the FR. Therefore, we have to determine the safe and economical logistics for the transportation of these spent fuels by considering their transportation risks and costs. In this study, we developed four transportation scenarios for a maritime transportation by considering the type of transportation casks and transport means in order to suggest safe and economical transportation logistics for the spent fuels in Korea. And, we estimated and compared the transportation risks for these four transportation scenarios. Also, we estimated and compared the transportation risks resulting from accidents during the transportation of PWR and PHWR spent fuels by road trailers from the CISF and the FR. From the results of this study, we found that risks resulting from accidents during the transportation of the spent fuels have a very low radiological risk activity with a manageable safety and health consequences. The results of this study can be used as basic data for the development of safe and economical logistics for a transportation of the spent fuels in Korea by considering the transportation costs for the four scenarios which will be needed in the near future.

  7. Alternative Fuel Guidelines for Alternative Transportation Systems.

    Science.gov (United States)

    2011-01-31

    The Volpe Center documented the increased use of alternative fuels on vehicles owned and operated by federal land management agencies. For each alternative fuel type, the Volpe Center documented the availability of vehicles, fueling mechanisms and pr...

  8. FREVAP-6, Metal Fission Products Release from HTGR Fuel Elements

    International Nuclear Information System (INIS)

    Pierce, V.H.

    2005-01-01

    1 - Description of problem or function: The FREVAP type of code for estimating the release of longer-lived metallic fission products from HTGR fuel elements has been developed to take into account the combined effects of the retention of metallic fission products by fuel particles and the rather strong absorption of these fission products by the graphite of the fuel elements. Release calculations are made on the basis that the loss of fission product nuclides such as strontium, cesium, and barium is determined by their evaporation from the graphite surfaces and their transpiration induced by the flowing helium coolant. The code is devised so that changes of fission rate (fuel element power), fuel temperature, and graphite temperature may be incorporated into the calculation. Temperature is quite important in determining release because, in general, both release from fuel particles and loss by evaporation (transpiration) vary exponentially with the reciprocal of the absolute temperature. NESC0301/02: This version differs from the previous one in the following points: The source and output files were converted from BCD to ASCII coding. 2 - Method of solution: A problem is defined as having a one-dimensional segment made up of three parts - (1) the fission product source (fuel particles) in series with, (2) a non-source and absorption part (element graphite) and (3) a surface for evaporation to the coolant (graphite-helium interface). More than one segment may be connected (possibly segments stacked axially) by way of the coolant. At any given segment, a continuity equation is solved assuming equilibrium between the source term, absorption term, evaporation at coolant interface and the partial pressure of the fission product isotope in the coolant. 3 - Restrictions on the complexity of the problem - Maxima of: 5 isotopes; 10 time intervals for time-dependent variable; 49 segments (times number of isotopes); 5 different output print time-steps

  9. Induction Heating Model of Cermet Fuel Element Environmental Test (CFEET)

    Science.gov (United States)

    Gomez, Carlos F.; Bradley, D. E.; Cavender, D. P.; Mireles, O. R.; Hickman, R. R.; Trent, D.; Stewart, E.

    2013-01-01

    Deep space missions with large payloads require high specific impulse and relatively high thrust to achieve mission goals in reasonable time frames. Nuclear Thermal Rockets (NTR) are capable of producing a high specific impulse by employing heat produced by a fission reactor to heat and therefore accelerate hydrogen through a rocket nozzle providing thrust. Fuel element temperatures are very high (up to 3000 K) and hydrogen is highly reactive with most materials at high temperatures. Data covering the effects of high-temperature hydrogen exposure on fuel elements are limited. The primary concern is the mechanical failure of fuel elements due to large thermal gradients; therefore, high-melting-point ceramics-metallic matrix composites (cermets) are one of the fuels under consideration as part of the Nuclear Cryogenic Propulsion Stage (NCPS) Advance Exploration System (AES) technology project at the Marshall Space Flight Center. The purpose of testing and analytical modeling is to determine their ability to survive and maintain thermal performance in a prototypical NTR reactor environment of exposure to hydrogen at very high temperatures and obtain data to assess the properties of the non-nuclear support materials. The fission process and the resulting heating performance are well known and do not require that active fissile material to be integrated in this testing. A small-scale test bed; Compact Fuel Element Environmental Tester (CFEET), designed to heat fuel element samples via induction heating and expose samples to hydrogen is being developed at MSFC to assist in optimal material and manufacturing process selection without utilizing fissile material. This paper details the analytical approach to help design and optimize the test bed using COMSOL Multiphysics for predicting thermal gradients induced by electromagnetic heating (Induction heating) and Thermal Desktop for radiation calculations.

  10. Alternative Fuels in Transportation : Workforce needs and opportunities in support of reducing reliance on petroleum fuels

    Science.gov (United States)

    2016-01-01

    An overreliance on foreign oil and the negative impacts of using petroleum fuels on the worlds climate have prompted energy policies that support the diversification of transport fuels and aggressive work to transition to non-petroleum options. Th...

  11. Fuel element replacement and cooling water radioactivity at the Musashi reactor

    International Nuclear Information System (INIS)

    Nozaki, T.; Honda, T.; Horiuchi, N.; Aizawa, O.; Sato, T.

    1988-01-01

    The Musashi reactor (TRIGA-II, 100kW) has been operated without any serious troubles since 1963. In 1985 the old Al-cladded fuel elements were replaced with new stainless cladded ones in order to insure a long and safe operation. By using a semi-automatic equipment the old fuel elements have been transferred into the bulk-shielding experimental pool, which was remodelled for the spent-fuel storage. In order to reduce the exposure during the transfer work, the old fuel elements were cooled in the core tank for 3 months. After the replacement, the radioactivities in the cooling water have been drastically changed. The activity of Na-24 decreased about one decade, and the activities of Cr-51, Mn-54, Mn-56, Co-58 and Co-60 increased about two decades. At this conference we will report on the following points: (1) semi-automatic equipment for the transportation of the Al-cladded spent fuel, (2) structure of spent-fuel storage pool, and (3) radioactivity change in the cooling water. (author)

  12. Review of fuel element development for nuclear rocket engines

    International Nuclear Information System (INIS)

    Taub, J.M.

    1975-06-01

    The Los Alamos Scientific Laboratory (LASL) entered the nuclear propulsion field in 1955 and began work on all aspects of a nuclear propulsion program involving uranium-loaded graphite fuels, hydrogen propellant, and a target exhaust temperature of approximately 2500 0 C. A very extensive uranium-loaded graphite fuel element technology evolved from the program. Selection and composition of raw materials for the extrusion mix had to be coupled with heat treatment studies to give optimum element properties. The highly enriched uranium in the element was incorporated as UO 2 , pyrocarbon-coated UC 2 , or solid solution UC . ZrC particles. An extensive development program resulted in successful NbC or ZrC coatings on elements to withstand hydrogen corrosion at elevated temperatures. Hot gas, thermal shock, thermal stress, and NDT evaluation procedures were developed to monitor progress in preparation of elements with optimum properties. Final evaluation was made in reactor tests at NRDS. Aerojet-General, Westinghouse Astronuclear Laboratory, and the Oak Ridge Y-12 Plant of Union Carbide Nuclear Company entered the program in the early 1960's, and their activities paralleled those of LASL in fuel element development. (U.S.)

  13. Temperature Analysis and Failure Probability of the Fuel Element in HTR-PM

    International Nuclear Information System (INIS)

    Yang Lin; Liu Bing; Tang Chunhe

    2014-01-01

    Spherical fuel element is applied in the 200-MW High Temperature Reactor-Pebble-bed Modular (HTR-PM). Each spherical fuel element contains approximately 12,000 coated fuel particles in the inner graphite matrix with a diameter of 50mm to form the fuel zone, while the outer shell with a thickness of 5mm is a fuel-free zone made up of the same graphite material. Under high burnup irradiation, the temperature of fuel element rises and the stress will result in the damage of fuel element. The purpose of this study is to analyze the temperature of fuel element and to discuss the stress and failure probability. (author)

  14. Fabrication of fuel elements interplay between typical SNR Mark Ia specifications and the fuel element fabrication

    International Nuclear Information System (INIS)

    Biermann, W.K.; Heuvel, H.J.; Pilate, S.; Vanderborck, Y.; Pelckmans, E.; Vanhellemont, G.; Roepenack, H.; Stoll, W.

    1987-01-01

    The core and fuel were designed for the SNR-300 first core by Interatom GmbH and Belgonucleaire. The fuel was fabricated by Alkem/RBU and Belgonucleaire. Based on the preparation of drawings and specifications and on the results of the prerun fabrication, an extensive interplay took place between design requirements, specifications, and fabrication processes at both fuel plants. During start-up of pellet and pin fabrication, this solved such technical questions as /sup 239/Pu equivalent linear weight, pellet density, stoichiometry of the pellets, and impurity content. Close cooperation of designers and manufacturers has allowed manufacture of 205 fuel assemblies without major problems

  15. Neutron physical aspects of the storage of BWR fuel elements

    International Nuclear Information System (INIS)

    Woloch, F.; Sdouz, G.; Suda, M.

    1980-01-01

    For the storage of BWR fuel elements in a high density fuel rack using boronated steel absorbers and in a fuel rack with a larger pitch without absorber, criticality calculations are performed. The cooling water density is varied for the storage without absorbers. For the selected pitches of 16.5 cm for the high density fuel rack and 25 cm for the fuel rack without absorber respectively the ksub(infinitely) values of 0.933 and 0.748 are obtained. The dependence of the results on different calculational methods and on the influence of the variation of three important design parameters, i.e. of the concentration of boron, of the thickness of the boronated steel and of the watergap is investigated for the high density fuel rack. The average isothermal temperature coefficient is obtained for the high density fuel rack as -4.5 x 10 -40 sup(0)C -1 and as approx. 2.0 x 10 -40 sup(0)C -1 for the fuel rack without absorbers. For both ways of storage the aspects of safety of the results are discussed thoroughly. (orig.) 891 RW/orig. 892 CKA [de

  16. Sustainable fuel for the transportation sector

    Science.gov (United States)

    Agrawal, Rakesh; Singh, Navneet R.; Ribeiro, Fabio H.; Delgass, W. Nicholas

    2007-01-01

    A hybrid hydrogen-carbon (H2CAR) process for the production of liquid hydrocarbon fuels is proposed wherein biomass is the carbon source and hydrogen is supplied from carbon-free energy. To implement this concept, a process has been designed to co-feed a biomass gasifier with H2 and CO2 recycled from the H2-CO to liquid conversion reactor. Modeling of this biomass to liquids process has identified several major advantages of the H2CAR process. (i) The land area needed to grow the biomass is CAR process shows the potential to supply the entire United States transportation sector from that quantity of biomass. (iii) The synthesized liquid provides H2 storage in an open loop system. (iv) Reduction to practice of the H2CAR route has the potential to provide the transportation sector for the foreseeable future, using the existing infrastructure. The rationale of using H2 in the H2CAR process is explained by the significantly higher annualized average solar energy conversion efficiency for hydrogen generation versus that for biomass growth. For coal to liquids, the advantage of H2CAR is that there is no additional CO2 release to the atmosphere due to the replacement of petroleum with coal, thus eliminating the need to sequester CO2. PMID:17360377

  17. Core analysis during transition from 37-element fuel to CANFLEX-NU fuel in CANDU 6

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Chang Joon; Suk, Ho Chun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1999-12-31

    An 1200-day time-dependent fuel-management for the transition from 37-element fuel to CANFLEX-NU fuel in a CANDU 6 reactor has been simulated to show the compatibility of the CANFLEX-NU fuel with the reactor operation. The simulation calculations were carried out with the RFSP code, provided by cell averaged fuel properties obtained from the POWDERPUFS-V code. The refueling scheme for both fuels was an eight bundle shift at a time. The simulation results show that the maximum channel and bundle powers were maintained below the license limit of the CANDU 6. This indicates that the CANFLEX-NU fuel bundle is compatible with the CANDU 6 reactor operation during the transition period. 3 refs., 2 figs., 1 tab. (Author)

  18. Core analysis during transition from 37-element fuel to CANFLEX-NU fuel in CANDU 6

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Chang Joon; Suk, Ho Chun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    An 1200-day time-dependent fuel-management for the transition from 37-element fuel to CANFLEX-NU fuel in a CANDU 6 reactor has been simulated to show the compatibility of the CANFLEX-NU fuel with the reactor operation. The simulation calculations were carried out with the RFSP code, provided by cell averaged fuel properties obtained from the POWDERPUFS-V code. The refueling scheme for both fuels was an eight bundle shift at a time. The simulation results show that the maximum channel and bundle powers were maintained below the license limit of the CANDU 6. This indicates that the CANFLEX-NU fuel bundle is compatible with the CANDU 6 reactor operation during the transition period. 3 refs., 2 figs., 1 tab. (Author)

  19. Irradiated nuclear fuel transport from Japan to Europe

    International Nuclear Information System (INIS)

    Kavanagh, M.T.; Shimoyama, S.

    1976-01-01

    Irradiated nuclear fuel has been transported from Japan to Europe since 1969, although U.K. experience goes back almost two decades. Both magnox and oxide fuel have been transported, and the technical requirements associated with each type of fuel are outlined. The specialized ships used by British Nuclear Fuels Limited (BNFL) for this transport are described, as well as the ships being developed for future use in the Japan trade. The ship requirements are related to the regulatory position both in the United Kingdom and internationally, and the Japanese regulatory requirements are described. Finally, specific operational experience of a Japanese reactor operator is described

  20. Design of a transportation cask for irradiated CANDU fuel

    International Nuclear Information System (INIS)

    Nash, K.E.; Gavin, M.E.

    1983-01-01

    A major step in the development of a large-scale transportation system for irradiated CANDU fuel is being made by Ontario Hydro in the design and construction of a demonstration cask by 1988/89. The system being designed is based on dry transportation with the eventual fully developed system providing for dry fuel loading and unloading. Research carried out to date has demonstrated that it is possible to transport irradiated CANDU fuel in a operationally efficient and simple manner without any damage which would prejudice subsequent automated fuel handling

  1. Energy and the transport sector. [For countries with no fossil fuels

    Energy Technology Data Exchange (ETDEWEB)

    Olson, P E

    1979-01-01

    This article describes the current energy situation from both the global viewpoint and the viewpoint of countries with no indigenous sources of fossil fuels. The lack of fossil fuels necessitates a substitution with indigenous sources of energy, where feasible. Long-distance railway transport is a self-evident element in the expanding transport sector. In view of the proven high energy efficiency of electric railway systems, there is every incentive for a more active investment policy in railway electrification. This applies to both medium-distance transportation of freight and passengers and different electric mass transit systems.

  2. Experience of European irradiated fuel transport - the first four hundred tonnes

    International Nuclear Information System (INIS)

    Curtis, H.W.

    1977-01-01

    The paper describes the successful integration of the experience of its three shareholders into an international company providing an irradiated fuel transport service throughout Europe. The experience of transporting more than 400 tonnes of irradiated uranium from fifteen power reactors is used to illustrate the flexibility which the transport organisation requires when the access and handling facilities are different at almost every reactor. Variations in fuel cross sections and lengths of fuel elements used in first generation reactors created the need for first generation flasks with sufficient variants to accommodate all reactor fuels but the trend now is towards standardisation of flasks to perhaps two basic types. Increases in fuel rating have raised the flask shielding and heat dissipation requirements and have influenced the design of later flasks. More stringent criticality acceptance criteria have tended to reduce the flask capacity below the maximum number of elements which could physically be contained. Reprocessing plant acceptance criteria initiated because of the presence of substantial quantities of loose crud released in the flask and the need to transport substantial numbers of failed elements have also reduced the flask capacity. Different modes of transport have been developed to cater for the various limitations on access to reactor sites arising from geographical and routing considerations. The safety record of irradiated fuel transport is examined with explanation of the means whereby this has been achieved. The problems of programming the movement of a pool of flasks for fifteen reactors in eight countries are discussed together with the steps taken to ensure that the service operates fairly to give priority to those reactors with the greatest problems. The transport of European irradiated fuel can be presented as an example of international collaboration which works

  3. Design evaluation of the HTGR fuel element size reduction system

    International Nuclear Information System (INIS)

    Strand, J.B.

    1978-06-01

    A fuel element size reduction system for the ''cold'' pilot plant of the General Atomic HTGR Reference Recycle Facility has been designed and tested. This report is both an evaluation of the design based on results of initial tests and a description of those designs which require completion or modification for hot cell use. 11 figures

  4. Storage frame for long fuel elements for nuclear reactors

    International Nuclear Information System (INIS)

    Ristow, U.; Krainer, F.; Heinz, G.

    1986-01-01

    Vertical shafts with a cross section suitable for the fuel element cross section and made of metal can have corrugations for spacing from one another. These corrugations are machined parallel to the wall surface of the shafts. One thus obtains great accuracy of distancing. (orig./HP) [de

  5. Leakage monitoring equipment of fuel element by delayed neutron method

    International Nuclear Information System (INIS)

    Ji Changsong; Zhang Shulan; Zhang Shuheng

    1999-01-01

    Based on monitoring results of delayed neutrons from reactor first circle water, the leakage of reactor fuel elements is monitored. A monitoring equipment consisted of an array of 3 He proportional counter tubes with 75 s delay has been developed. The neutron detection efficiency of 6.1% is obtained

  6. Fuel element database: developer handbook; Entwicklerhandbuch zur Brennelement-Datenbank

    Energy Technology Data Exchange (ETDEWEB)

    Dragicevic, M [Atominstitut der Oesterreichischen Universitaeten (Austria)

    2004-09-15

    The fuel elements database which was developed for Atomic Institute of the Austrian Universities is described. The software uses standards like HTML, PHP and SQL. For the standard installation freely available software packages such as MySQL database or the PHP interpreter from Apache Software Foundation and Java Script were used. (nevyjel)

  7. Facility for electrochemical dissolution of rejected fuel elements

    International Nuclear Information System (INIS)

    Deniskin, V.P.; Filatov, O.N.; Konovalov, E.A.; Kolesnikov, B.P.; Bukharin, A.D.

    2003-01-01

    A facility for electrochemical dissolution of rejected fuel elements with the stainless steel can and uranium of 90% enrichment is described. The start-adjustment works and trial-commercial tests of the facility are carried out. A s a result its technological parameters are determined [ru

  8. Experimental study of water flow in nuclear fuel elements

    International Nuclear Information System (INIS)

    Rodrigues, Lorena Escriche; Rezende, Hugo Cesar; Mattos, Joao Roberto Loureiro de; Barros Filho, Jose Afonso; Santos, Andre Augusto Campagnole dos

    2013-01-01

    This work aims to develop an experimental methodology for investigating the water flow through rod bundles after spacer grids of nuclear fuel elements of PWR type reactors. Speed profiles, with the device LDV (Laser Doppler Velocimetry), and the pressure drop between two sockets located before and after the spacer grid, using pressure transducers were measured

  9. Competitive strength by rearrangement of fuel element activities

    International Nuclear Information System (INIS)

    Pekarek, H.

    1993-01-01

    The fuel element activities of Siemens AG and Siemens Power Corporation (SPC) were merged, in particular by creation of a world-wide manufacturing network; establishment of priorities in research and development; intensified standardization of products and processes; continued quality improvement by TQM (Total Quality Management), and by fusion of European marketing systems. (orig./DG) [de

  10. Holding device for gas-cooled reactor fuel elements

    International Nuclear Information System (INIS)

    Hensolt, T.

    1980-01-01

    The sheathed fuel elements of the GCFR are inserted with their pedestal in a grid plate arranged below the reactor core and are clamped there. The clamping force as well as the force required for hydraulic holding-down against the flow pressure of the coolant are applied through the differential pressure between inlet and outlet of the coolant. (DG) [de

  11. Method to fabricate block fuel elements for high temperature reactors

    International Nuclear Information System (INIS)

    Hrovat, M.; Rachor, L.

    1977-01-01

    The fabrication of block fuel elements for gas-cooled high temperature reactors can be improved upon by adding 0.2 to 2 wt.% of a hydrocarbon compound to the lubricating mixture prior to pressing. Hexanol or octanol are named as substances. The dimensional accuracy of the block is thus improved. 2 examples illustrate the method. (RW) [de

  12. Method to fabricate block fuel elements for high temperature reactors

    International Nuclear Information System (INIS)

    Hrovat, M.; Rachor, L.

    1978-01-01

    The fabrication of block fuel elements for gas-cooled high temperature reactors can be improved upon by adding 0.2 to 2 wt.% of a hydrocarbon compound to the lubricating mixture prior to pressing. Hexanol or octanol are named as substances. The dimensional accuracy of the block is thus improved. 2 examples illustrate the method. (orig./PW)

  13. Poolside inspection, repair and reconstitution of LWR fuel elements

    International Nuclear Information System (INIS)

    1993-03-01

    The purpose of the meeting was to review the state of the art in the area of poolside inspection, repair and reconstitution of light water fuel elements. In the present publication it appears that techniques of inspection, repair and reconstitution of fuel elements have been developed by fuel suppliers and are now routinely and successfully applied in many countries. For the first time, the subject of control rod poolside examination was dealt with, poolside inspection and repair of a MOX assembly were reported and the inspection and repair of WWER assemblies were examined. Compared to the results of the previous meeting, present developments in the area aim now at reaching better economics, better reliability, reduction of personal doses and waste volume. Thirty-six participants representing twelve countries attended the meeting. Fifteen papers were presented in two sessions. An abstract was prepared for each of these papers. Refs, figs, tabs, diagrams, pictures and photos

  14. Method to produce fuel element blocks for HTR reactors

    International Nuclear Information System (INIS)

    Hrovat, M.; Rachor, L.

    1977-01-01

    The patent claim relates to one partial step of the multi-stage pressing process in the production of fuel elements. A binder resin with a softening point at least 15 0 C but preferably 25-40 0 C above the melting point of the lubricant is proposed. The pressed block is expelled from the forging die in the temperature interval between the melting point of the lubricant and the softening point of the binder resin. The purpose of the invention is that the pressed fuel element blocks are expelled from the machine tool without damage at a pressure low enough to protect the mechanical integrity of the coated fuel particles or fertile particles. (UA) [de

  15. Fuel element for high-temperature nuclear power reactors

    International Nuclear Information System (INIS)

    Schloesser, J.

    1974-01-01

    The fuel element of the HTGR consists of a spherical graphite body with a spherical cavity. A deposit of fissile material, e.g. coated particles of uranium carbide, is fixed to the inner wall using binders. In addition to the fissile material, there are concentric deposits of fertile material, e.g. coated thorium carbide particles. The remaining cavity is filled with a graphite mass, preferably graphite powder, and the filling opening with a graphite stopper. At the beginning of the reactor operation, the fissile material layer provides the whole power. With progressing burn-up, the energy production is taken over by the fertile layer, which provides the heat production until the end of burn-up. Due to the relatively small temperature difference between the outer wall of the outer graphite body and the maximum fuel temperature, the power of the fuel element can be increased. (DG) [de

  16. Heat diffusion in cylindrical fuel elements of water cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Randles, J [Technical Assessments and Services Division, Atomic Energy Establishment, Winfrith, Dorchester, Dorset (United Kingdom)

    1961-09-15

    This report contains a theoretical study of heat diffusion in the cylindrical fuel elements of water reactors. After setting up appropriate boundary conditions on the temperature, the steady state Fourier equation is solved both for a flat and a tilted fission power source. It is shown that source tilting does not have an appreciable effect on the peak fuel temperature while the heat flux to the coolant suffers a circumferential variation of less than a half of that of the fission power. In the last section, the theory is extended to include the effect of a flat, time dependent fission power. The time dependent Fourier equation is solved by means of a Dini series of Bessel functions which is shown to be rapidly convergent. From this series is derived expressions for the fuel element transfer functions required in reactor servo-analysis. These have the form of a rapidly convergent series of time-lag terms. (author)

  17. Modeling and Simulation of a Nuclear Fuel Element Test Section

    Science.gov (United States)

    Moran, Robert P.; Emrich, William

    2011-01-01

    "The Nuclear Thermal Rocket Element Environmental Simulator" test section closely simulates the internal operating conditions of a thermal nuclear rocket. The purpose of testing is to determine the ideal fuel rod characteristics for optimum thermal heat transfer to their hydrogen cooling/working fluid while still maintaining fuel rod structural integrity. Working fluid exhaust temperatures of up to 5,000 degrees Fahrenheit can be encountered. The exhaust gas is rendered inert and massively reduced in temperature for analysis using a combination of water cooling channels and cool N2 gas injectors in the H2-N2 mixer portion of the test section. An extensive thermal fluid analysis was performed in support of the engineering design of the H2-N2 mixer in order to determine the maximum "mass flow rate"-"operating temperature" curve of the fuel elements hydrogen exhaust gas based on the test facilities available cooling N2 mass flow rate as the limiting factor.

  18. Fuel Chemistry Research | Transportation Research | NREL

    Science.gov (United States)

    Fuel Chemistry Research Fuel Chemistry Research Photo of a hand holding a beaker containing a clear oils. Photo by Dennis Schroeder, NREL NREL's fuel chemistry research explores how biofuels, advanced , emissions control catalysts, and infrastructure materials. Results from NREL's fuel chemistry studies feed

  19. Nuclear reactor fuel element with a cluster of parallel fuel pins

    International Nuclear Information System (INIS)

    Macfall, D.; Butterfield, C.E.; Butterfield, R.S.

    1977-01-01

    An improvement of the design of nuclear reactor fuel elements is described and illustrated by the example of a gas-cooled, graphite-moderated nuclear reactor. The fuel element has a cluster of parallel fuel pins with an outer can of structure material and an inner sleeve, as well as tie bars and spacing devices for all of these parts. The fuel element designed according to the invention allows lasy assembling and disassembling before and after use. During use, no relative axial motions are possible; nevertheless, the graphite sleeve is at no time subject to tensile stress: the individual parts are held in position from below by a single holding device. (UWI) [de

  20. MX 8: the next generation high capacity system for the transport of fresh MOX fuel

    International Nuclear Information System (INIS)

    Potelle, F.; Issard, H.

    1998-01-01

    The choice of reprocessing policy was made a long time ago in France, leading to the development of an advanced Pu recycling industry. In 1987, Saint Laurent was the first French reactor to be loaded with fresh MOX fuel. Transnucleaire, then in charge of transport packaging development, created the FS 69 concept, derived from the classical RCC concept for the transport of UO 2 fresh fuel. On the other hand, Cogema, as the main actor in the field of fuel cycle and thus in transport matters, developed the associated security truck and security caisson in order to provide the transport system with the acceptable Physical Protection devices required by French Authorities. As a whole, the security truck and the FS 69 have now been used for more than ten years with a remarkable level of efficiency and safety. Indeed, more than 600 fresh MOX fuel elements have been delivered, without any incident, both regarding safety or fuel integrity requirements. But, as a matter of fact, the replacement of FS 69 transport system is now scheduled for several reasons. First of all, the burnups achieved with UO 2 fuel progressed together with its enrichment within the last ten years, and the MOX 'equivalence' also implies that its Pu content be increased to enhance its reactor performances: from 5.25 % of Pu content today, the MOX fuel will reach 7% tomorrow, and almost 10% the day after tomorrow. Lastly, the reprocessing/recycling policy has been confirmed and amplified, leading to an increasing number of 'moxified' reactors. As a consequence, the French utility (EDF), the fuel designer (Fragema, the joint venture between Framatome and Cogema), the fuel manufacturer (Cogema), and the transporter (Transnucleaire) joined in a specific working group devoted to the development of the MX 8, the next generation high capacity system for the land transport of MOX fuel. (authors)