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Sample records for carbide fuel elements

  1. Study on the performance of fuel elements with carbide and carbide-nitride fuel

    International Nuclear Information System (INIS)

    Golovchenko, Yu.M.; Davydov, E.F.; Maershin, A.A.

    1985-01-01

    Characteristics, test conditions and basic results of material testing of fuel elements with carbide and carbonitride fuel irradiated in the BOR-60 reactor up to 3-10% burn-up at specific power rate of 55-70 kW/m and temperatures of the cladding up to 720 deg C are described. Increase of cladding diameter is stated mainly to result from pressure of swelling fuel. The influence of initial efficient porosity of the fuel on cladding deformation and fuel stoichiometry on steel carbonization is considered. Utilization of carbide and carbonitride fuel at efficient porosity of 20% at the given test modes is shown to ensure their operability up to 10% burn-up

  2. Fabrication of uranium carbide/beryllium carbide/graphite experimental-fuel-element specimens

    International Nuclear Information System (INIS)

    Muenzer, W.A.

    1978-01-01

    A method has been developed for fabricating uranium carbide/beryllium carbide/graphite fuel-element specimens for reactor-core-meltdown studies. The method involves milling and blending the raw materials and densifying the resulting blend by conventional graphite-die hot-pressing techniques. It can be used to fabricate specimens with good physical integrity and material dispersion, with densities of greater than 90% of the theoretical density, and with a uranium carbide particle size of less than 10 μm

  3. Advances in carbide fuel element development for fast reactor application

    International Nuclear Information System (INIS)

    Dienst, W.; Kleykamp, H.; Muehling, G.; Reiser, H.; Steiner, H.; Thuemmler, F.; Wedermeyer, H.; Weimar, P.

    1977-01-01

    The features of the carbide fuel development programme are reviewed and evaluated. Single pin and bundle irradiations are carried out under thermal, epithermal and fast flux conditions, the latter in the DFR and KNK-II reactors. Several fuel concepts in the region of representative SNR clad temperatures are compared by parameter and performance tests. A conservative concept is based on He-bonded 8 mm pins with (U,Pu)C pellets and a smear density of 75% TD, operating at 800 W/cm rod power and burnup to 70 MWd/kg. The preparation of mixed carbide fuels is carried out by carbothermic reduction of the oxides in different methods supported by equivalent carbon content, grain size and phase distribution analysis. The fuel for subassembly performance tests is produced in a pilot plant of 0,5 t/year capacity. Compatibility studies reveal that cladding carburization is the only chemical interaction with carbide fuels. This effect leads to a reduction in ductility of the stainless steel. Fission products apparently play no role in the compatibility behaviour. Comprehensive studies lead to reliable information on the chemical and thermodynamic state of the fuel under irradiation. The swelling of carbide fuels and the fission gas release are examined and analysed. Cladding plastic strain by fuel swelling occurs during steady-state operation because the irradiation creep is rather slow compared to oxide fuels. The cladding strain observed depends on the fuel porosity and the cladding strength. The development of carbide fuel pins is complemented by the application of comprehensive computer models. In addition to the steady-state tests power cycling and safety tests are under performance. Up to 1980 the results are summarized for the final design and specification. The development target of the present program is to fabricate several subassemblies for test operation in the SNR 300 by 1981

  4. Irradiation performance of helium-bonded uranium--plutonium carbide fuel elements

    International Nuclear Information System (INIS)

    Latimer, T.W.; Petty, R.L.; Kerrisk, J.F.; DeMuth, N.S.; Levine, P.J.; Boltax, A.

    1979-01-01

    The current irradiation program of helium-bonded uranium--plutonium carbide elements is achieving its original goals. By August 1978, 15 of the original 171 helium-bonded elements had reached their goal burnups including one that had reached the highest burnup of any uranium--plutonium carbide element in the U.S.--12.4 at.%. A total of 66 elements had attained burnups over 8 at.%. Only one cladding breach had been identified at that time. In addition, the systematic and coordinated approach to the current steady-state irradiation tests is yielding much needed information on the behavior of helium-bonded carbide fuel elements that was not available from the screening tests (1965 to 1974). The use of hyperstoichiometric (U,Pu)C containing approx. 10 vol% (U,Pu) 2 C 3 appears to combine lower swelling with only a slightly greater tendency to carburize the cladding than single-phase (U,Pu)C. The selected designs are providing data on the relationship between the experimental parameters of fuel density, fuel-cladding gap size, and cladding type and various fuel-cladding mechanical interaction mechanisms

  5. Joining elements of silicon carbide

    International Nuclear Information System (INIS)

    Olson, B.A.

    1979-01-01

    A method of joining together at least two silicon carbide elements (e.g.in forming a heat exchanger) is described, comprising subjecting to sufficiently non-oxidizing atmosphere and sufficiently high temperature, material placed in space between the elements. The material consists of silicon carbide particles, carbon and/or a precursor of carbon, and silicon, such that it forms a joint joining together at least two silicon carbide elements. At least one of the elements may contain silicon. (author)

  6. Design and performance of sodium-bonded uranium--plutonium carbide fuel elements

    International Nuclear Information System (INIS)

    Kerrisk, J.F.; DeMuth, N.S.; Petty, R.L.; Latimer, T.W.; Vitti, J.A.; Jones, L.J.

    1979-01-01

    Recent results from irradiation tests indicate that sodium-bonded elements provide a practical advanced fuel element design for use in LMFBRs. Shroud tubes have effectively controlled fuel-cladding mechanical interaction; thicker and stronger claddings have also been effective in this respect. Burnups to 11 at.% have been achieved under typical operating conditions. A hetrogeneous core with a breeding ratio of 1.55 and a compound system doubling time of less than 13 years has been designed using these element designs

  7. Design report for an annular fuel element for accommodation of a carbide test bundle on the ring position of the KNK II/2 test zone

    International Nuclear Information System (INIS)

    Haefner, H.E.

    1982-03-01

    This report describes an annular oxide element with Mark II rods for accommodation of a 19-pin carbide test bundle on position 201 in the test zone of the second core of KNK II as well as its behavior during the period of operation. The ring element comprises within a driver wrapper in three rows of pins 102 fuel pins of 7.6 mm diameter and six structural rods for fixing the spark eroded spacers. The report deals with the ring element with its individual components fuel rod, bundle, wrappers, head and foot and describes methods, criteria and results concerning the design. The carbide test bundle to be accommodated by the annular carrier element will be treated in a separate report. The loadability of the annular element with its components is demonstrated by generally valid standards for strength criteria

  8. Postirradiation results and evaluation of helium-bonded uranium--plutonium carbide fuel elements irradiated in EBR-II. Interim report

    International Nuclear Information System (INIS)

    Latimer, T.W.; Barner, J.O.; Kerrisk, J.F.; Green, J.L.

    1976-02-01

    An evaluation was made of the performance of 74 helium-bonded uranium-plutonium carbide fuel elements that were irradiated in EBR-II at 38-96 kW/m to 2-12 at. percent burnup. Only 38 of these elements have completed postirradiation examination. The higher failure rate found in fuel elements which contained high-density (greater than 95 percent theoretical density) fuel than those which contained low-density (77-91 percent theoretical density) fuel was attributed to the limited ability of the high-density fuel to swell into the void space provided in the fuel element. Increasing cladding thickness and original fuel-cladding gap size were both found to influence the failure rates for elements containing low-density fuel. Lower cladding strain and higher fission-gas release were found in high-burnup fuel elements having smear densities of less than 81 percent. Fission-gas release was usually less than 5 percent for high-density fuel, but increased with burnup to a maximum of 37 percent in low-density fuel. Maximum carburization in elements attaining 5-10 at. percent burnup and clad in Types 304 or 316 stainless steel and Incoloy 800 ranged from 36-80 μm and 38-52 μm, respectively. Strontium and barium were the fission products most frequently found in contact with the cladding but no penetration of the cladding by uranium, plutonium, or fission products was observed

  9. 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

  10. 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

  11. 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)

  12. Fission product phases in irradiated carbide fuels

    International Nuclear Information System (INIS)

    Ewart, F.T.; Sharpe, B.M.; Taylor, R.G.

    1975-09-01

    Oxide fuels have been widely adopted as 'first charge' fuels for demonstration fast reactors. However, because of the improved breeding characteristics, carbides are being investigated in a number of laboratories as possible advanced fuels. Irradiation experiments on uranium and mixed uranium-plutonium carbides have been widely reported but the instances where segregate phases have been found and subjected to electron probe analysis are relatively few. Several observations of such segregate phases have now been made over a period of time and these are collected together in this document. Some seven fuel pins have been examined. Two of the irradiations were in thermal materials testing reactors (MTR); the remainder were experimental assemblies of carbide gas bonded oxycarbide and sodium bonded oxycarbide in the Dounreay Fast Reactor (DFR). All fuel pins completed their irradiation without failure. (author)

  13. Square lattice honeycomb tri-carbide fuels for 50 to 250 KN variable thrust NTP design

    International Nuclear Information System (INIS)

    Anghaie, Samim; Knight, Travis; Gouw, Reza; Furman, Eric

    2001-01-01

    Ultrahigh temperature solid solution of tri-carbide fuels are used to design an ultracompact nuclear thermal rocket generating 950 seconds of specific impulse with scalable thrust level in range of 50 to 250 kilo Newtons. Solid solutions of tri-carbide nuclear fuels such as uranium-zirconium-niobium carbide. UZrNbC, are processed to contain certain mixing ratio between uranium carbide and two stabilizing carbides. Zirconium or niobium in the tri-carbide could be replaced by tantalum or hafnium to provide higher chemical stability in hot hydrogen environment or to provide different nuclear design characteristics. Recent studies have demonstrated the chemical compatibility of tri-carbide fuels with hydrogen propellant for a few to tens of hours of operation at temperatures ranging from 2800 K to 3300 K, respectively. Fuel elements are fabricated from thin tri-carbide wafers that are grooved and locked into a square-lattice honeycomb (SLHC) shape. The hockey puck shaped SLHC fuel elements are stacked up in a grooved graphite tube to form a SLHC fuel assembly. A total of 18 fuel assemblies are arranged circumferentially to form two concentric rings of fuel assemblies with zirconium hydride filling the space between assemblies. For 50 to 250 kilo Newtons thrust operations, the reactor diameter and length including reflectors are 57 cm and 60 cm, respectively. Results of the nuclear design and thermal fluid analyses of the SLHC nuclear thermal propulsion system are presented

  14. 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.)

  15. Behaviour of a VVER-1000 fuel element with boron carbide/steel absorber tested under severe fuel damage conditions in the CORA facility (Results of experiment CORA-W2)

    International Nuclear Information System (INIS)

    Hagen, S.; Hofmann, P.; Noack, V.; Schanz, G.; Schumacher, G.; Sepold, L.

    1994-10-01

    The 'Severe Fuel Damage' (SFD) experiments of the Kernforschungszentrum Karlsruhe (KfK), Federal Republic of Germany, were carried out in the out-of-pile facility 'CORA' as part of the international Severe Fuel Damage (SFD) research. The experimental program was set up to provide information on the failure mechanisms of Light Water Reactor (LWR) fuel elements in a temperature range from 1200 C to 2000 C and in few cases up to 2400 C. Between 1987 and 1992 a total of 17 CORA experiments with two different bundle configurations, i.e. PWR (Pressurized Water Reactor) and BWR (Boiling Water Reactor) bundles were performed. These assemblies represented 'Western-type' fuel elements with the pertinent materials for fuel, cladding, grid spacer, and absorber rod. At the end of the experimental program two VVER-1000 specific tests were run in the CORA facility with identical objectives but with genuine VVER-type materials. The experiments, designated CORA-W1 and CORA-W2 were conducted on February 18, 1993 and April 21, 1993, respectively. Test bundle CORA-W1 was without absorber material whereas CORA-W2 contained one absorber rod (boron carbide/steel). As in the earlier CORA tests the test bundles were subjected to temperature transients of a slow heatup rate in a steam environment. The transient phases of the tests were initiated with a temperature ramp rate of 1 K/s. With these conditions a so-called small-break LOCA was simulated. The temperature escalation due to the exothermal zircon/niobium-steam reaction started at about 1200 C, leading the bundles to maximum temperatures of approximately 1900 C. The thermal response of bundle CORA-W2 is comparable to that of CORA-W1. In test CORA-W2, however, the temperature front moved faster from the top to the bottom compared to test CORA-W1 [de

  16. 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

  17. Irradiation of a 19 pin subassembly with mixed carbide fuel in KNK II

    Science.gov (United States)

    Geithoff, D.; Mühling, G.; Richter, K.

    1992-06-01

    The presentation deals with the fabrication, irradiation and nondestructive postirradiation examinations of LMR fuel pins with mixed (U, Pu)-carbide fuels. The mixed carbide fuel was fabricated by the European Institute of Transuranium Elements using various fabrication procedures. Fuel composition varied therefore in a wide range of tolerances with respect to oxygen and phase content and microstructure. The 19 carbide pins were irradiated in the fast neutron flux of the KNK II reactor to a burn-up of about 7 at% without any failure in the centre of a KNK "carrier element" at a maximum linear rating of 800 W/cm. After dismantling in the Hot Cells of KfK nondestructive examinations were carried out comprising dimensional controls, radiography, γ-scanning and eddy-current testing. The results indicate differences in fuel behaviour with respect to composition of the fuel.

  18. 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

  19. Conceptual design study of LMFBR core with carbide fuel

    International Nuclear Information System (INIS)

    Tezuka, H.; Hojuyama, T.; Osada, H.; Ishii, T.; Hattori, S.; Nishimura, T.

    1987-01-01

    Carbide fuel is a hopeful candidate for demonstration FBR(DFBR) fuel from the plant cost reduction point of view. High thermal conductivity and high heavy metal content of carbide fuel lead to high linear heat rate and high breeding ratio. We have analyzed carbide fuel core characteristics and have clarified the concept of carbide fuel core. By survey calculation, we have obtained a correlation map between core parameters and core characteristics. From the map, we have selected a high efficiency core whose features are better than those of an oxide core, and have obtained reactivity coefficients. The core volume and the reactor fuel inventory are approximately 20% smaller, and the burn-up reactivity loss is 50% smaller compared with the oxide fuel core. These results will reduce the capital cost. The core reactivity coefficients are similar to the conventional oxide DFBR's. Therefore the carbide fuel core is regarded as safe as the oxide core. Except neutron fluence, the carbide fuel core has better nuclear features than the oxide core

  20. Safety research needs for carbide and nitride fueled LMFBR's. Final report

    International Nuclear Information System (INIS)

    Kastenberg, W.E.

    1975-01-01

    The results of a study initiated at UCLA during the academic year 1974--1975 to evaluate and review the potential safety related research needs for carbide and nitride fueled LMFBR's are presented. The tasks included the following: (1) Review Core and primary system designs for any significant differences from oxide fueled reactors, (2) Review carbide (and nitride) fuel element irradiation behavior, (3) Review reactor behavior in postulated accidents, (4) Examine analytical methods of accident analysis to identify major gaps in models and data, and (5) Examine post accident heat removal. (TSS)

  1. 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

  2. Silver diffusion through silicon carbide in microencapsulated nuclear fuels TRISO

    International Nuclear Information System (INIS)

    Cancino T, F.; Lopez H, E.

    2013-10-01

    The silver diffusion through silicon carbide is a challenge that has persisted in the development of microencapsulated fuels TRISO (Tri structural Isotropic) for more than four decades. The silver is known as a strong emitter of gamma radiation, for what is able to diffuse through the ceramic coatings of pyrolytic coal and silicon carbide and to be deposited in the heat exchangers. In this work we carry out a recount about the art state in the topic of the diffusion of Ag through silicon carbide in microencapsulated fuels and we propose the role that the complexities in the grain limit can have this problem. (Author)

  3. Ternary carbide uranium fuels for advanced reactor design applications

    International Nuclear Information System (INIS)

    Knight, Travis; Anghaie, Samim

    1999-01-01

    Solid-solution mixed uranium/refractory metal carbides such as the pseudo-ternary carbide, (U, Zr, Nb)C, hold significant promise for advanced reactor design applications because of their high thermal conductivity and high melting point (typically greater than 3200 K). Additionally, because of their thermochemical stability in a hot-hydrogen environment, pseudo-ternary carbides have been investigated for potential space nuclear power and propulsion applications. However, their stability with regard to sodium and improved resistance to attack by water over uranium carbide portends their usefulness as a fuel for advanced terrestrial reactors. An investigation into processing techniques was conducted in order to produce a series of (U, Zr, Nb)C samples for characterization and testing. Samples with densities ranging from 91% to 95% of theoretical density were produced by cold pressing and sintering the mixed constituent carbides at temperatures as high as 2650 K. (author)

  4. Separation of Nuclear Fuel Surrogates from Silicon Carbide Inert Matrix

    International Nuclear Information System (INIS)

    Baney, Ronald

    2008-01-01

    The objective of this project has been to identify a process for separating transuranic species from silicon carbide (SiC). Silicon carbide has become one of the prime candidates for the matrix in inert matrix fuels, (IMF) being designed to reduce plutonium inventories and the long half-lives actinides through transmutation since complete reaction is not practical it become necessary to separate the non-transmuted materials from the silicon carbide matrix for ultimate reprocessing. This work reports a method for that required process

  5. Gas cooled fast breeder reactors using mixed carbide fuel

    International Nuclear Information System (INIS)

    Kypreos, S.

    1976-09-01

    The fast reactors being developed at the present time use mixed oxide fuel, stainless-steel cladding and liquid sodium as coolant (LMFBR). Theoretical and experimental designing work has also been done in the field of gas-cooled fast breeder reactors. The more advanced carbide fuel offers greater potential for developing fuel systems with doubling times in the range of ten years. The thermohydraulic and physics performance of a GCFR utilising this fuel is assessed. One question to be answered is whether helium is an efficient coolant to be coupled with the carbide fuel while preserving its superior neutronic performance. Also, an assessment of the fuel cycle cost in comparison to oxide fuel is presented. (Auth.)

  6. Nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    Hindle, E.D.

    1981-01-01

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

  7. Nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    Hindle, E.D.

    1984-01-01

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

  8. A review of the breeding potentials of carbide, nitride and oxide fueled LMFBRs and GCFRs

    International Nuclear Information System (INIS)

    Handa, Muneo

    1977-11-01

    The effects of design parameters in large variation on compound system doubling time of large advanced-fueled LMFBR are described on the base of recent U.S. results. The fuel element design by Combustion Engineering Inc. in step-by-step substitution of the initial oxide fuel subassemblies with carbide ones is explained. Breeding characteristics of the oxide-fueled LMFBR and its potential design modifications are expounded. The gas cooled fast breeder program in West Germany and in the United States are briefed. Definitions of the breeding ratio and doubling time in overall fuel cycle are given. (auth.)

  9. Design, Fabrication and Performance of Boron-Carbide Control Elements

    International Nuclear Information System (INIS)

    Brammer, H.A.; Jacobson, J.

    1964-01-01

    A control blade design, incorporating boron-carbide (B 4 C) in stainless-steel tubes, was introduced into service in boiling water reactors in April 1961. Since that time this blade has become the standard reference control element in General Electric boiling-water reactors, replacing the 2% boron-stainless-steel blades previously used. The blades consist of a sheathed, cruciform array of small vertical stainless-steel tubes filled with compácted boron-carbide powder. The boron-carbide powder is confined longitudinally into several independent compartments by swaging over ball bearings located inside the tubes. The development and use of boron-carbide control rods is discussed in five phases: 1. Summary of experience with boron-steel blades and reasons for transition to boron-carbide control; 2. Design of the boron-carbide blade, beginning with developmental experiments, including early measurements performed in the AEC ''Control Rod Material and Development Program'' at the Vallecitos Atomic Laboratory, through a description of the final control blade configuration; 3. Fabrication of the blades and quality control procedures; 4. Results of confirmatory pre-operational mechanical and reactivity testing; and 5. Post-operational experience with the blades, including information on the results of mechanical inspection and reactivity testing after two years of reactor service. (author) [fr

  10. 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.)

  11. 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

  12. Failure analysis of carbide fuels under transient overpower (TOP) conditions

    International Nuclear Information System (INIS)

    Nguyen, D.H.

    1980-06-01

    The failure of carbide fuels in the Fast Test Reactor (FTR) under Transient Overpower (TOP) conditions has been examined. The Beginning-of-Cycle Four (BOC-4) all-oxide base case, at $.50/sec ramp rate was selected as the reference case. A coupling between the advanced fuel performance code UNCLE-T and HCDA Code MELT-IIIA was necessary for the analysis. UNCLE-T was used to determine cladding failure and fuel preconditioning which served as initial conditions for MELT-III calculations. MELT-IIIA determined the time of molten fuel ejection from fuel pin

  13. 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.)

  14. 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

  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. 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

  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. 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

  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. GEN IV: Carbide Fuel Elaboration for the 'Futurix Concepts' experiment

    International Nuclear Information System (INIS)

    Vaudez, Stephane; Riglet-Martial, Chantal; Paret, Laurent; Abonneau, Eric

    2008-01-01

    In order to collect information on the behaviour of the future GFR (Gas Fast Reactor) fuel under fast neutron irradiation, an experimental irradiation program, called 'Futurix-concepts' has been launched at the CEA. The considered concept is a composite material made of a fissile fuel embedded in an inert ceramic matrix. Fissile fuel pellets are made of UPuN or UPuC while ceramics are SiC for the carbide fuel and TiN for the nitride fuel. This paper focuses on the description of the carbide composite fabrication. The UPuC pellets are manufactured using a metallurgical powder process. Fabrication and handling of the fuels are carried out in glove boxes under a nitrogen atmosphere. Carbide fuel is synthesized by carbo-thermic reduction under vacuum of a mixture of actinide oxide and graphitic carbon up to 1550 deg. C. After ball milling, the UPuC powder is pressed to create hexagonal or spherical compacts. They are then sintered up to 1750 deg. C in order to obtain a density of 85 % of the theoretical one. The sintered pellets are inserted into an inert and tight capsule of SiC. In order to control the gap between the fuel and the matrix precisely, the pellets are abraded. The inert matrix is then filled with the pellets and the whole system is sealed by a BRASiC R process at high temperature under a helium atmosphere. Fabrication of the sample to be irradiated was done in 2006 and the irradiation began in May 2007 in the Phenix reactor. This presentation will detail and discuss the results obtained during this fabrication phase. (authors)

  4. Present status of uranium-plutonium mixed carbide fuel development for LMFBRs

    International Nuclear Information System (INIS)

    Handa, Muneo; Suzuki, Yasufumi

    1984-01-01

    The feature of carbide fuel is that it has the doubling time as short as about 13 years, that is, close to one half as compared with oxide fuel. The development of the carbide fuel in the past 10 years has been started in amazement. Especially in the program of new fuel development in USA started in 1974, He and Na bond fuel attained the burnup of 16 a/o without causing the breaking of cladding tubes. In 1984, the irradiation of the assembly composed of 91 fuel pins in the FFTF is expected. On the other hand in Japan, the fuel research laboratory was constructed in 1974 in the Oarai Laboratory, Japan Atomic Energy Research Institute, to carry out the studies on carbide fuel. In the autumn of 1982, two carbide fuel pins with different chemical composition have been successfully made. Accordingly, the recent status of the development is explained. The uranium-plutonium mixed carbide fuel is suitable to liquid metal-cooled fast breeder reactors because of large heat conductivity and the high density of nuclear fission substances. The thermal and nuclear characteristics of carbide fuel, the features of the reactor core using carbide fuel, the chemical and mechanical interaction of fuel and cladding tubes, the selection of bond materials, the manufacturing techniques for the fuel, the development of the analysis code for fuel behavior, and the research and development of carbide fuel in Japan are described. (Kako, I.)

  5. 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

  6. 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.

  7. 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)

  8. 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.)

  9. 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)

  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. 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

  13. 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.

  14. 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

  15. Present status of uranium-plutonium mixed carbide fuel development for LMFBR

    International Nuclear Information System (INIS)

    Handa, Muneo; Suzuki, Yasufumi.

    One Oarai characteristic of a carbide fuel is that its doubling time is about 13 years which is only about half as long as that of an oxide fuel. The development of carbide fuels in the past ten years has been truly remarkable. Especially, through the new fuel development program initiated in 1974 in the United States, success has been achieved with respect to He- and Na-bond fuels in obtaining a 16 a/o burning rate without damage to cladding tubes. In 1984 at FFTF, a radiation of a fuel assembly consisting 91 fuel pins is contemplated. On the other hand, in Japan, in 1974, a Fuel Research Wing specializing in the study of carbide fuels was constructed in the Oarai Laboratory of the Atomic Energy Research Institute and in the fall of 1982, was successful in fabricating two carbide fuel pins having different chemical compositions

  16. 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

  17. 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

  18. 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.)

  19. CANDU reactors with reactor grade plutonium/thorium carbide fuel

    Energy Technology Data Exchange (ETDEWEB)

    Sahin, Suemer [Atilim Univ., Ankara (Turkey). Faculty of Engineering; Khan, Mohammed Javed; Ahmed, Rizwan [Pakistan Institute of Engineering and Applied Sciences, Islamabad (Pakistan); Gazi Univ., Ankara (Turkey). Faculty of Technology

    2011-08-15

    Reactor grade (RG) plutonium, accumulated as nuclear waste of commercial reactors can be re-utilized in CANDU reactors. TRISO type fuel can withstand very high fuel burn ups. On the other hand, carbide fuel would have higher neutronic and thermal performance than oxide fuel. In the present work, RG-PuC/ThC TRISO fuels particles are imbedded body-centered cubic (BCC) in a graphite matrix with a volume fraction of 60%. The fuel compacts conform to the dimensions of sintered CANDU fuel compacts are inserted in 37 zircolay rods to build the fuel zone of a bundle. Investigations have been conducted on a conventional CANDU reactor based on GENTILLYII design with 380 fuel bundles in the core. Three mixed fuel composition have been selected for numerical calculation; (1) 10% RG-PuC + 90% ThC; (2) 30% RG-PuC + 70% ThC; (3) 50% RG-PuC + 50% ThC. Initial reactor criticality values for the modes (1), (2) and (3) are calculated as k{sub {infinity}}{sub ,0} = 1.4848, 1.5756 and 1.627, respectively. Corresponding operation lifetimes are {proportional_to} 2.7, 8.4, and 15 years and with burn ups of {proportional_to} 72 000, 222 000 and 366 000 MW.d/tonne, respectively. Higher initial plutonium charge leads to higher burn ups and longer operation periods. In the course of reactor operation, most of the plutonium will be incinerated. At the end of life, remnants of plutonium isotopes would survive; and few amounts of uranium, americium and curium isotopes would be produced. (orig.)

  20. Survey of post-irradiation examinations made of mixed carbide fuels

    International Nuclear Information System (INIS)

    Coquerelle, M.

    1997-01-01

    Post-irradiation examinations on mixed carbide, nitride and carbonitride fuels irradiated in fast flux reactors Rapsodie and DFR were carried out during the seventies and early eighties. In this report, emphasis was put on the fission gas release, cladding carburization and head-end gaseous oxidation process of these fuels, in particular, of mixed carbides. (author). 8 refs, 16 figs, 3 tabs

  1. 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.

  2. Properties of zirconium carbide for nuclear fuel applications

    Energy Technology Data Exchange (ETDEWEB)

    Katoh, Yutai; Vasudevamurthy, Gokul, E-mail: gvasudev@vcu.edu; Nozawa, Takashi; Snead, Lance L.

    2013-10-15

    Zirconium carbide (ZrC) is a potential coating, oxygen-gettering, or inert matrix material for advanced high temperature reactor fuels. ZrC has demonstrated attractive properties for these fuel applications including excellent resistance against fission product corrosion and fission product retention capabilities. However, fabrication of ZrC results in a range of stable sub-stoichiometric and carbon-rich compositions with or without substantial microstructural inhomogeneity, textural anisotropy, and a phase separation, leading to variations in physical, chemical, thermal, and mechanical properties. The effects of neutron irradiation at elevated temperatures, currently only poorly understood, are believed to be substantially influenced by those compositional and microstructural features further adding complexity to understanding the key ZrC properties. This article provides a survey of properties data for ZrC, as required by the United States Department of Energy’s advanced fuel programs in support of the current efforts toward fuel performance modeling and providing guidance for future research on ZrC for fuel applications.

  3. 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.)

  4. 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)

  5. Determining the minimum required uranium carbide content for HTGR UCO fuel kernels

    International Nuclear Information System (INIS)

    McMurray, Jacob W.; Lindemer, Terrence B.; Brown, Nicholas R.; Reif, Tyler J.; Morris, Robert N.; Hunn, John D.

    2017-01-01

    Highlights: • The minimum required uranium carbide content for HTGR UCO fuel kernels is calculated. • More nuclear and chemical factors have been included for more useful predictions. • The effect of transmutation products, like Pu and Np, on the oxygen distribution is included for the first time. - Abstract: Three important failure mechanisms that must be controlled in high-temperature gas-cooled reactor (HTGR) fuel for certain higher burnup applications are SiC layer rupture, SiC corrosion by CO, and coating compromise from kernel migration. All are related to high CO pressures stemming from O release when uranium present as UO 2 fissions and the O is not subsequently bound by other elements. In the HTGR kernel design, CO buildup from excess O is controlled by the inclusion of additional uranium apart from UO 2 in the form of a carbide, UC x and this fuel form is designated UCO. Here general oxygen balance formulas were developed for calculating the minimum UC x content to ensure negligible CO formation for 15.5% enriched UCO taken to 16.1% actinide burnup. Required input data were obtained from CALPHAD (CALculation of PHAse Diagrams) chemical thermodynamic models and the Serpent 2 reactor physics and depletion analysis tool. The results are intended to be more accurate than previous estimates by including more nuclear and chemical factors, in particular the effect of transmuted Pu and Np oxides on the oxygen distribution as the fuel kernel composition evolves with burnup.

  6. 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)

  7. 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)

  8. 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.)

  9. 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

  10. 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

  11. UK irradiation experience relevant to advanced carbide fuel concepts for LMFBR's

    International Nuclear Information System (INIS)

    Bagley, K.Q.; Batey, W.; Paris, R.; Sloss, W.M.; Snape, G.P.

    1977-01-01

    Despite discouraging prognoses of fabrication and reprocessing problems, it is recognized that the quest for a carbide fuel pin design which fully exploits the favourable density and thermal conductivity of (U,Pu) monocarbide must be maintained. Studies in aid of carbide fuel development have, therefore, continued in the UK in parallel with those on oxide, albeit at a substantially lower level of effort, and a sufficient body of irradiation experience has been accumulated to allow discrimination of realistic fuel pin designs

  12. Advanced Silicon Carbide from Molecular Engineering and Actinide Fuels

    International Nuclear Information System (INIS)

    Meyer, D.J.M.; Garcia, J.; Guillaneux, D.; Wong-Chi-Man, M.; Moreau, J.J.E.

    2008-01-01

    In the frame of nuclear fuels studies for generation IV, carbides or oxycarbides assemblies are one of the engaged material for high temperature reactors. The design of the fuels is not yet defined but some structures are actually considered with SiC as matrix for the actinide fuel. In this work we have studied the synthesis of a multi-scale structure controlled SiC matrix using molecular silicon organometallic precursors. The aim of this work was to develop a way to obtain multi-scale SiC matrix material which could be engineered to fit in any fuel structure defined for generation IV fuels. The control of this multi-scale structure was done using several simulation methods specific of the low temperature solution synthesis of the precursor. In a first step, we have focused our effort on the synthesis of the SiC material. A first level of template was successfully done by the use of solid silica 500 nm balls. A second level of template was studied by the use of meso-porous silica, structured at a 50 nm level. At least, supra-molecular simulation in non aqueous media was considered with the difficulty to build a molecular assembly (inverse micelles). In a second step, we have functionalized the primary silane phase with actinide complexing agent in order to blend directly the actinide inside this primary phase in a controlled way. During these studies, a new one pot synthesis route to obtain the functionalized primary silane phase was developed. (authors)

  13. 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)

  14. 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

  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. 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

  17. 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.)

  18. 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

  19. Characterization of fuel swelling in helium-bonded carbide fuel pins

    International Nuclear Information System (INIS)

    Louie, D.L.Y.

    1987-08-01

    This work is not only the first attempt at characterizing the swelling of (U,Pu)C fuel pellets, but it also represents the only detailed examinations on carbide fuel swelling at high fuel burnups (4 to 16 at. %). This characterization includes the contributions of fission gases, cracks and solid fission products to fuel swelling. Significantly, the contributions of fission gases and cracks were determined by using the image analysis technique (IAT) which allows researchers to take areal measurements of the irradiated fuel porosity and cracks from the photographs of metallographic fuel samples. However, because areal measurements for varying depths in the fuel pellet could not be obtained, the crack areal measurements could not be converted into volumetric quantities. Consequently, in this situation, an areal fuel swelling analysis was used. The macroscopic fission-gas induced fuel swelling (MAS) caused by fission-gas bubbles and pores > 1 μm was determined using the measured irradiated fuel porosity because the measuring range of IAT is limited to bubbles and pores >1 μm. Conversely, for fuel swelling induced by fission-gas bubbles < 1 μm, the microscopic fission-gas induced fuel swelling (MIS) was estimated using an areal fuel swelling model

  20. Porous nuclear fuel element with internal skeleton for high-temperature gas-cooled nuclear reactors

    Science.gov (United States)

    Youchison, Dennis L.; Williams, Brian E.; Benander, Robert E.

    2013-09-03

    Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

  1. Porous nuclear fuel element for high-temperature gas-cooled nuclear reactors

    Science.gov (United States)

    Youchison, Dennis L [Albuquerque, NM; Williams, Brian E [Pacoima, CA; Benander, Robert E [Pacoima, CA

    2011-03-01

    Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

  2. 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)

  3. Pulverization of boron element and proportions of boron carbide in boron

    International Nuclear Information System (INIS)

    Lang, F.M.; Finck, C.

    1956-01-01

    It is possible to reduce boron element into fine powder by means of a mortar and pestle made of sintered boron carbide, the ratio of boron carbide introduced being less than one per cent. Boron element at our disposal is made of sharp edged, dark brown, little grains of average size greater than 5 μ. Grain sizes smaller than 1μ are required for applying thin layers of such boron. (author) [fr

  4. Analysis of possibilities for functional capacity for work rise of reactor fuel elements at nuclear engine regime

    International Nuclear Information System (INIS)

    Deryavko, I.I.; Perepelkin, I.G.; Pivovarov, O.S.; Storozhenko, A.N.; Tarasov, V.I.

    2000-01-01

    The principle results of carbide fuel rods testing during series of IVG.1 reactor starts up at regime simulating nuclear engine regime of nuclear moving power unit are given. Considerable degradation of initial fuel elements status increasing from start up to start up and which could resulted fail of separate technological channels is shown. Origin case of extreme degradation of fuel elements status are insufficient thermal strength of fuel elements operation in the field brittle state of sintered carbide material, Possible ways of artificial reinforce of fuel elements of low temperature sections, increasing its thermal strength up to required level

  5. 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)

  6. Nuclear fuel element end fitting

    International Nuclear Information System (INIS)

    Jabsen, F.S.

    1979-01-01

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

  7. 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.)

  8. 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.)

  9. Mixed Uranium/Refractory Metal Carbide Fuels for High Performance Nuclear Reactors

    International Nuclear Information System (INIS)

    Knight, Travis; Anghaie, Samim

    2002-01-01

    Single phase, solid-solution mixed uranium/refractory metal carbides have been proposed as an advanced nuclear fuel for advanced, high-performance reactors. Earlier studies of mixed carbides focused on uranium and either thorium or plutonium as a fuel for fast breeder reactors enabling shorter doubling owing to the greater fissile atom density. However, the mixed uranium/refractory carbides such as (U, Zr, Nb)C have a lower uranium densities but hold significant promise because of their ultra-high melting points (typically greater than 3700 K), improved material compatibility, and high thermal conductivity approaching that of the metal. Various compositions of (U, Zr, Nb)C were processed with 5% and 10% metal mole fraction of uranium. Stoichiometric samples were processed from the constituent carbide powders, while hypo-stoichiometric samples with carbon-to-metal (C/M) ratios of 0.92 were processed from uranium hydride, graphite, and constituent refractory carbide powders. Processing techniques of cold uniaxial pressing, dynamic magnetic compaction, sintering, and hot pressing were investigated to optimize the processing parameters necessary to produce high density (low porosity), single phase, solid-solution mixed carbide nuclear fuels for testing. This investigation was undertaken to evaluate and characterize the performance of these mixed uranium/refractory metal carbides for high performance, ultra-safe nuclear reactor applications. (authors)

  10. 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

  11. Loss-of-flow transient characterization in carbide-fueled LMFBRs

    International Nuclear Information System (INIS)

    Rothrock, R.B.; Morgan, M.M.; Baars, R.E.; Elson, J.S.; Wray, M.L.

    1985-01-01

    One of the benefits derived from the use of carbide fuel in advanced Liquid Metal Fast Breeder Reactors (LMFBRs) is a decreased vulnerability to certain accidents. This can be achieved through the combination of advanced fuel performance with the enhanced reactivity feedback effects and passive shutdown cooling systems characteristic of the current 'inherently safe' plant concepts. The calculated core response to an unprotected loss of flow (ULOF) accident has frequently been used as a benchmark test of these designs, and the advantages of a high-conductivity fuel in relation to this type of transient have been noted in previous analyses. To evaluate this benefit in carbide-fueled LMFBRs incorporating representative current plant design features, limited calculations have been made of a ULOF transient in a small ('modular') carbide-fueled LMFBR

  12. 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

  13. 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.)

  14. 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)

  15. Development of a Robust Tri-Carbide Fueled Reactor for Multi-Megawatt Space Power and Propulsion Applications

    International Nuclear Information System (INIS)

    Samim Anghaie; Knight, Travis W.; Plancher, Johann; Gouw, Reza

    2004-01-01

    An innovative reactor core design based on advanced, mixed carbide fuels was analyzed for nuclear space power applications. Solid solution, mixed carbide fuels such as (U,Zr,Nb)c and (U,Zr, Ta)C offer great promise as an advanced high temperature fuel for space power reactors

  16. 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)

  17. Role of alloying elements and carbides in the chlorine-induced corrosion of steels and alloys

    Directory of Open Access Journals (Sweden)

    Hans Jürgen Grabke

    2004-03-01

    Full Text Available The high temperature corrosion of steels and Ni-base alloys in oxidizing and chloridizing environments is of practical interest in relation to problems in waste incineration plants and power plants using Cl containing fuels. The behaviour of the most important alloying elements Fe, Cr, Ni, Mo, Mn, Si, Al upon corrosion in an oxidizing and chloridizing atmosphere was elucidated: the reactions and kinetics can be largely understood on the base of thermodynamic data, i.e. free energy of chloride formation, vapor pressure of the chlorides and oxygen pressure pO2 needed for the conversion chlorides -> oxides. The mechanism is described by 'active oxidation', comprising inward penetration of chlorine into the scale, formation of chlorides at the oxide/metal interface, evaporation of the chlorides and conversion of the evaporating chlorides into oxides, which occurs in more or less distance from the surface (depending on pO2. This process leads to loose, fragile, multilayered oxides which are unprotective (therefore: active oxidation. Fe and Cr are rapidly transferred into such scale, Ni and Mo are relatively resistant. In many cases, the grain boundaries of the materials are strongly attacked, this is due to a susceptibility of chromium carbides to chloridation. In contrast the carbides Mo2C, TiC and NbC are less attacked than the matrix. Alloys on the basis Fe-Cr-Si proved to be rather resistant, and the alloying elements Ni and Mo clearly retard the attack in an oxidizing and chloridizing environment.

  18. 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

  19. Calculation of vapour pressures over mixed carbide fuels

    International Nuclear Information System (INIS)

    Joseph, M.; Mathews, C.K.

    1988-01-01

    Vapour pressure over the uranium-plutonium mixed carbide (Usub(l-p) Pusub(p C) was calculated in the temperature range of 1300-9000 for various compositions (p=0.1 to 0.7). Effects of variation of the sesquicarbide content were also studied. The principle of corresponding states was applied to UC and mixed carbides to obtain the equation of state. (author)

  20. 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.)

  1. 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

  2. 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

  3. Technology of the production of breeder fuel elements

    International Nuclear Information System (INIS)

    Funke, P.

    1976-01-01

    A survey is presented of the fabrication of oxide and carbide fuels and of the fuel rod for fast breeders (KNK, SNR-300). The advantages of the chosen methods are explained. The main points of development concerning the oxide fuel rod are gone into. The process sequence for plutonium oxide and plutonium carbide processing is presented in a flow chart. (HR) [de

  4. Estimation of sesqui-carbide fraction for MARK-I fuel

    International Nuclear Information System (INIS)

    Vana Varamban, S.; Ananthasivan, K.

    2016-01-01

    Sesqui-carbide content of FBTR bi-phasic mixed carbide is specified as 5-20 wt.%. For each batch of fuel production, the sesqui-carbide (M2C3) content is being determined by a K-ratio method using XRD information. There is a need to evolve an alternate method for qualitative determination of M2C3 content for a fabricated FBTR fuel pellet. Two independent approaches resulted in a correlation between overall carbon content and the M2C3 phase fraction. The thermodynamic calculations agree well with the stoichiometric correlation between the overall carbon content and the M2C3 phase fraction in FBTR MARK I fuel

  5. 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

  6. 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.)

  7. 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.)

  8. Status of steady-state irradiation testing of mixed-carbide fuel designs

    International Nuclear Information System (INIS)

    Harry, G.R.

    1983-01-01

    The steady-state irradiation program of mixed-carbide fuels has demonstrated clearly the ability of carbide fuel pins to attain peak burnup greater than 12 at.% and peak fluences of 1.4 x 10 23 n/cm 2 (E > 0.1 MeV). Helium-bonded fuel pins in 316SS cladding have achieved peak burnups of 20.7 at.% (192 MWd/kg), and no breaches have occurred in pins of this design. Sodium-bonded fuel pins in 316SS cladding have achieved peak burnups of 15.8 at.% (146 MWd/kg). Breaches have occurred in helium-bonded fuel pins in PE-16 cladding (approx. 5 at.% burnup) and in D21 cladding (approx. 4 at.% burnup). Sodium-bonded fuel pins achieved burnups over 11 at.% in PE-16 cladding and over 6 at.% in D9 and D21 cladding

  9. Design and fuel fabrication processes for the AC-3 mixed-carbide irradiation test

    International Nuclear Information System (INIS)

    Latimer, T.W.; Chidester, K.M.; Stratton, R.W.; Ledergerber, G.; Ingold, F.

    1992-01-01

    The AC-3 test was a cooperative U.S./Swiss irradiation test of 91 wire-wrapped helium-bonded U-20% Pu carbide fuel pins irradiated to 8.3 at % peak burnup in the Fast Flux Test Facility. The test consisted of 25 pins that contained spherepac fuel fabricated by the Paul Scherrer Institute (PSI) and 66 pins that contained pelletized fuel fabricated by the Los Alamos National Laboratory. Design of AC-3 by LANL and PSI was begun in 1981, the fuel pins were fabricated from 1983 to 1985, and the test was irradiated from 1986 to 1988. The principal objective of the AC-3 test was to compare the irradiation performance of mixed-carbide fuel pins that contained either pelletized or sphere-pac fuel at prototypic fluence and burnup levels for a fast breeder reactor

  10. Analysis of refabricated fuel: determination of carbon in uranium plutonium mixed carbide

    International Nuclear Information System (INIS)

    Huwyler, S.

    1977-09-01

    In developing uranium plutonium mixed carbide which represents an advanced fuel for breeder reactors carbon analysis is an important means of determining the stoichiometry. Methods of carbon determination are briefly reviewed. The carbon determination using a LECO WR-12 Carbon Determinator is treated in detail and experience of three years operation communicated. Problems arising from operating the LECO-apparatus in a glove box are discussed. It is pointed out that carbon determination with the LECO-apparatus is a very fast method with good precision and well suited for the routine analysis of mixed carbide fuel. The accuracy of the method is checked by means of a standard. (Auth.)

  11. 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)

  12. 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

  13. Nerva Fuel Element Development Program Summary Report - July 1966 through June 1972 Extrusion Studies

    Energy Technology Data Exchange (ETDEWEB)

    Napier, J. M.

    1973-09-21

    This part of the completion report pertaining to the NERVA graphite fuel element program covers data collected during the extrusion studies. The physical properties of the fuel element reached the following values: coefficient of thermal expansion (CTE) - 7.0 x 10-6/o C (25 - l,OOOo C); modulus of elasticity - 1.5 x lo6 psi; flexural strength - - 8,000 psi; ultimate strain to failure - 5,500 pidin; good thermal stress resistance. Matrices were produced which could be vapor coated with crack-free films of zirconium carbide. The CTE of the matrix was almost equal to the CTE of the zirconium carbide coating.

  14. 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)

  15. 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

  16. Review of the literature for dry reprocessing oxide, metal, and carbide fuel: The AIROX, RAHYD, and CARBOX pyrochemical processes

    Energy Technology Data Exchange (ETDEWEB)

    Hoyt, R.C.; Rhee, B.W. [Rockwell International Corp., Canoga Park, CA (United States). Energy Systems Group

    1979-09-30

    The state of the art of dry processing oxide, carbide, and metal fuel has been determined through an extensive literature review. Dry processing in one of the most proliferation resistant fuel reprocessing technologies available to date, and is one of the few which can be exported to other countries. Feasibility has been established for oxide, carbide, and metal fuel on a laboratory scale, and large-scale experiments on oxide and carbide fuel have shown viability of the dry processing concept. A complete dry processing cycle has been demonstrated by multicycle processing-refabrication-reirradiation experiments on oxide fuel. Additional experimental work is necessary to: (1) demonstrate the complete fuel cycle for carbide and metal fuel, (2) optimize dry processing conditions, and (3) establish fission product behavior. Dry process waste management is easier than for an aqueous processing facility since wastes are primarily solids and gases. Waste treatment can be accomplished by techniques which have been, or are being, developed for aqueous plants.

  17. A high-temperature, short-duration method of fabricating surrogate fuel microkernels for carbide-based TRISO nuclear fuels

    International Nuclear Information System (INIS)

    Vasudevamurthy, G.; Radecka, A.; Massey, C.

    2015-01-01

    High-temperature gas-cooled reactor technology is a frontrunner among generation IV nuclear reactor designs. Among the advanced nuclear fuel forms proposed for these reactors, dispersion-type fuel consisting of microencapsulated uranium di-oxide kernels, popularly known as tri-structural isotropic (TRISO) fuel, has emerged as the fuel form of choice. Generation IV gas-cooled fast reactors offer the benefit of recycling nuclear waste with increased burn-ups in addition to producing the required power and hydrogen. Uranium carbide has shown great potential to replace uranium di-oxide for use in these fast spectrum reactors. Uranium carbide microkernels for fast reactor TRISO fuel have traditionally been fabricated by long-duration carbothermic reduction and sintering of precursor uranium dioxide microkernels produced using sol-gel techniques. These long-duration conversion processes are often plagued by issues such as final product purity and process parameters that are detrimental to minor actinide retention. In this context a relatively simple, high-temperature but relatively quick-rotating electrode arc melting method to fabricate microkernels directly from a feedstock electrode was investigated. The process was demonstrated using surrogate tungsten carbide on account of its easy availability, accessibility and the similarity of its melting point relative to uranium carbide and uranium di-oxide.

  18. A high-temperature, short-duration method of fabricating surrogate fuel microkernels for carbide-based TRISO nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

    Vasudevamurthy, G.; Radecka, A.; Massey, C. [Virginia Commonwealth Univ., Richmond, VA (United States). High Temperature Materials Lab.

    2015-07-01

    High-temperature gas-cooled reactor technology is a frontrunner among generation IV nuclear reactor designs. Among the advanced nuclear fuel forms proposed for these reactors, dispersion-type fuel consisting of microencapsulated uranium di-oxide kernels, popularly known as tri-structural isotropic (TRISO) fuel, has emerged as the fuel form of choice. Generation IV gas-cooled fast reactors offer the benefit of recycling nuclear waste with increased burn-ups in addition to producing the required power and hydrogen. Uranium carbide has shown great potential to replace uranium di-oxide for use in these fast spectrum reactors. Uranium carbide microkernels for fast reactor TRISO fuel have traditionally been fabricated by long-duration carbothermic reduction and sintering of precursor uranium dioxide microkernels produced using sol-gel techniques. These long-duration conversion processes are often plagued by issues such as final product purity and process parameters that are detrimental to minor actinide retention. In this context a relatively simple, high-temperature but relatively quick-rotating electrode arc melting method to fabricate microkernels directly from a feedstock electrode was investigated. The process was demonstrated using surrogate tungsten carbide on account of its easy availability, accessibility and the similarity of its melting point relative to uranium carbide and uranium di-oxide.

  19. 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

  20. 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.)

  1. 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

  2. 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)

  3. Fuel elements assembling for the DON project exponential experience

    International Nuclear Information System (INIS)

    Anca Abati, R. de

    1966-01-01

    It is described the fuel unit used in the DON exponential experience, the manufacturing installments and tools as well as the stages in the fabrication.These 74 elements contain each 19 cartridges loaded with synterized urania, uranium carbide and indium, gold, and manganese probes. They were arranged in calandria-like tubes and the process-tube. This last one containing a cooling liquid simulating the reactor organic. Besides being used in the DON reactor exponential experience they were used in critic essays by the substitution method in the French reactor AQUILON II. (Author) 6 refs

  4. 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.)

  5. 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.)

  6. 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.)

  7. 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)

  8. 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

  9. Performance of a sphere-pac mixed carbide fuel pin irradiated in the Dounreay Fast Reactor (DFR 527/1 experiment)

    International Nuclear Information System (INIS)

    Bischoff, K.; Smith, L.; Stratton, R.W.

    1980-10-01

    The DFR 527/1 experiment was the first irradiation of EIR sphere-pac uranium-plutonium mixed carbide fuel in a fast flux. The experiment has been successfully irradiated to a burn-up of 7.3% FIMA at ratings between 45 and 62 kW m - 1 and clad temperatures between 300 and 600 0 C. Restructuring and elemental redistribution has been found to be similar to the pattern established for pellet type fuel and follows effects seen in earlier sphere-pac carbide tests. Gas release of 12-14% has been measured. A preliminary comparison of radial temperature distribution calculations using a first version of the fuel behaviour modelling code SPECKLE with the actual metallography has been attempted. (Auth.)

  10. 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)

  11. Evaluation of Codisposal Viability for TH/U Carbide (Fort Saint Vrain HTGR) DOE-Owned Fuel

    International Nuclear Information System (INIS)

    Radulescu, H.

    2001-01-01

    There are more than 250 forms of US Department of Energy (DOE)-owned spent nuclear fuel (SNF). Due to the variety of the spent nuclear fuel, the National Spent Nuclear Fuel Program has designated nine representative fuel groups for disposal criticality analyses based on fuel matrix, primary fissile isotope, and enrichment. The Fort Saint Vrain reactor (FSVR) SNF has been designated as the representative fuel for the Th/U carbide fuel group. The FSVR SNF consists of small particles (spheres of the order of 0.5-mm diameter) of thorium carbide or thorium and high-enriched uranium carbide mixture, coated with multiple, thin layers of pyrolytic carbon and silicon carbide, which serve as miniature pressure vessels to contain fission products and the U/Th carbide matrix. The coated particles are bound in a carbonized matrix, which forms fuel rods or ''compacts'' that are loaded into large hexagonal graphite prisms. The graphite prisms (or blocks) are the physical forms that are handled in reactor loading and unloading operations, and which will be loaded into the DOE standardized SNF canisters. The results of the analyses performed will be used to develop waste acceptance criteria. The items that are important to criticality control are identified based on the analysis needs and result sensitivities. Prior to acceptance to fuel from the Th/U carbide fuel group for disposal, the important items for the fuel types that are being considered for disposal under the Th/U carbide fuel group must be demonstrated to satisfy the conditions determined in this report

  12. Evaluation of Codisposal Viability for TH/U Carbide (Fort Saint Vrain HTGR) DOE-Owned Fuel

    Energy Technology Data Exchange (ETDEWEB)

    H. radulescu

    2001-09-28

    There are more than 250 forms of US Department of Energy (DOE)-owned spent nuclear fuel (SNF). Due to the variety of the spent nuclear fuel, the National Spent Nuclear Fuel Program has designated nine representative fuel groups for disposal criticality analyses based on fuel matrix, primary fissile isotope, and enrichment. The Fort Saint Vrain reactor (FSVR) SNF has been designated as the representative fuel for the Th/U carbide fuel group. The FSVR SNF consists of small particles (spheres of the order of 0.5-mm diameter) of thorium carbide or thorium and high-enriched uranium carbide mixture, coated with multiple, thin layers of pyrolytic carbon and silicon carbide, which serve as miniature pressure vessels to contain fission products and the U/Th carbide matrix. The coated particles are bound in a carbonized matrix, which forms fuel rods or ''compacts'' that are loaded into large hexagonal graphite prisms. The graphite prisms (or blocks) are the physical forms that are handled in reactor loading and unloading operations, and which will be loaded into the DOE standardized SNF canisters. The results of the analyses performed will be used to develop waste acceptance criteria. The items that are important to criticality control are identified based on the analysis needs and result sensitivities. Prior to acceptance to fuel from the Th/U carbide fuel group for disposal, the important items for the fuel types that are being considered for disposal under the Th/U carbide fuel group must be demonstrated to satisfy the conditions determined in this report.

  13. 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.)

  14. GEN IV: Carbide Fuel Elaboration for the 'Futurix Concepts' experiment

    Energy Technology Data Exchange (ETDEWEB)

    Vaudez, Stephane; Riglet-Martial, Chantal; Paret, Laurent; Abonneau, Eric [Commissariat a l' Energie Atomique (C.E.A.), Direction de l' Energie Nucleaire, Centre d' Etudes de Cadarache, 13108 Saint Paul lez Durance Cedex (France)

    2008-07-01

    In order to collect information on the behaviour of the future GFR (Gas Fast Reactor) fuel under fast neutron irradiation, an experimental irradiation program, called 'Futurix-concepts' has been launched at the CEA. The considered concept is a composite material made of a fissile fuel embedded in an inert ceramic matrix. Fissile fuel pellets are made of UPuN or UPuC while ceramics are SiC for the carbide fuel and TiN for the nitride fuel. This paper focuses on the description of the carbide composite fabrication. The UPuC pellets are manufactured using a metallurgical powder process. Fabrication and handling of the fuels are carried out in glove boxes under a nitrogen atmosphere. Carbide fuel is synthesized by carbo-thermic reduction under vacuum of a mixture of actinide oxide and graphitic carbon up to 1550 deg. C. After ball milling, the UPuC powder is pressed to create hexagonal or spherical compacts. They are then sintered up to 1750 deg. C in order to obtain a density of 85 % of the theoretical one. The sintered pellets are inserted into an inert and tight capsule of SiC. In order to control the gap between the fuel and the matrix precisely, the pellets are abraded. The inert matrix is then filled with the pellets and the whole system is sealed by a BRASiC{sup R} process at high temperature under a helium atmosphere. Fabrication of the sample to be irradiated was done in 2006 and the irradiation began in May 2007 in the Phenix reactor. This presentation will detail and discuss the results obtained during this fabrication phase. (authors)

  15. Gamma scanning of mixed carbide and oxide fuel pins irradiated in FBTR

    International Nuclear Information System (INIS)

    Jayaraj, V.V.; Padalakshmi, M.; Ulaganathan, T.; Venkiteswaran, C.N.; Divakar, R.; Joseph, Jojo; Bhaduri, A.K.

    2016-01-01

    Fission in nuclear fuels results in a number of fission products that are gamma emitters in the energy range of 100 keV to 3 MeV. The gamma emitting fission products are therefore amenable for detection by gamma detectors. Assessment of the fission product distribution and their migration behavior through gamma scanning is important for characterizing the in reactor behavior of the fuel. Gamma scanning is an important non destructive technique used to evaluate the behavior of irradiated fuels. As a part of Post Irradiation Examinations (PIE), axial gamma scanning has been carried out on selected fuel pins of the FBTR Mark I mixed carbide fuel sub-assemblies and PFBR MOX test fuel sub-assembly irradiated in FBTR. This paper covers the results of gamma scanning and correlation of gamma scanning results with other PIE techniques

  16. Post irradiation examinations of uranium-plutonium mixed carbide fuels irradiated at low linear power rate

    International Nuclear Information System (INIS)

    Maeda, Atsushi; Sasayama, Tatsuo; Iwai, Takashi; Aizawa, Sakuei; Ohwada, Isao; Aizawa, Masao; Ohmichi, Toshihiko; Handa, Muneo

    1988-11-01

    Two pins containing uranium-plutonium carbide fuels which are different in stoichiometry, i.e. (U,Pu)C 1.0 and (U,Pu)C 1.1 , were constructed into a capsule, ICF-37H, and were irradiated in JRR-2 up to 1.0 at % burnup at the linear heat rate of 420 W/cm. After being cooled for about one year, the irradiated capsule was transferred to the Reactor Fuel Examination Facility where the non-destructive examinations of the fuel pins in the β-γ cells and the destructive ones in two α-γ inert gas atmosphere cells were carried out. The release rates of fission gas were low enough, 0.44 % from (U,Pu)C 1.0 fuel pin and 0.09% from (U,Pu)C 1.1 fuel pin, which is reasonable because of the low central temperature of fuel pellets, about 1000 deg C and is estimated that the release is mainly governed by recoil and knock-out mechanisms. Volume swelling of the fuels was observed to be in the range of 1.3 ∼ 1.6 % for carbide fuels below 1000 deg C. Respective open porosities of (U,Pu)C 1.0 and (U,Pu)C 1.1 fuel were 1.3 % and 0.45 %, being in accordance with the release behavior of fission gas. Metallographic observation of the radial sections of pellets showed the increase of pore size and crystal grain size in the center and middle region of (U,Pu)C 1.0 pellets. The chemical interaction between fuel pellets and claddings in the carbide fuels is the penetration of carbon in the fuels to stainless steel tubes. The depth of corrosion layer in inner sides of cladding tubes ranged 10 ∼ 15 μm in the (U,Pu)C 1.0 fuel and 15 #approx #25 μm in the (U,Pu)C 1.1 fuel, which is correlative with the carbon potential of fuels posibly affecting the amount of carbon penetration. (author)

  17. 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.)

  18. 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

  19. High 240Pu FTR/EMC experiments and analysis: Carbide fuel and UO2 blanket subassembly worths

    International Nuclear Information System (INIS)

    Ombrellaro, P.A.

    1977-06-01

    Carbide-plutonium fuel and UO 2 blanket subassembly worth measurements performed at ANL in the EMC/LWR were analyzed. Composition exchange worth calculations were performed for: (a) the replacement of high- 240 Pu fuel composition for low- 240 Pu fuel composition and carbide-plutonium fuel composition, successively, in the center subassembly of the core; (b) the replacement of low- 240 Pu fuel composition for carbide--plutonium fuel composition in one outer driver subassembly; and (c) the replacement of the radial reflector composition with UO 2 blanket composition in one subassembly of the radial reflector. The composition exchange worth calculations were performed in two-dimensional x,y geometry, using diffusion theory and perturbation theory. Each method produces about the same calculated-to-experimental bias factors

  20. 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

  1. 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

  2. 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

  3. 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

  4. High burnup, high power irradiation behavior of helium-bonded mixed carbide fuel pins

    International Nuclear Information System (INIS)

    Levine, P.J.; Nayak, U.P.; Boltax, A.

    1983-01-01

    Large diameter (9.4 mm) helium-bonded mixed carbide fuel pins were successfully irradiated in EBR-II to high burnup (12%) at high power levels (100 kW/m) with peak cladding midwall temperatures of 550 0 C. The wire-wrapped pins were clad with 0.51-mm-thick, 20% cold-worked Type 316 stainless steel and contained hyperstoichiometric (Usub(0.8)Pusub(0.2))C fuel covering the smeared density range from 75-82% TD. Post-irradiation examinations revealed: extensive fuel-cladding mechanical interaction over the entire length of the fuel column, 35% fission gas release at 12% burnup, cladding carburization and fuel restructuring. (orig.)

  5. The compatibility of stainless steels with particles and powders of uranium carbide and low-sulphur UCS fuels

    International Nuclear Information System (INIS)

    Venter, S.

    1978-05-01

    Slightly hyperstoichiometric (U,Pu)C is a potential nuclear fuel for fast breeder reactors. The excess carbon above the stoichiometric amount results in a higher carbon activity in the fuel, and carbon is transferred to the stainless steel cladding, resulting in embrittlement of the cladding. It is with this problem of carbon transfer from the fuel to the cladding that this thesis is concerned. For practical reasons, UC and not (U,Pu)C was used as the fuel. The theory of decarburisation of carbide fuel and the carburisation of stainless steel, the facilities constructed for the project at the Atomic Energy Board, and the experimental techniques used, including preparation of the fuels, are discussed. The effect of a number of variables of uranium carbide fuel on its compatibility behaviour with stainless steels was investigated, as well as the effect om microstructure and type of stainless steel (304, 304 L and 316) on the rate of carburisation. These studies can be briefly summarised under the following headings: powder-particle size; surface oxidation of uranium carbide; preparation temperature of uranium carbide; low sulfur UCS fuels; uranium sulfide and the microstructure and type of steel. The author concludes that: the effect of surface oxidation and particle size must be taken into account when evaluating out-of-pile tests; the possible effects of surface oxidation must be taken into account when considering vibro-compacted carbide fuels; there is no advantage in replacing a fraction of the carbon atoms by sulphur atoms in slightly hyperstoichiometric carbide fuels, and the type and thermo-mechanical treatment of the stainless steel used as cladding material in a fuel pin is not important as far as the rate of carburisation by the fuel is concerned

  6. 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)

  7. Radionuclide Inventories for DOE SNF Waste Stream and Uranium/Thorium Carbide Fuels

    International Nuclear Information System (INIS)

    K.L. Goluoglu

    2000-01-01

    The objective of this calculation is to generate radionuclide inventories for the Department of Energy (DOE) spent nuclear fuel (SNF) waste stream destined for disposal at the potential repository at Yucca Mountain. The scope of this calculation is limited to the calculation of two radionuclide inventories; one for all uranium/thorium carbide fuels in the waste stream and one for the entire waste stream. These inventories will provide input in future screening calculations to be performed by Performance Assessment to determine important radionuclides

  8. 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)

  9. 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)

  10. Evaluation of catalytic properties of tungsten carbide for the anode of microbial fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Rosenbaum, Miriam; Zhao, Feng; Quaas, Marion; Wulff, Harm; Schroeder, Uwe; Scholz, Fritz [Universitaet Greifswald, Institut fuer Biochemie, Felix-Hausdorff-Strasse 4, 17487 Greifswald (Germany)

    2007-07-31

    In this communication we discuss the properties of tungsten carbide, WC, as anodic electrocatalyst for microbial fuel cell application. The electrocatalytic activity of tungsten carbide is evaluated in the light of its preparation procedure, its structural properties as well as the pH and the composition of the anolyte solution and the catalyst load. The activity of the noble-metal-free electrocatalyst towards the oxidation of several common microbial fermentation products (hydrogen, formate, lactate, ethanol) is studied for microbial fuel cell conditions (e.g., pH 5, room temperature and ambient pressure). Current densities of up to 8.8 mA cm{sup -2} are achieved for hydrogen (hydrogen saturated electrolyte solution), and up to 2 mA cm{sup -2} for formate and lactate, respectively. No activity was observed for ethanol electrooxidation. The electrocatalytic activity and chemical stability of tungsten carbide is excellent in acidic to pH neutral potassium chloride electrolyte solutions, whereas higher phosphate concentrations at neutral pH support an oxidative degradation. (author)

  11. 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.

  12. Present status and further objectives of SNR fuel element development

    International Nuclear Information System (INIS)

    Karsten, G.

    Within the scope of the fuel element development program for the fast breeder reactor SNR 300, 500 fuel pins have been irradiated since 1964, 250 of them in fast flux. Results indicate that the maximum nominal target burnup of 90.000 MWd/t of the SNR 300 Mk Ia possibly can be reached. The main problems, which arise from clad swelling and internal corrosion, can be met by special pretreatments of the austenitic stainless steel 1.4970 and a fuel stoichiometry of 1.97. Beyond this target burnup either material property improvements have to be made or burnup reductions have to be accepted. The remaining questions can be answered by the use of the SNR 300 as a test reactor. A further target is the development of a carbide fuel element, which should be very effective in a high power breeder reactor because of its low fissile inventory and high breeding gain. This development program will also be finalized in the SNR 300. (U.S.)

  13. Molybdenum Carbide Synthesis Using Plasmas for Fuel Cells

    Science.gov (United States)

    2013-06-01

    S. A. Hong, I. H. Oh, and S. J. Shin, “Performance and life time analysis of the kW-class PEMFC stack,” Journal of Power Sources, vol. 106, pp. 295...pp. 591–596, 1998. [25] M. Gotz and H. Wendt, “Binary and ternary anode catalyst formulations including the elements W, Sn and Mo for PEMFCs ...and R. C. Urian, “Electrocatalysis of CO Tolerance by Carbon-Supported PtMo Electrocatalysts in PEMFCs ,” Journal of Electrochemical Society, vol

  14. 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

  15. 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

  16. Multi-criteria methodology to design a sodium-cooled carbide-fueled Gen-IV reactor

    International Nuclear Information System (INIS)

    Stauff, N.

    2011-01-01

    Compared with earlier plant designs (Phenix, Super-Phenix, EFR), Gen IV Sodium-cooled Fast Reactor requires improved economics while meeting safety and non-proliferation criteria. Mixed Oxide (U-Pu)O 2 fuels are considered as the reference fuels due to their important and satisfactory feedback experience. However, innovative carbide (U-Pu)C fuels can be considered as serious competitors for a prospective SFR fleet since carbide-fueled SFRs can offer another type of optimization which might overtake on some aspects the oxide fuel technology. The goal of this thesis is to reveal the potentials of carbide by designing an optimum carbide-fueled SFR with competitive features and a naturally safe behavior during transients. For a French nuclear fleet, a 1500 MW(e) break-even core is considered. To do so, a multi-physic approach was developed taking into account neutronics, fuel thermo-mechanics and thermal-hydraulic at a pre-design stage. Simplified modeling with the calculation of global neutronic feedback coefficients and a quasi-static evaluation was developed to estimate the behavior of a core during overpower transients, loss of flow and/or loss of heat removal transients. The breakthrough of this approach is to provide the designer with an overall view of the iterative process, emphasizing the well-suited innovations and the most efficient directions that can improve the SFR design project.This methodology was used to design a core that benefits from the favorable features of carbide fuels. The core developed is a large carbide-fueled SFR with high power density, low fissile inventory, break-even capability and forgiving behaviors during the un-scrammed transients studied that should prevent using expensive mitigate systems. However, the core-peak burnup is unlikely to significantly exceed 100 MWd/kg because of the large swelling of the carbide fuel leading to quick pellet-clad mechanical interaction and the low creep capacity of carbide. Moderate linear power fuel

  17. 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.)

  18. 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.)

  19. 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.

  20. 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.

  1. 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.)

  2. 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.

  3. 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

  4. 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

  5. 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)

  6. 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

  7. 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

  8. 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)

  9. 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

  10. 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

  11. 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

  12. 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

  13. 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

  14. 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)

  15. 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.)

  16. 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.)

  17. 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

  18. 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

  19. 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.

  20. 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

  1. Evaluation of Aluminum-Boron Carbide Neutron Absorbing Materials for Interim Storage of Used Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Lumin [Univ. of Michigan, Ann Arbor, MI (United States). Department of Nuclear Engineering and Radiological Science; Wierschke, Jonathan Brett [Univ. of Michigan, Ann Arbor, MI (United States). Department of Nuclear Engineering and Radiological Science

    2015-04-08

    The objective of this work was to understand the corrosion behavior of Boral® and Bortec® neutron absorbers over long-term deployment in a used nuclear fuel dry cask storage environment. Corrosion effects were accelerated by flowing humidified argon through an autoclave at temperatures up to 570°C. Test results show little corrosion of the aluminum matrix but that boron is leaching out of the samples. Initial tests performed at 400 and 570°C were hampered by reduced flow caused by the rapid build-up of solid deposits in the outlet lines. Analysis of the deposits by XRD shows that the deposits are comprised of boron trioxide and sassolite (H3BO3). The collection of boron- containing compounds in the outlet lines indicated that boron was being released from the samples. Observation of the exposed samples using SEM and optical microscopy show the growth of new phases in the samples. These phases were most prominent in Bortec® samples exposed at 570°C. Samples of Boral® exposed at 570°C showed minimal new phase formation but showed nearly the complete loss of boron carbide particles. Boron carbide loss was also significant in Boral samples at 400°C. However, at 400°C phases similar to those found in Bortec® were observed. The rapid loss of the boron carbide particles in the Boral® is suspected to inhibit the formation of the new secondary phases. However, Material samples in an actual dry cask environment would be exposed to temperatures closer to 300°C and less water than the lowest test. The results from this study conclude that at the temperature and humidity levels present in a dry cask environment, corrosion and boron leaching will have no effect on the performance of Boral® and Bortec® to maintain criticality control.

  2. Evaluation of Aluminum-Boron Carbide Neutron Absorbing Materials for Interim Storage of Used Nuclear Fuel

    International Nuclear Information System (INIS)

    Wang, Lumin; Wierschke, Jonathan Brett

    2015-01-01

    The objective of this work was to understand the corrosion behavior of Boral® and Bortec® neutron absorbers over long-term deployment in a used nuclear fuel dry cask storage environment. Corrosion effects were accelerated by flowing humidified argon through an autoclave at temperatures up to 570°C. Test results show little corrosion of the aluminum matrix but that boron is leaching out of the samples. Initial tests performed at 400 and 570°C were hampered by reduced flow caused by the rapid build-up of solid deposits in the outlet lines. Analysis of the deposits by XRD shows that the deposits are comprised of boron trioxide and sassolite (H 3 BO 3 ). The collection of boron- containing compounds in the outlet lines indicated that boron was being released from the samples. Observation of the exposed samples using SEM and optical microscopy show the growth of new phases in the samples. These phases were most prominent in Bortec® samples exposed at 570°C. Samples of Boral® exposed at 570°C showed minimal new phase formation but showed nearly the complete loss of boron carbide particles. Boron carbide loss was also significant in Boral samples at 400°C. However, at 400°C phases similar to those found in Bortec® were observed. The rapid loss of the boron carbide particles in the Boral® is suspected to inhibit the formation of the new secondary phases. However, Material samples in an actual dry cask environment would be exposed to temperatures closer to 300°C and less water than the lowest test. The results from this study conclude that at the temperature and humidity levels present in a dry cask environment, corrosion and boron leaching will have no effect on the performance of Boral® and Bortec® to maintain criticality control.

  3. 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

  4. 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)

  5. 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

  6. 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)

  7. Device for fracturing silicon-carbide coatings on nuclear-fuel particles

    Science.gov (United States)

    Turner, L.J.; Willey, M.G.; Tiegs, S.M.; Van Cleve, J.E. Jr.

    This invention is a device for fracturing particles. It is designed especially for use in hot cells designed for the handling of radioactive materials. In a typical application, the device is used to fracture a hard silicon-carbide coating present on carbon-matrix microspheres containing nuclear-fuel materials, such as uranium or thorium compounds. To promote remote control and facilitate maintenance, the particle breaker is pneumatically operated and contains no moving parts. It includes means for serially entraining the entrained particles on an anvil housed in a leak-tight chamber. The flow rate of the gas is at a value effecting fracture of the particles; preferably, it is at a value fracturing them into product particulates of fluidizable size. The chamber is provided with an outlet passage whose cross-sectional area decreases in the direction away from the chamber. The outlet is connected tangentially to a vertically oriented vortex-flow separator for recovering the product particulates entrained in the gas outflow from the chamber. The invention can be used on a batch or continuous basis to fracture the silicon-carbide coatings on virtually all of the particles fed thereto.

  8. Method for fracturing silicon-carbide coatings on nuclear-fuel particles

    Science.gov (United States)

    Turner, Lloyd J.; Willey, Melvin G.; Tiegs, Sue M.; Van Cleve, Jr., John E.

    1982-01-01

    This invention is a device for fracturing particles. It is designed especially for use in "hot cells" designed for the handling of radioactive materials. In a typical application, the device is used to fracture a hard silicon-carbide coating present on carbon-matrix microspheres containing nuclear-fuel material, such as uranium or thorium compounds. To promote remote control and facilitate maintenance, the particle breaker is pneumatically operated and contains no moving parts. It includes means for serially entraining the entrained particles on an anvil housed in a leak-tight chamber. The flow rate of the gas is at a value effecting fracture of the particles; preferably, it is at a value fracturing them into product particulates of fluidizable size. The chamber is provided with an outlet passage whose cross-sectional area decreases in the direction away from the chamber. The outlet is connected tangentially to a vertically oriented vortex-flow separator for recovering the product particulates entrained in the gas outflow from the chamber. The invention can be used on a batch or continuous basis to fracture the silicon-carbide coatings on virtually all of the particles fed thereto.

  9. A Review of Carbide Fuel Corrosion for Nuclear Thermal Propulsion Applications

    Science.gov (United States)

    Pelaccio, Dennis G.; El-Genk, Mohamed S.; Butt, Darryl P.

    1994-07-01

    At the operation conditions of interest in nuclear thermal propulsion reactors, carbide materials have been known to exhibit a number of life limiting phenomena. These include the formation of liquid, loss by vaporization, creep and corresponding gas flow restrictions, and local corrosion and fuel structure degradation due to excessive mechanical and/or thermal loading. In addition, the radiation environment in the reactor core can produce a substantial change in its local physical properties, which can produce high thermal stresses and corresponding stress fractures (cracking). Time-temperature history and cyclic operation of the nuclear reactor can also accelerate some of these processes. The University of New Mexico's Institute for Space Nuclear Power Studies, under NASA sponsorship has recently initiated a study to model the complicated hydrogen corrosion process. In support of this effort, an extensive review of the open literature was performed, and a technical expert workshop was conducted. This paper summarizes the results of this review.

  10. A review of carbide fuel corrosion for nuclear thermal propulsion applications

    Energy Technology Data Exchange (ETDEWEB)

    Pelaccio, D.G.; El-Genk, M.S. [Univ. of New Mexico, Albuquerque, NM (United States). Inst. for Space Nuclear Power Studies; Butt, D.P. [Los Alamos National Lab., NM (United States)

    1993-12-01

    At the operation conditions of interest in nuclear thermal propulsion reactors, carbide materials have been known to exhibit a number of life limiting phenomena. These include the formation of liquid, loss by vaporization, creep and corresponding gas flow restrictions, and local corrosion and fuel structure degradation due to excessive mechanical and/or thermal loading. In addition, the radiation environment in the reactor core can produce a substantial change in its local physical properties, which can produce high thermal stresses and corresponding stress fractures (cracking). Time-temperature history and cyclic operation of the nuclear reactor can also accelerate some of these processes. The University of New Mexico`s Institute for Space Nuclear Power Studies, under NASA sponsorship has recently initiated a study to model the complicated hydrogen corrosion process. In support of this effort, an extensive review of the open literature was performed, and a technical expert workshop was conducted. This paper summarizes the results of this review.

  11. Apparatus for surface treatment of U-Pu carbide fuel samples

    International Nuclear Information System (INIS)

    Fukushima, Susumu; Arai, Yasuo; Handa, Muneo; Ohmichi, Toshihiko; Shiozawa, Ken-ichi.

    1979-05-01

    Apparatus has been constructed for treating the surface of U-Pu carbide fuel samples for EPMA. The treatment is to clean off oxide layer on the surface, then coat with an electric-conductive material. The apparatus, safe in handling plutonium, operates as follows. (1) To avoid oxidation of the analyzing surface by oxygen and water in the air, series of cleaning and coating, i.e. ion-etching and ion-coating or ion-etching and vacuum-evaporation is done at the same time in an inert gas atmosphere. (2) Ion-etching is possible on samples embedded in non-electric-conductive and low heat-conductive resin. (3) Since the temperature rise in (2) is negligible, there is no deterioration of the samples. (author)

  12. 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

  13. 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.

  14. 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.

  15. 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)

  16. 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

  17. 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.

  18. Magnesium carbide synthesis from methane and magnesium oxide - a potential methodology for natural gas conversion to premium fuels and chemicals

    Energy Technology Data Exchange (ETDEWEB)

    Diaz, A.F.; Modestino, A.J.; Howard, J.B. [Massachusetts Institute of Technology, Cambridge, MA (United States)] [and others

    1995-12-31

    Diversification of the raw materials base for manufacturing premium fuels and chemicals offers U.S. and international consumers economic and strategic benefits. Extensive reserves of natural gas in the world provide a valuable source of clean gaseous fuel and chemical feedstock. Assuming the availability of suitable conversion processes, natural gas offers the prospect of improving flexibility in liquid fuels and chemicals manufacture, and thus, the opportunity to complement, supplement, or displace petroleum-based production as economic and strategic considerations require. The composition of natural gas varies from reservoir to reservoir but the principal hydrocarbon constituent is always methane (CH{sub 4}). With its high hydrogen-to-carbon ratio, methane has the potential to produce hydrogen or hydrogen-rich products. However, methane is a very chemically stable molecule and, thus, is not readily transformed to other molecules or easily reformed to its elements (H{sub 2} and carbon). In many cases, further research is needed to augment selectivity to desired product(s), increase single-pass conversions, or improve economics (e.g. there have been estimates of $50/bbl or more for liquid products) before the full potential of these methodologies can be realized on a commercial scale. With the trade-off between gas conversion and product selectivity, a major challenge common to many of these technologies is to simultaneously achieve high methane single-pass conversions and high selectivity to desired products. Based on the results of the scoping runs, there appears to be strong indications that a breakthrough has finally been achieved in that synthesis of magnesium carbides from MgO and methane in the arc discharge reactor has been demonstrated.

  19. Thermodynamic studies of thorium carbide fuel preparation and fuel-clad comptability

    International Nuclear Information System (INIS)

    Besmann, T.M.; Beahm, E.C.

    1979-01-01

    The carbothermic reduction of thorium and uranium-thorium dioxide to monocarbide has been assessed. Equilibrium calculations have yielded Th-C-O and U-Th-C-O phase equilibria and (CO) pressures generated during reduction. The (CO) pressures were found to be at least five orders of magnitude greater than any of the other 15 gaseous species considered. This confirms that the monocarbide can successfully be prepared by carbothermic reduction. The chemical compatibility of thorium carbides with the Cr-Fe-Ni content of clad alloys has been thermodynamically avaluated. Solid solutions of 5 > and 5 > and of 7 C 3 > and 7 C 3 > were the principal reaction products. The Cr-Fe-Ni content of 316 stainless steel showed much less reaction product than that for any of the other six alloys considered. (orig.) [de

  20. Review of experimental studies of zirconium carbide coated fuel particles for high temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Minato, Kazuo; Ogawa, Toru; Fukuda, Kousaku

    1995-03-01

    Experimental studies of zirconium carbide(ZrC) coated fuel particles were reviewed from the viewpoints of fuel particle designs, fabrication, characterization, fuel performance, and fission product retentiveness. ZrC is known as a refractory and chemically stable compound, so ZrC is a candidate to replace the silicon carbide(SiC) coating layer of the Triso-coated fuel particles. The irradiation experiments, the postirradiation heating tests, and the out-of-reactor experiments showed that the ZrC layer was less susceptible than the SiC layer to chemical attack by fission products and fuel kernels, and that the ZrC-coated fuel particles performed better than the standard Triso-coated fuel particles at high temperatures, especially above 1600degC. The ZrC-coated fuel particles demonstrated better cesium retention than the standard Triso-coated fuel particles though the ZrC layer showed a less effective barrier to ruthenium than the SiC layer. (author) 51 refs

  1. Pulverization of boron element and proportions of boron carbide in boron; Broyage de bore element et dosage de carbure de bore dans le bore

    Energy Technology Data Exchange (ETDEWEB)

    Lang, F M; Finck, C [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1956-07-01

    It is possible to reduce boron element into fine powder by means of a mortar and pestle made of sintered boron carbide, the ratio of boron carbide introduced being less than one per cent. Boron element at our disposal is made of sharp edged, dark brown, little grains of average size greater than 5 {mu}. Grain sizes smaller than 1{mu} are required for applying thin layers of such boron. (author) [French] Il est possible de pulveriser finement du bore element au moyen de mortier et pilon en carbure de bore fritte, le taux de carbure de bore introduit etant inferieur a 1 pour cent. Le bore element dont nous disposons est constitue de petits grains brun fonce, a aretes vives, de dimension moyenne superieure a 5 {mu}. L'application de ce bore en couches minces demande des grains de dimensions inferieures a 1 {mu}. (aute0008.

  2. 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

  3. 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

  4. 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)

  5. 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

  6. 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)

  7. 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

  8. 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

  9. 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

  10. Formation of scandium carbides and scandium oxycarbide from the elements at high-(P, T) conditions

    International Nuclear Information System (INIS)

    Juarez-Arellano, Erick A.; Winkler, Bjoern; Bayarjargal, Lkhamsuren; Friedrich, Alexandra; Milman, Victor; Kammler, Daniel R.; Clark, Simon M.; Yan Jinyuan; Koch-Mueller, Monika; Schroeder, Florian; Avalos-Borja, Miguel

    2010-01-01

    Synchrotron diffraction experiments with in situ laser heated diamond anvil cells and multi-anvil press synthesis experiments have been performed in order to investigate the reaction of scandium and carbon from the elements at high-(P,T) conditions. It is shown that the reaction is very sensitive to the presence of oxygen. In an oxygen-rich environment the most stable phase is ScO x C y , where for these experiments x=0.39 and y=0.50-0.56. If only a small oxygen contamination is present, we have observed the formation of Sc 3 C 4 , Sc 4 C 3 and a new orthorhombic ScC x phase. All the phases formed at high pressures and temperatures are quenchable. Experimentally determined elastic properties of the scandium carbides are compared to values obtained by density functional theory based calculations. - Graphical Abstract Legend (TOC Figure): Table of Contents Figure Selected images recorded with a MAR345 image plate detector show the reaction of α-Sc and graphite at high-(P,T) conditions. Left: mixture of α-Sc and graphite. Right: recovered sample after laser heated the diamond anvil cell.

  11. The application of the German reg. guides ('elements of calculation') for radioactive discharges via exhaust air and waste water on fuel element fabrication

    International Nuclear Information System (INIS)

    Hille, R.; Rudolph, W.

    1978-01-01

    The fuel element fabricating plants at Hanau are handlung uranium, plutonium and thorium. The process essentially of converting these heavy metals into oxide, carbide or metal compounds. Thereby occur radioactive discharges into the exhaust air and the waste water. The most important pathway for exposure from these substances is inhalation, the released radionuclides mostly being α-emitters. Compared to this the external irradiation from immersion in γ, β, and neutron radiation is of less importance. (orig./HP) [de

  12. 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)

  13. 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

  14. 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)

  15. Evaluation of the mechanical performance of silicon carbide in TRISO fuel at high temperatures

    Energy Technology Data Exchange (ETDEWEB)

    Rohbeck, Nadia, E-mail: nadia.rohbeck@manchester.ac.uk; Xiao, Ping, E-mail: p.xiao@manchester.ac.uk

    2016-09-15

    The HTR design envisions fuel operating temperatures of up to 1000 °C and in case of an accident even 1600 °C are conceivable. To ensure safety in all conditions a thorough understanding of the impact of an extreme temperature environment is necessary. This work assesses the high temperature mechanical performance of the silicon carbide (SiC) layer within the tristructural-isotropic (TRISO) fuel particle as it poses the main barrier against fission product release into the primary circuit. Therefore, simulated fuel was fabricated by fluidized bed chemical vapour deposition; varying the deposition conditions resulted in strongly differing SiC microstructures for the various samples. Subsequently the TRISO particles were annealed in inert atmosphere at temperatures ranging from 1600 °C up to 2200 °C. Scanning electron microscopy and Raman spectroscopy showed that strong disintegration of the SiC layer occurred from 2100 °C onwards, but initial signs of porosity formation were visible already at 1800 °C. Still, the elastic modulus and hardness as measured by nanoindentation were hardly impaired. After annealing stoichiometric SiC coatings showed a reduction in fracture strength as determined by a modified crush test, however the actual annealing temperature from 1600 °C to 2000 °C had no measureable effect. Furthermore, a technique was developed to measure the elastic modulus and hardness in situ up to 500 °C using a high temperature nanoindentation facility. This approach allows conducting tests while the specimen and indenter tip are heated to a specific measurement temperature, thus obtaining reliable values for the temperature dependent mechanical properties of the material. For the SiC layer in TRISO particles it was found that the elastic modulus decreased slightly from room temperature up to 500 °C, whereas the hardness was reduced more severely to approximately half of its ambient temperature value.

  16. Evaluation of the Mechanical Performance of Silicon Carbide in TRISO Fuel at High Temperatures

    International Nuclear Information System (INIS)

    Rohbeck, N.; Xiao, P.

    2014-01-01

    The HTR design envisions fuel operating temperatures of up to 1000°C and in case of an accident even 1600°C are conceivable. To ensure safety in all conditions a thorough understanding of the impact of an extreme temperature environment is necessary. This work assesses the high temperature mechanical performance of the silicon carbide (SiC) layer within the tristructural-isotropic (TRISO) fuel particle as it poses the main barrier against fission product release into the primary circuit. Therefore simulated fuel was fabricated by fluidized bed chemical vapour deposition; varying the deposition conditions resulted in strongly differing SiC microstructures for the various samples. Subsequently the TRISO particles were annealed in inert atmosphere at temperatures ranging from 1600°C up to 2200°C. Scanning electron microscopy and Raman spectroscopy showed that strong disintegration of the SiC layer occurred from 2100°C onwards, but initial signs of porosity formation were visible already at 1800°C. Still, the elastic modulus and hardness as measured by nanoindentation were hardly impaired. After annealing stoichiometric SiC coatings showed a reduction in fracture strength as determined by a modified crush test, however the actual annealing temperature from 1600°C to 2000°C had no measureable effect. Furthermore, a technique was developed to measure the elastic modulus and hardness in-situ up to 500°C using a high temperature nanoindentation facility. This approach allows conducting numerous tests on small sample volumes and thus promises to improve our knowledge of irradiation effects on the mechanical properties. For the SiC layer in TRISO particles it was found that the elastic modulus decreased slightly from room temperature up to 500°C, whereas the hardness was reduced more severely to approximately half of its ambient temperature value. (author)

  17. Evaluation of the mechanical performance of silicon carbide in TRISO fuel at high temperatures

    International Nuclear Information System (INIS)

    Rohbeck, Nadia; Xiao, Ping

    2016-01-01

    The HTR design envisions fuel operating temperatures of up to 1000 °C and in case of an accident even 1600 °C are conceivable. To ensure safety in all conditions a thorough understanding of the impact of an extreme temperature environment is necessary. This work assesses the high temperature mechanical performance of the silicon carbide (SiC) layer within the tristructural-isotropic (TRISO) fuel particle as it poses the main barrier against fission product release into the primary circuit. Therefore, simulated fuel was fabricated by fluidized bed chemical vapour deposition; varying the deposition conditions resulted in strongly differing SiC microstructures for the various samples. Subsequently the TRISO particles were annealed in inert atmosphere at temperatures ranging from 1600 °C up to 2200 °C. Scanning electron microscopy and Raman spectroscopy showed that strong disintegration of the SiC layer occurred from 2100 °C onwards, but initial signs of porosity formation were visible already at 1800 °C. Still, the elastic modulus and hardness as measured by nanoindentation were hardly impaired. After annealing stoichiometric SiC coatings showed a reduction in fracture strength as determined by a modified crush test, however the actual annealing temperature from 1600 °C to 2000 °C had no measureable effect. Furthermore, a technique was developed to measure the elastic modulus and hardness in situ up to 500 °C using a high temperature nanoindentation facility. This approach allows conducting tests while the specimen and indenter tip are heated to a specific measurement temperature, thus obtaining reliable values for the temperature dependent mechanical properties of the material. For the SiC layer in TRISO particles it was found that the elastic modulus decreased slightly from room temperature up to 500 °C, whereas the hardness was reduced more severely to approximately half of its ambient temperature value.

  18. 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

  19. Silver diffusion through silicon carbide in microencapsulated nuclear fuels TRISO; Difusion de plata a traves de carburo de silicio en combustibles nucleares microencapsulados TRISO

    Energy Technology Data Exchange (ETDEWEB)

    Cancino T, F.; Lopez H, E., E-mail: Felix.cancino@cinvestav.edu.mx [IPN, Centro de Investigacion y de Estudios Avanzados, Unidad Saltillo, Av. Industria Metalurgica No. 1062, Col. Ramos Arizpe, 25900 Saltillo, Coahuila (Mexico)

    2013-10-15

    The silver diffusion through silicon carbide is a challenge that has persisted in the development of microencapsulated fuels TRISO (Tri structural Isotropic) for more than four decades. The silver is known as a strong emitter of gamma radiation, for what is able to diffuse through the ceramic coatings of pyrolytic coal and silicon carbide and to be deposited in the heat exchangers. In this work we carry out a recount about the art state in the topic of the diffusion of Ag through silicon carbide in microencapsulated fuels and we propose the role that the complexities in the grain limit can have this problem. (Author)

  20. 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)

  1. 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

  2. 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.)

  3. 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

  4. 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)

  5. 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)

  6. 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

  7. 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)

  8. Study of a Tricarbide Grooved Ring Fuel Element for Nuclear Thermal Propulsion

    Science.gov (United States)

    Taylor, Brian; Emrich, Bill; Tucker, Dennis; Barnes, Marvin; Donders, Nicolas; Benensky, Kelsa

    2018-01-01

    Deep space exploration, especially that of Mars, is on the horizon as the next big challenge for space exploration. Nuclear propulsion, through which high thrust and efficiency can be achieved, is a promising option for decreasing the cost and logistics of such a mission. Work on nuclear thermal engines goes back to the days of the NERVA program. Currently, nuclear thermal propulsion is under development again in various forms to provide a superior propulsion system for deep space exploration. The authors have been working to develop a concept nuclear thermal engine that uses a grooved ring fuel element as an alternative to the traditional hexagonal rod design. The authors are also studying the use of carbide fuels. The concept was developed in order to increase surface area and heat transfer to the propellant. The use of carbides would also raise the operating temperature of the reactor. It is hoped that this could lead to a higher thrust to weight nuclear thermal engine. This paper describes the modeling of neutronics, heat transfer, and fluid dynamics of this alternative nuclear fuel element geometry. Fabrication experiments of grooved rings from carbide refractory metals are also presented along with material characterization and interactions with a hot hydrogen environment. Results of experiments and associated analysis are discussed. The authors demonstrated success in reaching desired densities with some success in material distribution and reaching a solid solution. Future work is needed to improve distribution of material, minimize oxidation during the milling process, and define a fabrication process that will serve for constructing grooved ring fuel rods for large system tests.

  9. Characterisation of nuclear dispersion fuels. The non-destructive examination of silicon carbide by selenium immersion

    Energy Technology Data Exchange (ETDEWEB)

    Ambler, J.F.R.; Ferguson, I.F.

    1974-07-15

    The non-destructive microscopic examination of silicon-carbide-coated spheres containing uranium carbide, which involves immersing the coated spheres in selenium, is particularly suited for the examination of flaws in the coats but it is not possible to measure coating thicknesses by this method. Some coats are found to be opaque and this is related to their porosity. (auth)

  10. Irradiation behaviour of advanced fuel elements for the helium-cooled high temperature reactor (HTR)

    International Nuclear Information System (INIS)

    Nickel, H.

    1990-05-01

    The design of modern HTRs is based on high quality fuel. A research and development programme has demonstrated the satisfactory performance in fuel manufacturing, irradiation testing and accident condition testing of irradiated fuel elements. This report describes the fuel particles with their low-enriched UO 2 kernels and TRISO coating, i.e. a sequence of pyrocarbon, silicon carbide, and pyrocarbon coating layers, as well as the spherical fuel element. Testing was performed in a generic programme satisfying the requirements of both the HTR-MODUL and the HTR 500. With a coating failure fraction less than 2x10 -5 at the 95% confidence level, the results of the irradiation experiments surpassed the design targets. Maximum accident temperatures in small, modular HTRs remain below 1600deg C, even in the case of unrestricted core heatup after depressurization. Here, it was demonstrated that modern TRISO fuels retain all safety-relevant fission products and that the fuel does not suffer irreversible changes. Isothermal heating tests have been extended to 1800deg C to show performance margins. Ramp tests to 2500deg C demonstrate the limits of present fuel materials. A long-term programm is planned to improve the statistical significance of presently available results and to narrow remaining uncertainty limits. (orig.) [de

  11. 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)

  12. 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)

  13. Gas chromatographic determination of Di-n-butyl phosphate in radioactive lean organic solvent of FBTR carbide fuel reprocessing

    International Nuclear Information System (INIS)

    Velavendan, P.; Ganesh, S.; Pandey, N.K.; Kamachi Mudali, U.; Natarajan, R.

    2011-01-01

    In the present work Di-n- butyl phosphate (DBP) a degraded product of Tri-n-butyl phosphate (TBP) formed by acid hydrolysis and radiolysis in the PUREX process was analyzed. Lean organic streams of different fuel burn-up FBTR carbide fuel reprocessing solution was determined by standard Gas Chromatographic technique. The method involves the conversion of non-volatile Di-n-butyl phosphate into volatile and stable derivatives by the action of diazomethane and then determined by Gas Chromatograph (GC). A calibration graph was made for DBP concentration range of 200-2000 ppm with correlation coefficient of 0.99587 and RSD 1.2 %. (author)

  14. 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)

  15. 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.)

  16. 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)

  17. 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

  18. ELASTO-PLASTIC DEFORMATION OF COMPOSITE POWDERS WITH LAYERED CARBON AND CARBIDE-FORMING ELEMENT COATING

    Directory of Open Access Journals (Sweden)

    V. N. Kovalevsky

    2012-01-01

    Full Text Available Coating structure formation under magnetron spraying of titanium and carbon cathodes and combined cathodes, namely cobalt (EP 131 – nickel, tungsten – carbon have been investigated under conditions of carbide separate synthesis within the temperature range of 650–1200 °C. Usage of cobalt and nickel particles as matrix material leads to their rapid thermal expansion under heating during sintering process in the dilatometer. Subsequent plastic deformation of sintered samples provides obtaining a composite powder material that is a composite with framing structure of cobalt, titanium and tungsten carbides in the coatings.

  19. Process for the manufacture of a fuel catalyst made of tungsten carbide for electrochemical fuel cells. Verfahren zur Herstellung eines Brennstoffkatalysators aus Wolframcarbid fuer elektrochemische Brennstoffzellen

    Energy Technology Data Exchange (ETDEWEB)

    Baresel, D.; Gellert, W.; Scharner, P.

    1982-05-19

    The invention refers to a process for the manufacture of a fuel catalyst made of tungsten carbide for the direct generation of electrical energy by the oxidation of hydrogen, formaldehyde or formic acid in electrochemical fuel cells. Tungsten carbide is obtained by carburisation of tungsten or tungsten oxide by carbon monoxide. The steps of the process are as follows: dissolving the commercial-quality tungstic acid in ammonium hydroxide; precipitating the tungstic acid with concentrated hydrochloric acid; drying in a vacuum and then heating to 200/sup 0/C to remove the water of crystallisation forming tungsten trioxide; and mixing the tungsten trioxide with zinc powder and heating to 600/sup 0/C. The zinc oxide is dissolved with hydrochloric acid after cooling. The finely divided tungsten obtained in this way is converted with carbon monoxide in a quartz tube at 700/sup 0/C.

  20. 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)

  1. 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)

  2. 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.

  3. 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

  4. 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

  5. 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.

  6. 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.)

  7. 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

  8. 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

  9. Advanced Characterization Techniques for Silicon Carbide and Pyrocarbon Coatings on Fuel Particles for High Temperature Reactors (HTR)

    Energy Technology Data Exchange (ETDEWEB)

    Basini, V.; Charollais, F. [CEA Cadarache, DEN/DEC/SPUA, BP 1, 13108 St Paul Lez Durance (France); Dugne, O. [CEA Marcoule, DEN/DTEC/SCGS BP 17171 30207 Bagnols sur Ceze (France); Garcia, C. [Laboratoire des Composites Thermostructuraux (LCTS), UMR CNRS 5801, 3 allee de La Boetie, 33600 Pessac (France); Perez, M. [CEA Grenoble DRT/DTH/LTH, 17 rue des Martyrs, 38054 Grenoble cedex 9 (France)

    2008-07-01

    Cea and AREVA NP have engaged an extensive research and development program on HTR (high temperature reactor) fuel. The improving of safety of (very) high temperature reactors (V/HTR) is based on the quality of the fuel particles. This requires a good knowledge of the properties of the four-layers TRISO particles designed to retain the uranium and fission products during irradiation or accident conditions. The aim of this work is to characterize exhaustively the structure and the thermomechanical properties of each unirradiated layer (silicon carbide and pyrocarbon coatings) by electron microscopy (SEM, TEM), selected area electronic diffraction (SEAD), thermo reflectance microscopy and nano-indentation. The long term objective of this study is to define pertinent parameters for fuel performance codes used to better understand the thermomechanical behaviour of the coated particles. (authors)

  10. Synthesis of carbide fuels from nano-structured precursors: impact on carbo-reduction and physico-chemical properties

    International Nuclear Information System (INIS)

    Saravia, Alvaro

    2015-01-01

    The classical way classically used for manufacturing carbide fuels consists of carbo-reducing at high temperature (1600 C) and under primary vacuum a mixture of AnO 2 and graphite powders. These conditions are disadvantageous for the synthesis of mixed (U,Pu)C carbides on account of plutonium volatilization. Therefore, one of the main aims of these studies is to decrease the carbo-reduction temperature. The experiments focused mainly on the lowering of the uranium oxide temperature. This result has been obtained with the use of uranium oxide and carbon nano-structured precursors. To achieve this goal colloidal suspensions of uranium oxide have been prepared and stabilized by cellulosic ethers. Cellulosic ethers are both stabiliser for uranium oxide nanoparticles and carbon source for carbo-reduction. It has been shown that these precursors are more efficient for carbo-reduction than the standard precursors: a reduction of 300 C of carbo-reduction temperature has been obtained. The impact of these precursors on carbo-reduction and on physico-chemical properties as well as the structural and microstructural characterizations of the obtained carbides have been carried out. (author) [fr

  11. 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)

  12. Metallographic examination of damaged N reactor spent nuclear fuel element SFEC5,4378

    Energy Technology Data Exchange (ETDEWEB)

    Marschman, S.C.; Pyecha, T.D.; Abrefah, J.

    1997-08-01

    N-Reactor spent nuclear fuel (SNF) is currently residing underwater in the K Basins at the Hanford site, in Richland, Washington. This report presents results of the metallographic examination of specimens cut from an SNF element (Mark IV-E) with breached cladding. The element had resided in the K-West (KW) Storage Basin for at least 10 years after it was discharged from the N-Reactor. The storage containers in the KW Basin were nominally closed, isolating the SNF elements from the open pool environment. Seven specimens from this Mark IV-E outer fuel element were examined using an optical metallograph. Included were two specimens that had been subjected to a conditioning process recommended by the Independent Technical Assessment Team, two specimens that had been subjected to a conditioning process recommended in the Integrated Process Strategy Report, and three that were in the as-received, as-cut condition. One of the as-received specimens had been cut from the damaged (or breached) end of the element. All other specimens were cut from the undamaged mid-region of the fuel element. The specimens were visually examined to (1) identify uranium hydride inclusions present in the uranium metal fuel, (2) measure the thickness of the oxide layer formed on the uranium edges and assess the apparent integrity and adhesion of the oxide layer, and (3) look for features in the microstructure that might provide an insight into the various corrosion processes that occurred during underwater storage in the KW Basin. These features included, but were not limited to, the integrity of the cladding and the fuel-to-cladding bond, obvious anomalies in the microstructure, excessive pitting or friability of the fuel matrix, and obvious anomalies in the distribution of uranium hydride or uranium carbide inclusions. Also, the observed metallographic features of the conditioned specimens were compared with those of the as-received (unconditioned) specimens. 11 refs., 93 figs., 2 tabs.

  13. 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)

  14. 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.

  15. Current status of the transient integral fuel element performance code URANUS

    International Nuclear Information System (INIS)

    Preusser, T.; Lassmann, K.

    1983-01-01

    To investigate the behavior of fuel pins during normal and off-normal operation, the integral fuel rod code URANUS has been extended to include a transient version. The paper describes the current status of the program system including a presentation of newly developed models for hypothetical accident investigation. The main objective of current development work is to improve the modelling of fuel and clad material behavior during fast transients. URANUS allows detailed analysis of experiments until the onset of strong material transport phenomena. Transient fission gas analysis is carried out due to the coupling with a special version of the LANGZEIT-KURZZEIT-code (KfK). Fuel restructuring and grain growth kinetics models have been improved recently to better characterize pre-experimental steady-state operation; transient models are under development. Extensive verification of the new version has been carried out by comparison with analytical solutions, experimental evidence, and code-to-code evaluation studies. URANUS, with all these improvements, has been successfully applied to difficult fast breeder fuel rod analysis including TOP, LOF, TUCOP, local coolant blockage and specific carbide fuel experiments. Objective of further studies is the description of transient PCMI. It is expected that the results of these developments will contribute significantly to the understanding of fuel element structural behavior during severe transients. (orig.)

  16. 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

  17. 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

  18. 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)

  19. 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)

  20. 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

  1. 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

  2. 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

  3. 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

  4. 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

  5. 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

  6. Fluoride-Salt-Cooled High-Temperature Reactor (FHR) with Silicon-Carbide-Matrix Coated-Particle Fuel

    International Nuclear Information System (INIS)

    Forsberg, C. W.; Snead, Lance Lewis; Katoh, Yutai

    2012-01-01

    The FHR is a new reactor concept that uses coated-particle fuel and a low-pressure liquid-salt coolant. Its neutronics are similar to a high-temperature gas-cooled reactor (HTGR). The power density is 5 to 10 times higher because of the superior cooling properties of liquids versus gases. The leading candidate coolant salt is a mixture of 7 LiF and BeF 2 (FLiBe) possessing a boiling point above 1300 C and the figure of merit ρC p (volumetric heat capacity) for the salt slightly superior to water. Studies are underway to define a near-term base-line concept while understanding longer-term options. Near-term options use graphite-matrix coated-particle fuel where the graphite is both a structural component and the primary neutron moderator. It is the same basic fuel used in HTGRs. The fuel can take several geometric forms with a pebble bed being the leading contender. Recent work on silicon-carbide-matrix (SiCm) coated-particle fuel may create a second longer-term fuel option. SiCm coated-particle fuels are currently being investigated for use in light-water reactors. The replacement of the graphite matrix with a SiCm creates a new family of fuels. The first motivation behind the effort is to take advantage of the superior radiation resistance of SiC compared to graphite in order to provide a stable matrix for hosting coated fuel particles. The second motivation is a much more rugged fuel under accident, repository, and other conditions.

  7. 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)

  8. 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

  9. Highly efficient transition metal and nitrogen co-doped carbide-derived carbon electrocatalysts for anion exchange membrane fuel cells

    Science.gov (United States)

    Ratso, Sander; Kruusenberg, Ivar; Käärik, Maike; Kook, Mati; Puust, Laurits; Saar, Rando; Leis, Jaan; Tammeveski, Kaido

    2018-01-01

    The search for an efficient electrocatalyst for oxygen reduction reaction (ORR) to replace platinum in fuel cell cathode materials is one of the hottest topics in electrocatalysis. Among the many non-noble metal catalysts, metal/nitrogen/carbon composites made by pyrolysis of cheap materials are the most promising with control over the porosity and final structure of the catalyst a crucial point. In this work we show a method of producing a highly active ORR catalyst in alkaline media with a controllable porous structure using titanium carbide derived carbon as a base structure and dicyandiamide along with FeCl3 or CoCl2 as the dopants. The resulting transition metal-nitrogen co-doped carbide derived carbon (M/N/CDC) catalyst is highly efficient for ORR electrocatalysis with the activity in 0.1 M KOH approaching that of commercial 46.1 wt.% Pt/C. The catalyst materials are also investigated by scanning electron microscopy, Raman spectroscopy and X-ray photoelectron spectroscopy to characterise the changes in morphology and composition causing the raise in electrochemical activity. MEA performance of M/N/CDC cathode materials in H2/O2 alkaline membrane fuel cell is tested with the highest power density reached being 80 mW cm-2 compared to 90 mW cm-2 for Pt/C.

  10. 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

  11. 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)

  12. 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.

  13. 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

  14. 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

  15. 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

  16. 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.)

  17. 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

  18. 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

  19. Studies relating to construction materials to be used in different options for head end treatment in reprocessing of mixed carbide fuel of plutonium and uranium

    International Nuclear Information System (INIS)

    Rajan, S.K.; Palamalai, A.; Ravi, T.N.; Sampath, M.; Raman, V.R.; Balasubramanian, G.R.

    1993-01-01

    Mixed carbide of uranium and plutonium has been chosen as the fuel for the first core of Fast Breeder Test Reactor, installed in the Indira Gandhi Centre for Atomic Research. Reprocessing of this fuel is one of the vital steps to prove the viability of the fuel cycle. The head end treatment process introduces constraints in the reprocessing of carbide fuel when compared to the commonly used mixed oxide fuel. Three head end processes, namely direct oxidation, pyrohydrolysis and direct dissolution in nitric acid with oxidation of organic acids were considered for study for exercising the choice. The paper briefly describes the three processes. In each process probable material of construction and related problems are discussed. (author). 3 refs, 5 figs, 7 tabs

  20. Atomic pile Directorate, Department of Metallurgy, Departments of Technology, Department of Fuel Elements and Structures, Division of Study of Fuel Elements - Semi annual report on the 1968-10-1

    International Nuclear Information System (INIS)

    Arnaud, M.; Tortel, J.; Viallet, H.; Marinot, R.; Rulleau, A.; Lestiboudois, G.; Rousseau, G.; Faussat, A.; Ollier, H.; Truffert, J.; Ferrier, C.; Courcon, P.; Rendu, M.; Dieumegard, M.; Bret, A.

    1968-01-01

    This document gathers a set of reports of studies performed on nuclear fuel elements. The addressed topics are: creep behaviour of UMo and UMoAl tubes and pellets under the action of an external pressure (creep strength of tubes under external pressure, creep strength of pellets under external pressure, uncertainties on irradiation parameters in Pegase), problems related to centring devices (measurements and tests), irradiations of ring elements in power reactors, uranium/sheath metallurgical relationship for Bugey and influence of irradiation (cartridge behaviour in Pegase, long duration irradiation in power reactors, extrapolation in Bugey of results obtained in G2), theoretical study of kinetic oxidation phenomena in metal fuels, tests of leaking cartridges in EdF2, evolution of pressure in EL4 type irradiated fuel rods with ZrCu liners with respect to the conductivity integral, a focus on irradiations of Z0 type fuel elements in Pegase, cluster safety tests with uranium carbide in pile and out of pile, a review of studies performed on fuel elements with blowhole, and application of neutrography to fuel elements

  1. A study on the formation of uranium carbide in an induction furnace

    International Nuclear Information System (INIS)

    Song, In Young; Lee, Yoon Sang; Kim, Eung Soo; Lee, Don Bae; Kim, Chang Kyu

    2005-01-01

    Uranium is a typical carbide-forming element. Three carbides, UC, U 2 C 3 and UC 2 , are formed in the uranium-carbon system. The most important of these as fuel is uranium monocarbide UC. It is well known that Uranium carbides can be obtained by three basic methods: 1) by reaction of uranium metal with carbon; 2) by reaction of uranium metal powder with gaseous hydrocarbons; 3) by reaction of uranium oxides with carbon. The use of uranium monocarbide, or materials based on it, has great prospects as fuel for nuclear reactors. It is quite possible that uranium dicarbide UC 2 may also acquire great importance as a fuel, particularly in dispersion fuel elements with graphite matrix. In the present study, uranium carbides are obtained by direct reaction of uranium metal with graphite in a high frequency induction furnace

  2. 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

  3. 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

  4. 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)

  5. 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.)

  6. 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

  7. 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.)

  8. 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)

  9. 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

  10. Study of carbide-forming element interlayers for diamond nucleation and growth on silicon and WC-Co substrates

    International Nuclear Information System (INIS)

    Tang, Y.; Li, Y.S.; Yang, Q.; Hirose, A.

    2010-01-01

    Diamond nucleation and growth on several typical carbide-forming elements (CFE) (Ti, Cr and W) coated Si and WC-Co substrates were studied. The ion beam sputtered CFE interlayers show an amorphous/nanocrystalline microstructure. The diamond formed on the CFE coated substrates shows higher nucleation density and rate and finer grain structure than on uncoated substrates. Consequently, nanocrystalline diamond thin films can be formed on the CFE coated substrates under conventional microcrystalline diamond growth conditions. Among the three tested CFE interlayers, diamond has the highest nucleation density and rate on W layer and the lowest on Ti layer. The diamond nucleation density and rate on CFE coated WC-Co are much higher than those on widely used metal nitride coated WC-Co.

  11. In situ thermal imaging and three-dimensional finite element modeling of tungsten carbide-cobalt during laser deposition

    International Nuclear Information System (INIS)

    Xiong Yuhong; Hofmeister, William H.; Cheng Zhao; Smugeresky, John E.; Lavernia, Enrique J.; Schoenung, Julie M.

    2009-01-01

    Laser deposition is being used for the fabrication of net shapes from a broad range of materials, including tungsten carbide-cobalt (WC-Co) cermets (composites composed of a metallic phase and a hard refractory phase). During deposition, an unusual thermal condition is created for cermets, resulting in rather complex microstructures. To provide a fundamental insight into the evolution of such microstructures, we studied the thermal behavior of WC-Co cermets during laser deposition involving complementary results from in situ high-speed thermal imaging and three-dimensional finite element modeling. The former allowed for the characterization of temperature gradients and cooling rates in the vicinity of the molten pool, whereas the latter allowed for simulation of the entire sample. By combining the two methods, a more robust analysis of the thermal behavior was achieved. The model and the imaging results correlate well with each other and with the alternating sublayers observed in the microstructure.

  12. 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)

  13. Low temperature chemical processing of graphite-clad nuclear fuels

    Science.gov (United States)

    Pierce, Robert A.

    2017-10-17

    A reduced-temperature method for treatment of a fuel element is described. The method includes molten salt treatment of a fuel element with a nitrate salt. The nitrate salt can oxidize the outer graphite matrix of a fuel element. The method can also include reduced temperature degradation of the carbide layer of a fuel element and low temperature solubilization of the fuel in a kernel of a fuel element.

  14. 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.

  15. 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)

  16. Synthesis of Binary Transition Metal Nitrides, Carbides and Borides from the Elements in the Laser-Heated Diamond Anvil Cell and Their Structure-Property Relations

    Directory of Open Access Journals (Sweden)

    Lkhamsuren Bayarjargal

    2011-09-01

    Full Text Available Transition metal nitrides, carbides and borides have a high potential for industrial applications as they not only have a high melting point but are generally harder and less compressible than the pure metals. Here we summarize recent advances in the synthesis of binary transition metal nitrides, carbides and borides focusing on the reaction of the elements at extreme conditions generated within the laser-heated diamond anvil cell. The current knowledge of their structures and high-pressure properties like high-(p; T stability, compressibility and hardness is described as obtained from experiments.

  17. Synthesis of Binary Transition Metal Nitrides, Carbides and Borides from the Elements in the Laser-Heated Diamond Anvil Cell and Their Structure-Property Relations

    Science.gov (United States)

    Friedrich, Alexandra; Winkler, Björn; Juarez-Arellano, Erick A.; Bayarjargal, Lkhamsuren

    2011-01-01

    Transition metal nitrides, carbides and borides have a high potential for industrial applications as they not only have a high melting point but are generally harder and less compressible than the pure metals. Here we summarize recent advances in the synthesis of binary transition metal nitrides, carbides and borides focusing on the reaction of the elements at extreme conditions generated within the laser-heated diamond anvil cell. The current knowledge of their structures and high-pressure properties like high-(p,T) stability, compressibility and hardness is described as obtained from experiments. PMID:28824101

  18. 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

  19. The significance of strength of silicon carbide for the mechanical integrity of coated fuel particles for HTRs

    International Nuclear Information System (INIS)

    Bongartz, K.; Scheer, A.; Schuster, H.; Taeuber, K.

    1975-01-01

    Silicon carbide (SiC) and pyrocarbon are used as coating material for the HTR fuel particles. The PyC shell having a certain strength acts as a pressure vessel for the fission gases whereas the SiC shell has to retain the solid fission products in the fuel kernel. For measuring the strength of coating material the so-called Brittle Ring Test was developed. Strength and Young's modulus can be measured simultaneously with this method on SiC or PyC rings prepared out of the coating material of real fuel particles. The strength measured on the ring under a certain stress distribution which is characteristic for this method is transformed with the aid of the Weibull formalism for brittle fracture into the equivalent strength of the spherical coating shell on the fuel particle under uniform stress caused by the fission gas pressure. The values measured for the strength of the SiC were high (400-700MN/m 2 ), it could therefore be assumed that a SiC layer might contribute significantly also to the mechanical strength of the fuel coating. This assumption was confirmed by an irradiation test on coated particles with PyC-SiC-PyC coatings. There were several particles with all PyC layers broken during the irradiation, whereas the SiC layers remained intact having to withstand the fission gas pressure alone. This fact can only be explained assuming that the strength of the SiC is within the range of the values measured with the brittle ring test. The result indicates that, in optimising the coating of a fuel particle, the PyC layers of a multilayer coating should be considered alone as prospective layers for the SiC. The SiC shell, besides acting as a fission product barrier, is then also responsible for the mechanical integrity of the particle

  20. 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

  1. 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

  2. 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)

  3. 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

  4. 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

  5. Corrosion resistant cemented carbide

    International Nuclear Information System (INIS)

    Hong, J.

    1990-01-01

    This paper describes a corrosion resistant cemented carbide composite. It comprises: a granular tungsten carbide phase, a semi-continuous solid solution carbide phase extending closely adjacent at least a portion of the grains of tungsten carbide for enhancing corrosion resistance, and a substantially continuous metal binder phase. The cemented carbide composite consisting essentially of an effective amount of an anti-corrosion additive, from about 4 to about 16 percent by weight metal binder phase, and with the remaining portion being from about 84 to about 96 percent by weight metal carbide wherein the metal carbide consists essentially of from about 4 to about 30 percent by weight of a transition metal carbide or mixtures thereof selected from Group IVB and of the Periodic Table of Elements and from about 70 to about 96 percent tungsten carbide. The metal binder phase consists essentially of nickel and from about 10 to about 25 percent by weight chromium, the effective amount of an anti-corrosion additive being selected from the group consisting essentially of copper, silver, tine and combinations thereof

  6. 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

  7. 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

  8. 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

  9. 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)

  10. 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

  11. Fabrication and control of fuels made of mixed carbides (U, Pu)C

    International Nuclear Information System (INIS)

    Lorenzelli, R.; Delaroche, P.

    1980-01-01

    Fabrication of this type of advanced fuel is described. The fuel is prepared by reduction of oxides with carbon and natural sintering. Density, thermal stability and thermal conductibility are more particularly studied [fr

  12. 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)

  13. 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)

  14. 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.)

  15. 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

  16. 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)

  17. 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

  18. Universal high-temperature heat treatment furnace for FBR mixed uranium and plutonium carbide fuel

    International Nuclear Information System (INIS)

    Handa, Muneo; Takahashi, Ichiro; Watanabe, Hitoshi

    1978-10-01

    A universal high-temperature heat treatment furnace for LMFBR advanced fuels was installed in Plutonium Fuel Laboratory, Oarai Research Establishment. Design, construction and performance of the apparatus are described. With the apparatus, heat treatment of the fuel under a controlled gas atmosphere and quenching of the fuel with blowing helium gas are possible. Equipment to measure impurity gas release of the fuel is also provided. Various plutonium enclosure techniques, e.g., a gas line filter with new exchange mechanics, have been developed. In performance test, results of the enclosure techniques are described. (author)

  19. 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)

  20. 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.)

  1. 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)

  2. 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)

  3. 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

  4. 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

  5. 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)

  6. 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

  7. 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

  8. 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

  9. 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)

  10. 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

  11. 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

  12. Synthesis and characterization of nanostructured titanium carbide for fuel cell applications

    Energy Technology Data Exchange (ETDEWEB)

    Singh, Paviter; Singh, Harwinder; Singh, Bikramjeet; Kaur, Manpreet; Kaur, Gurpreet; Kumar, Akshay, E-mail: akshaykumar.tiet@gmail.com [Advanced Functional Material Laboratory, Department of Nanotechnology,, Sri Guru Granth Sahib World University, Fatehgarh Sahib-140 406 Punjab (India); Kumar, Manjeet [Department of Materials Engineering, Defense Institute of Advanced Technology (DU), Pune-411 025 (India); Bala, Rajni [Department of Mathematics Punjabi University Patiala-147 002 Punjab (India)

    2016-04-13

    Titanium carbide (TiC) nanoparticles have been successfully synthesized by carbo-thermic reaction of titanium and acetone at 800 °C. This method is relatively low temperature synthesis route. It can be used for large scale production of TiC. The synthesized nanoparticles have been characterized by X-ray diffraction (XRD), scanning electron microscopy (SEM) and differential thermal analyzer (DTA) techniques. XRD analysis confirmed the formation of single phase TiC. XRD analysis confirmed that the particles are spherical in shape with an average particle size of 13 nm. DTA analysis shows that the phase is stable upto 900 °C and the material can be used for high temperature applications.

  13. 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)

  14. 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

  15. 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)

  16. 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

  17. Innovative coating of nanostructured vanadium carbide on the F/M cladding tube inner surface for mitigating the fuel cladding chemical interactions

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Yong [Univ. of Florida, Gainesville, FL (United States); Phillpot, Simon [Univ. of Florida, Gainesville, FL (United States)

    2017-11-29

    500 oC, respectively. The coating layer contains both carbon and vanadium elements as quantified by WED, and the phases mainly consist of a mixture of V2C and VC, which was confirmed using X-ray diffraction patterns. In addition, the ratio between V and C varies with processing temperature, and it was observed that a higher temperature promotes the carbon adsorption and increases thickness of the coating. With optimized deposition conditions, we can apply the coating technique toward the actual T91 cladding materials, and provide the possibilities for the real application in sodium-cooled fast reactors (SFRs). Diffusion couple experiments were performed at both 550 oC and 690 oC, which corresponds to normal and aggressive operating temperatures, respectively. The results show that vanadium carbide coating with wider thickness (8 µm) and lower carbon concentration (27 at.%) reduced the width of the inter diffusion region, indicating that vanadium carbide coating can mitigate FCCI effectively. In specific, inter-diffusion between Fe and Ce was prohibited over most area, but Ce diffusion occurred toward the coating and the Fe substrate through thinner coating layer, which needs further optimization in terms of uniform coating thickness. Overall, it is concluded that this coating process can be successfully applied onto the inner surface of HT9 cladding tubes and the FCCI can be effectively mitigated if not totally eliminated.

  18. 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)

  19. 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)

  20. 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

  1. 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)

  2. 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

  3. 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)

  4. 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)

  5. 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

  6. 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

  7. 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)

  8. 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)

  9. 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)

  10. 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

  11. 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)

  12. 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.

  13. 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.

  14. Finite Element Analysis Modeling of Chemical Vapor Deposition of Silicon Carbide

    Science.gov (United States)

    2014-06-19

    concentrations. This is the method by which species adsorb to the surface of the substrate. The movement resulting from diffusion is governed by...itself. This can be treacherous, however. The mesh is what the entire finite element method is built upon. If the movement of the backbone has... Brownian Motion Algorithm for Tow Scale Modeling of Chemical Vapor Infiltration. Computational Materials Science, 1871-1878. !178 23. Wang, C. & D

  15. 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

  16. 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.)

  17. 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)

  18. 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)

  19. 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)

  20. 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

  1. 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.

  2. 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)

  3. 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.

  4. 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

  5. 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

  6. 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)

  7. 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)

  8. Thermal-Hydraulic Aspects of Changing the Nuclear Fuel-Cladding Materials from Zircaloy to Silicon Carbides

    International Nuclear Information System (INIS)

    Niceno, Bojan; Pouchon, Manuel

    2014-01-01

    The accident in Fukushima has drastically shown the drawbacks of Zircaloy claddings despite their beneficial properties in normal use. The effect of the lack of cooling and the production of hydrogen would not have been so strong if the fuel cladding had not consisted of a zirconium (or metal) alloy. International activities have been started to search for an alternative to Zircaloy, however, still on a limited basis. A project sponsored by Swissnuclear has been conducted at Paul Scherrer Institute (PSI) with the aim to close the gap in knowledge on application of silicon carbides (SiC) as potential replacement for Zircaloys as material for nuclear fuel cladding. The work was interdisciplinary, result of collaboration between different laboratories at PSI, and has focused on SiC cladding material properties, implication of its usage on neutronics and on thermal-hydraulics. This paper summarizes thermal-hydraulic aspects of changing Zircaloy for SiC as the cladding material. The change of cladding material inevitably changes the surface properties thus making a significant impact on boiling curve, and critical heat flux (CHF). Low chemical reactivity of SiC means fewer particles in the flow (less crud), which leads to fewer failures, but also decreases the CHF. Due to differences in physical properties between SiC and Zircaloys, higher brittleness of SiC in particular, might have impact on fuel-rod assembly design, which has direct influence on flow patterns and heat transfer in the fuel assembly. Higher melting (i.e. decomposition) point for SiC means that severe accident management guidelines (SAMG) should have to be re-assessed. Not only would the core degrade later than in the case of conventional fuels, but the production of hydrogen would be quite different as well. All these issues are explored in this work in two steps; first the SiC properties which may have influence on thermal-hydraulics are outlined, then each thermal-hydraulic issues is explained from

  9. 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

  10. 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

  11. 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)

  12. 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

  13. Methanol electro-oxidation on platinum modified tungsten carbides in direct methanol fuel cells: a DFT study.

    Science.gov (United States)

    Sheng, Tian; Lin, Xiao; Chen, Zhao-Yang; Hu, P; Sun, Shi-Gang; Chu, You-Qun; Ma, Chun-An; Lin, Wen-Feng

    2015-10-14

    In exploration of low-cost electrocatalysts for direct methanol fuel cells (DMFCs), Pt modified tungsten carbide (WC) materials are found to be great potential candidates for decreasing Pt usage whilst exhibiting satisfactory reactivity. In this work, the mechanisms, onset potentials and activity for electrooxidation of methanol were studied on a series of Pt-modified WC catalysts where the bare W-terminated WC(0001) substrate was employed. In the surface energy calculations of a series of Pt-modified WC models, we found that the feasible structures are mono- and bi-layer Pt-modified WCs. The tri-layer Pt-modified WC model is not thermodynamically stable where the top layer Pt atoms tend to accumulate and form particles or clusters rather than being dispersed as a layer. We further calculated the mechanisms of methanol oxidation on the feasible models via methanol dehydrogenation to CO involving C-H and O-H bonds dissociating subsequently, and further CO oxidation with the C-O bond association. The onset potentials for the oxidation reactions over the Pt-modified WC catalysts were determined thermodynamically by water dissociation to surface OH* species. The activities of these Pt-modified WC catalysts were estimated from the calculated kinetic data. It has been found that the bi-layer Pt-modified WC catalysts may provide a good reactivity and an onset oxidation potential comparable to pure Pt and serve as promising electrocatalysts for DMFCs with a significant decrease in Pt usage.

  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. Ultimate storage of spent fuel elements of the AVR test power station in the Asse salt mine

    International Nuclear Information System (INIS)

    Wolf, J.

    1975-02-01

    With regard to the ultimate storage of irradiated AVR pebble-bed reactor carbide fuel elements in the saline of Asse, a number of tests and calculations are presented to demonstrate that there is no credible possibility of the MCA (maximum credible accident) defined for the saline. The safety of persons is not threatened during the operation of spent fuel storage nor at any later time (extrapolation up to approx. 1,000 years after storage). 1,000 fuel elements at a time are packed up in gas-tight containers which are stacked in boreholes. The boreholes are then sealed with concrete. Lay-out and functions of the special airlock and transportation systems - from the packing of the containers in a hot cell to the final storage in the borehole - are described with special reference to aspects of the safety of the overall procedure. The possible accidents in the mine are discussed in detail. 85 Kr and T release rates are determined in laboratory tests by heating of the spherical fuel elements. Tests with fuel elements embedded in salt or stagnant brine were carried out at varies temperatures to investigate their behaviour in final storage. Kr and T release, extraction of fission products, mechanical resistance and corrosion were examined in these tests. Finally, the permeability of salt and salt concrete to radioactive gases were investigated in a special experimental arrangement. The diffusion and permeability coefficients obtained for 85 Kr, HT and HTO allow an estimation of the gas discharge of the stored fuel element. (RB/AK) [de

  3. 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.

  4. 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

  5. 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.

  6. 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

  7. 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

  8. 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)

  9. 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)

  10. 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

  11. 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

  12. 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.)

  13. 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.

  14. 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)

  15. 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

  16. 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

  17. 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

  18. 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

  19. 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)

  20. 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

  1. 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)

  2. Fuel fabrication processes, design and experimental conditions for the joint US-Swiss mixed carbide test in FFTF (AC-3 test)

    International Nuclear Information System (INIS)

    Stratton, R.W.; Ledergerber, G.; Ingold, F.; Latimer, T.W.; Chidester, K.M.

    1993-01-01

    The preparation of mixed carbide fuel for a joint (US-Swiss) irradiation test in the US Fast Flux Test Facility (FFTF) is described, together with the experiment design and the irradiation conditions. Two fabrication routes were compared. The US produced 66 fuel pins containing pellet fuel via the powder-pellet (dry) route, and the Swiss group produced 25 sphere pac pins of mixed carbide using the internal gelation (wet) route. Both sets of fuel met all t the requirements of the specifications concerning soichiometry, chemical composition and structure. The pin designs were as similar as possible. The test operated successfully in the FFTF for 620 effective full power days until October 1988 and reached over 8% burn up with peak powers of around 80 kW/m. The conclusions were that the choice of sphere pac or pellet fuel for reactor application is dependent on preferred differences in fabrication (e.g. economics and environmental factors) and not on differences in irradiation behaviour. (orig.)

  3. 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

  4. 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

  5. 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.

  6. 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)

  7. 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)

  8. 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

  9. 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

  10. 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)

  11. 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)

  12. Effect of deposition conditions on the properties of pyrolytic silicon carbide coatings for high-temperature gas-cooled reactor fuel particles

    International Nuclear Information System (INIS)

    Stinton, D.P.; Lackey, W.J.

    1977-10-01

    Silicon carbide coatings on HTGR microsphere fuel act as the barrier to contain metallic fission products. Silicon carbide coatings were applied by the decomposition of CH 3 SiCl 3 in a 13-cm-diam (5-in.) fluidized-bed coating furnace. The effects of temperature, CH 3 SiCl 3 supply rate and the H 2 :CH 3 SiCl 3 ratio on coating properties were studied. Deposition temperature was found to control coating density, whole particle crushing strength, coating efficiency, and microstructure. Coating density and microstructure were also partially determined by the H 2 :CH 3 SiCl 3 ratio. From this work, it appears that the rate at which high quality SiC can be deposited can be increased from 0.2 to 0.5 μm/min

  13. 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)

  14. 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

  15. 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

  16. 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

  17. 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)

  18. 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

  19. 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

  20. 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

  1. 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

  2. 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

  3. 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

  4. 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)

  5. 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

  6. 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

  7. 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)

  8. 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)

  9. 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.

  10. Neutronics performances study of silicon carbide as an inert matrix to achieve very high burn-up for light water reactor fuels

    International Nuclear Information System (INIS)

    Chabert, C.; Coulon-Picard, E.; Pelletier, M.

    2007-01-01

    In order to extend the actual limits of light water reactors, the Cea has put emphasis on the exploration of major fuel innovations that would allow us to increase the competitiveness, the safety and flexibility, while keeping the standard PWR environment. Different fuel concepts have been chosen and are actually studied to evaluate their advantages and drawbacks. The objectives of these new fuels are to increase the safety performances and to achieve a very high burn-up. One concept is a CERCER fuel with silicon carbide (SiC) as an inert matrix devoted to reduce the fuel temperature at nominal conditions. Besides the investigation of the neutronic performance, analyses on the thermomechanical performances, the fuel fabrication, the fuel reprocessing and economic aspects have been performed. This paper presents particularly neutronic results obtained for the CERCER fuel. The results show that a very high burn-up, a high safety performance and a better competitiveness cannot be achieved with this fuel concept. (authors)

  11. Pilot production of 325 kg of uranium carbide

    International Nuclear Information System (INIS)

    Clozet, C.; Dessus, J.; Devillard, J.; Guibert, M.; Morlot, G.

    1969-01-01

    This report describes the pilot fabrication of uranium carbide rods to be mounted in bundles and assayed in two channels of the EL 4 reactor. The fabrication process includes: - elaboration of uranium carbide granules by carbothermic reduction of uranium dioxide; - electron bombardment melting and continuous casting of the granules; - machining of the raw ingots into rods of the required dimensions; finally, the rods will be piled-up to make the fuel elements. Both qualitative and quantitative results of this pilot production chain are presented and discussed. (authors) [fr

  12. 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

  13. Gamma scanning of full scale HTR fuel elements

    International Nuclear Information System (INIS)

    Harrison, T.A.; Simpson, J.A.H.; Nabielek, H.

    1983-04-01

    Gamma scanning for the determination of burn-up and fission product inventory has been developed at the Dragon Project, suitable for measurements on fuel elements and segments from full-sized integral block elements. This involved the design and construction of a new lead flask with sophisticated collimator design. State-of-the art gamma spectrometric equipment was set up to cope with strong variations of count-rate and high data throughput. Software efforts concentrated on the calculation of the self absorption and absorption corrections in the complicated geometry of multi-hole graphite block segments with a corrugated circumference. The techniques described here are applicable to the non-destructive examination of a wide range of fuel element designs. (author)

  14. Fracture and Residual Characterization of Tungsten Carbide Cobalt Coatings on High Strength Steel

    National Research Council Canada - National Science Library

    Parker, Donald S

    2003-01-01

    Tungsten carbide cobalt coatings applied via high velocity oxygen fuel thermal spray deposition are essentially anisotropic composite structures with aggregates of tungsten carbide particles bonded...

  15. Fabrication procedures for manufacturing high uranium concentration dispersion fuel elements

    International Nuclear Information System (INIS)

    Souza, J.A.B.; Durazzo, M.

    2010-01-01

    IPEN developed and made available for routine production the technology for manufacturing dispersion type fuel elements for use in research reactors. However, the fuel produced at IPEN is limited to the uranium concentration of 3.0 gU/cm 3 by using the U 3 Si 2 -Al dispersion. Increasing the uranium concentration of the fuel is interesting by the possibility of increasing the reactor core reactivity and lifetime of the fuel. It is possible to increase the concentration of uranium in the fuel up to the technological limit of 4.8 gU/cm 3 for the U 3 Si 2 -Al dispersion, which is well placed around the world. This new fuel will be applicable in the new Brazilian-Multipurpose Reactor RMB. This study aimed to develop the manufacturing process of high uranium concentration fuel, redefining the procedures currently used in the manufacture of IPEN. This paper describes the main procedures adjustments that will be necessary. (author)

  16. Fabrication procedures for manufacturing high uranium concentration dispersion fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Souza, Jose Antonio Batista de; Durazzo, Michelangelo, E-mail: jasouza@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    IPEN developed and made available for routine production the technology for manufacturing dispersion type fuel elements for use in research reactors. However, the fuel produced at IPEN is limited to the uranium concentration of 3.0 g U/c m3 by using the U{sub 3}Si{sub 2}-Al dispersion. Increasing the uranium concentration of the fuel is interesting by the possibility of increasing the reactor core reactivity and lifetime of the fuel. It is possible to increase the concentration of uranium in the fuel up to the technological limit of 4.8 g U/c m3 for the U{sub 3}Si{sub 2}-Al dispersion, which is well placed around the world. This new fuel will be applicable in the new Brazilian- Multipurpose Reactor RMB. This study aimed to develop the manufacturing process of high uranium concentration fuel, redefining the procedures currently used in the manufacture of IPEN. This paper describes the main procedures adjustments that will be necessary. (author)

  17. Quality control procedures for HTGR fuel element components

    International Nuclear Information System (INIS)

    Delle, W.W.; Koizlik, K.; Luhleich, H.; Nickel, H.

    1976-08-01

    The growing use of nuclear reactors for the production of electric power throughout the world, and the consequent increase in the number of nuclear fuel manufacturers, is giving enhanced importance to the consideration of quality assurance in the production of nuclear fuels. The fuel is the place, where the radioactive fission products are produced in the reactor and, therefore, the integrity of the fuel is of utmost importance. The first and most fundamental means of insuring that integrity is through the exercise of properly designed quality assurance programmes during the manufacture of the fuel and other fuel element components. The International Atomic Energy Agency therefore conducted an International Seminar on Nuclear Fuel Quality Assurance in Oslo, Norway from 24 till 28 May, 1976. This KFA report contains a paper which was distributed preliminary during the seminar and - in the second part - the text of the oral presentation. The paper gives a summary of the procedures available in the present state for the production control of HTGR core materials and of the meaning of the particular properties for reactor operation. (orig./UA) [de

  18. Fabrication procedures for manufacturing high uranium concentration dispersion fuel elements

    International Nuclear Information System (INIS)

    Souza, Jose Antonio Batista de

    2011-01-01

    IPEN-CNEN/SP developed the technology to produce the dispersion type fuel elements for research reactors and made it available for routine production. Today, the fuel produced in IPEN-CNEN/SP is limited to the uranium concentration of 3.0 gU/cm 3 for U 3 Si 2 -Al dispersion-based and 2.3 gU/cm 3 for U 3 O 8 -Al dispersion. The increase of uranium concentration in fuel plates enables the reactivity of the reactor core reactivity to be higher and extends the fuel life. Concerning technology, it is possible to increase the uranium concentration in the fuel meat up to the limit of 4.8 gU/cm 3 in U 3 Si 2 -Al dispersion and 3.2 gU/cm 3 U 3 O 8 -Al dispersion. These dispersions are well qualified worldwide. This work aims to develop the manufacturing process of both fuel meats with high uranium concentrations, by redefining the manufacturing procedures currently adopted in the Nuclear Fuel Center of IPEN-CNEN/SP. Based on the results, it was concluded that to achieve the desired concentration, it is necessary to make some changes in the established procedures, such as in the particle size of the fuel powder and in the feeding process inside the matrix, before briquette pressing. These studies have also shown that the fuel plates, with a high concentration of U 3 Si 2 -Al, met the used specifications. On the other hand, the appearance of the microstructure obtained from U 3 O 8 -Al dispersion fuel plates with 3.2 gU/cm 3 showed to be unsatisfactory, due to the considerably significant porosity observed. The developed fabrication procedure was applied to U 3 Si 2 production at 4.8 gU/cm 3 , with enriched uranium. The produced plates were used to assemble the fuel element IEA-228, which was irradiated in order to check its performance in the IEA-R1 reactor at IPEN-CNEN/SP. These new fuels have potential to be used in the new Brazilian Multipurpose Reactor - RMB. (author)

  19. Equipment for detach the fuel elements of the irradiated candu fuel bundle

    International Nuclear Information System (INIS)

    Cojocaru, V.; Dinuta, G.

    2013-01-01

    Monitoring the behaviour of the fuel bundles during their combustion provides useful information for the operation of the nuclear power plant as well as for the fuel manufacturer. Before placing it inside the reactor, the fuel bundle is inspected visually, dimensionally and, during combustion in the reactor, its radioactive behaviour is monitored. The purpose of the presented equipment is to allow the visual external inspection of the damaged fuel bundle in order to identify visible defects and to detach the fuel element by breaking the welded connection between the cap and grid. These devices are operated using the handler devices already existing in the hot cells Post-Irradiation Examination Laboratory (LEPI). This equipment has been used successfully in the LEPI laboratory at SCN Pitesti to inspect the damaged fuel from Cernavoda NPP, in March 2013. (authors)

  20. The fabrication of nuclear fuel elements in Mexico

    International Nuclear Information System (INIS)

    Guerrero Morillo, H.L.

    1977-01-01

    The situation of nuclear electricity generation in Mexico in 1976 is described: two nuclear reactors were under construction but no definite programme on the type and start-up dates for the next power plants existed. However, the existence of a general plan on nuclear power plants is mentioned, which, according to the latest estimates, will provide 10,000MW installed by 1990. The national intention, as laid down in an appropriate Law, is to supply domestic nuclear fuel to the power reactors operating in the country, starting with the first reloading of the two BWRs at the first national station in Laguna Verde, required at the end of 1981 and 1982, respectively. Before this can be achieved and to provide the relatively small amounts of fuel elements for the two reactors, Mexico must adopt a strategy of fuel elements fabrication. The two main options are analysed: (1) to delay local fabrication until a national nuclear programme has been defined, meanwhile purchasing abroad the necessary initial cores and refuelling; (2) to start local fabrication of fuel elements as soon as possible in order to provide the first refuelling of the first unit of Laguna Verde, confronting the economic risks of such a decision with the advantages of immediate action. Both options are analysed in detail, comparing them especially from the economic point of view. Current information from potential licensors for design and manufacture are used in the analysis. (author)

  1. Fuel Element Mechanical Design for CAREM-25 Reactor

    International Nuclear Information System (INIS)

    Estevez, Esteban; Markiewicz, Mario; Gerding, Jose

    2000-01-01

    The Fuel Element mechanical design and spider-control reactivity and security rods assembly for the CAREM-25 reactor is introduced. The CAREM-25 Fuel Element has a hexagonal cross section with 127 positions, in a triangular arrangement.There are 108 positions for the fuel rods while the guide tubes and instrumentation tube occupy the 19 remaining positions.From the structural point of view, the fuel element is being composed by a framework formed by the guides and instrumentation tubes, 4 spacer grids and the upper and lower coupling pieces.The spider is a plane piece, with a central body and six radial branches in T form, which has holes where the absorber rods are fitted.The central body ends in a joint in the upper side, which allows connect the assembly whit the reactor control mechanisms.The absorber rods are made of a neutron absorber material (Ag-In-Cd) hermetically closed in a stainless steel cladding. In this work are determined, in addition to the basic design, the operational conditions, the functional requirements to be satisfied and in agreement with those, the adopted criteria and limits to avoid systematics failure during normal operation conditions. The proposed program for the verification and evaluation of design is detailed.To consolidate the design, a prototype was manufactures, based on drawings and specifications needed for its construction

  2. The behaviour of spherical HTR fuel elements under accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Schenk, W; Naoumidis, A [Institute for Reactor Material, KFA Juelich (Germany)

    1985-07-01

    Hypothetical accidents may lead to significantly higher temperatures in HTR fuel than during normal operation. In order to obtain meaningful statements on fission product behaviour and release, irradiated spherical fuel elements containing a large number of coated particles (20,000-40,000) with burnups between 6 and 16% FIMA were heated at temperatures between 1400 and 2500 deg. C. HTI-pyrocarbon coating retains the gaseous fission products (e.g. Kr) very well up to about 2400 deg. C if the burnup does not exceed the specified value for THTR (11.5%). Cs diffuses through the pyrocarbon significantly faster than Kr and the diffusion is enhanced at higher fuel burnups because of irradiation induced kernel microstructure changes. Below about 1800 deg. C the Cs release rate is controlled by diffusion in the fuel kernel; above this temperature the diffusion in the pyrocarbon coating is the controlling parameter. An additional SiC coating interlayer (TRISO) ensures Cs retention up to 1600 deg. C. However, the release obtained in the examined fuel elements was only by a factor of three lower than through the HTI pyrocarbon. Solid fission products added to UO{sub 2}-TRISO particles to simulate high burnup behave in various ways and migrate to attack the SiC coating. Pd migrates fastest and changes the SiC microstructure making it permeable.

  3. Use of plate fuel elements for the RA3 reactor

    International Nuclear Information System (INIS)

    Parodi, C.; Parkanski, D.; Higa, M.; Marajofsky, A.

    1992-01-01

    The RA3 reactor is a pool reactor, redesigned for 5 MW dissipation. Nineteen plates are used in each fuel element. The utilization of 20% enriched U, gives the possibility of the development of rod type fuel with Al/U 3 O 8 cermets. The thermohydraulic and neutronic conditions are studied in this work in order to satisfy the stipulated power. In addition, the fabrication conditions of Al/U 3 O 8 and Al/U 3 O 8 /Zr H 2 cermets with densities within the limits imposed by the thermohydraulics and neutronics conditions are studied. (author)

  4. Radial heat conduction in a power reactor fuel element

    International Nuclear Information System (INIS)

    Ventura, M.A.

    1998-01-01

    Two radial conduction models, one for steady state and another for unsteady state, in a nuclear power reactor fuel element are developed. The objective is to obtain the temperatures in the fuel pellet and the cladding. The lumped-parameter hypothesis are adopted to represent the system. Both models are verified and their results are compared with similar ones. A method to calculate the conductance in the gap between the UO 2 pellet and the clad and its associated uncertainty is included in the steady state model. (author) [es

  5. Characteristics and behaviour of the PHENIX fuel element

    International Nuclear Information System (INIS)

    Delpeyroux, P.; Balloffet, Y.; Blanchard, P.; Courcon, P.; Jallade, M.; Millet, P.; Rousseau, J.; Carteret, Y.; Coulon, P.

    1977-01-01

    The Phenix reactor has been in regular industrial operation for two years and has functioned very satisfactorily thanks in particular to the very good behaviour of the fuel element. A brief description is given of the fuel element and the operating conditions which were set for the fuel at the time of start-up (50000 MWd/t). The surveillance scheme is then described with the examinations in the hot laboratory on the basis of which it was possible to achieve the nominal specific burn-up and then to clear the Phenix fuel for a specific burn-up of 60000 MWd/t or 7 at.%. The behaviour of the mixed oxide (U, Pu)O 2 is quite normal and conforms to predictions as regards the heat conditions, swelling and fission gas release. The corrosion reaction between the oxide and the clad is progressing slowly and affects only small thicknesses of cladding. The mechanical integrity of the clad under thermal stresses and the stresses produced by swelling and fission gas pressure do not pose any special problem. The present limitation of the irradiation level is essentially based on the permissible deformations due to swelling and irradiation creep in the fuel pin cladding and in the hexagonal tube. This corresponds to damage to the steel of the order of 80 dpa. The mechanical behaviour of the bundle of pins, its interaction with the hexagonal tube and the thermohydraulic consequences of the deformations are all satisfactory to date. The absence of fuel failures is also worth noting; the only burst can detected to date did not affect either the operation of the fuel assembly or the performance of the reactor [fr

  6. Process and equipment for locating defective fuel rods of a reactor fuel element

    International Nuclear Information System (INIS)

    Jester, A.; Honig, H.

    1977-01-01

    By this equipment, well-known processes for determining defective fuel rods of a reactor fuel element are improved in such a fashion that defective fuel rods can be located individually, so that it is possible to replace them. The equipment consists of a cylindrical test vessel open above, which accommodates the element to be tested, so that an annular space is left between the latter's external circumference and the wall of the vessel, and so that the fuel rods project above the vessel. A bell in the shape of a frustrum of a cone is inverted over the test vessel, which has an infra-red measuring equipment at a certain distance above the tops of the fuel rods. The fuel element to be tested together with the test vessel and hood are immersed in a basin full of water, which displaces water by means of gas from the hood. The post-shutdown heat increases the temperature in the water space of the test vessel, which is stabilised at 100 0 C. In each defective fuel rod the water which has penetrated the defective fuel rod previously, or does so now, starts to boil. The steam rising in the fuel rod raises the temperature of the defective fuel rod compared to all the sound ones. The subsequent measurement easily determines this. Where one can expect interference with the measurement by appreciable amounts of gamma rays, the measuring equipment is removed from the path of radiation by mirror deflection in a suitably shaped measuring hood. (FW) [de

  7. High surface area synthesis, electrochemical activity, and stability of tungsten carbide supported Pt during oxygen reduction in proton exchange membrane fuel cells

    Science.gov (United States)

    Chhina, H.; Campbell, S.; Kesler, O.

    The oxidation of carbon catalyst supports to carbon dioxide gas leads to degradation in catalyst performance over time in proton exchange membrane fuel cells (PEMFCs). The electrochemical stability of Pt supported on tungsten carbide has been evaluated on a carbon-based gas diffusion layer (GDL) at 80 °C and compared to that of HiSpec 4000™ Pt/Vulcan XC-72R in 0.5 M H 2SO 4. Due to other electrochemical processes occurring on the GDL, detailed studies were also performed on a gold mesh substrate. The oxygen reduction reaction (ORR) activity was measured both before and after accelerated oxidation cycles between +0.6 V and +1.8 V vs. RHE. Tafel plots show that the ORR activity remained high even after accelerated oxidation tests for Pt/tungsten carbide, while the ORR activity was extremely poor after accelerated oxidation tests for HiSpec 4000™. In order to make high surface area tungsten carbide, three synthesis routes were investigated. Magnetron sputtering of tungsten on carbon was found to be the most promising route, but needs further optimization.

  8. High surface area synthesis, electrochemical activity, and stability of tungsten carbide supported Pt during oxygen reduction in proton exchange membrane fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Chhina, H. [Automotive fuel cell corporation, 9000 Glenlyon Parkway, Burnaby, BC (Canada); Department of Mechanical and Industrial Engineering, 5 King' s College Road, University of Toronto, Toronto, Ontario (Canada); Campbell, S. [Automotive fuel cell corporation, 9000 Glenlyon Parkway, Burnaby, BC (Canada); Kesler, O. [Department of Mechanical and Industrial Engineering, 5 King' s College Road, University of Toronto, Toronto, Ontario (Canada)

    2008-04-15

    The oxidation of carbon catalyst supports to carbon dioxide gas leads to degradation in catalyst performance over time in proton exchange membrane fuel cells (PEMFCs). The electrochemical stability of Pt supported on tungsten carbide has been evaluated on a carbon-based gas diffusion layer (GDL) at 80 C and compared to that of HiSpec 4000 trademark Pt/Vulcan XC-72R in 0.5 M H{sub 2}SO{sub 4}. Due to other electrochemical processes occurring on the GDL, detailed studies were also performed on a gold mesh substrate. The oxygen reduction reaction (ORR) activity was measured both before and after accelerated oxidation cycles between +0.6 V and +1.8 V vs. RHE. Tafel plots show that the ORR activity remained high even after accelerated oxidation tests for Pt/tungsten carbide, while the ORR activity was extremely poor after accelerated oxidation tests for HiSpec 4000 trademark. In order to make high surface area tungsten carbide, three synthesis routes were investigated. Magnetron sputtering of tungsten on carbon was found to be the most promising route, but needs further optimization. (author)

  9. Elastic plastic analysis of fuel element assemblies - hexagonal claddings and fuel rods

    International Nuclear Information System (INIS)

    Mamoun, M.M.; Wu, T.S.; Chopra, P.S.; Rardin, D.C.

    1979-01-01

    Analytical studies have been conducted to investigate the structural, thermal, and mechanical behavior of fuel rods, claddings and fuel element assemblies of several designs for a conceptual Safety Test Facility (STF). One of the design objectives was to seek a geometrical configuration for a clad by maximizing the volume fraction of fuel and minimizing the resultant stresses set-up in the clad. The results of studies conducted on various geometrical configurations showed that the latter design objective can be achieved by selecting a clad of an hexagonal geometry. The analytical studies necessitated developing solutions for determining the stresses, strains, and displacements experienced by fuel rods and an hexagonal cladding subjected to thermal fuel-bowing loads acting on its internal surface, the external pressure of the coolant, and elevated temperatures. This paper presents some of the initially formulated analytical methods and results. It should be emphasized that the geometrical configuration considered in this paper may not necessarily be similar to that of the final design. Several variables have been taken into consideration including cladding thickness, the dimensions of the fuel rod, the temperature of the fuel and cladding, the external pressure of the cooling fluid, and the mechanical strength properties of fuel and cladding. A finite-element computer program, STRAW Code, has also been employed to generate several numerical results which have been compared with those predicted by employing the initially formulated solutions. The theoretically predicted results are in good agreement with those of the STRAW Code. (orig.)

  10. Optimizing the dual elemental thermal reactive deposition time in carbide layer formation on SUJ2 tool steel

    Science.gov (United States)

    Mochtar, Myrna Ariati; Putra, Wahyuaji Narottama; Mahardika, Bayu

    2018-05-01

    This paper presents developments contributing to the improvement of thermo-reactive deposition (TRD) process in producing hard carbide layers, on automotive components application. The problem in using FeV powder as a coating material that has been applied in the industries is it is high cost. In this study, FeCr powder coating material was mixed into FeV powder with a ratio of 35:65 weight percent. The SUJ2 steel pins components are processed at 980° C, with varying TRD time was 4,6,8 and 10 hours. Scanning Electron microscope (SEM), Electron Probe Micro Analyzer (EPMA) and X-ray diffraction (XRD) were applied to analyze the coating layers. The thickness of the carbide layer formed will increase with the longer processing time, which thickness at 4-10 hours is increase from 22.7 to 29.7 micron. The gained thickness tends to be homogeneous. Increasing the TRD process holding time results in a higher hardness of the carbide layerwith hardness at 4, 6, 8 and 10 hours is 2049, 2184, 2175 and 2343 HV. The wear rate at TRD holding time of 4-10 hours with the Ogoshi method was reduced from 5.1 × 10-4 mm3/m to 2.5 × 10-4 mm3/m. Optical microscope observations shows that substrate phases consisting of pearlite and cementite and grains that tend to enlarge with the addition of time. Carbide compounds that are formed are vanadium carbide (V8C7, V6C5, V2C) and chromium carbide (Cr3C2, Cr23C7, Cr3C7). While EDS-Linescan results show complex phase (Fe, V, Cr) xC formed. The research shows that addition of FeCr into FeV powder in TRD process in 980°C with optimum time of 10 hours processing meet the mechanical properties requirement of automotive components.

  11. Cumulative damage fraction design approach for LMFBR metallic fuel elements

    International Nuclear Information System (INIS)

    Johnson, D.L.; Einziger, R.E.; Huchman, G.D.

    1979-01-01

    The cumulative damage fraction (CDF) analytical technique is currently being used to analyze the performance of metallic fuel elements for proliferation-resistant LMFBRs. In this technique, the fraction of the total time to rupture of the cladding is calculated as a function of the thermal, stress, and neutronic history. Cladding breach or rupture is implied by CDF = 1. Cladding wastage, caused by interactions with both the fuel and sodium coolant, is assumed to uniformly thin the cladding wall. The irradiation experience of the EBR-II Mark-II driver fuel with solution-annealed Type 316 stainless steel cladding provides an excellent data base for testing the applicability of the CDF technique to metallic fuel. The advanced metal fuels being considered for use in LMFBRs are U-15-Pu-10Zr, Th-20Pu and Th-2OU (compositions are given in weight percent). The two cladding alloys being considered are Type 316 stainless steel and a titanium-stabilized Type 316 stainless steel. Both are in the cold-worked condition. The CDF technique was applied to these fuels and claddings under the assumed steady-state operating conditions

  12. Design and operational behaviour of the SNR-reactor fuel element structure

    International Nuclear Information System (INIS)

    Dietz, W.; Toebbe, H.

    1985-01-01

    The fuel element and core concept of a fast breeder reactor is described by the example of the SNR 300 (1st core), and the requirements made on the fuel elements with respect to burnup and neutron dose are listed for existing and projected plants. Irradiation experiments carried out and operational experience gained with fuel elements show that the residence time of the fuel elements is influenced mainly by the stability of shape of the fuel element components. The requirements made with reference to neutron loading for future advanced high-performance fuel elements can not be anticipated from the present state of experience. Besides optimization of fuel element design and checking-out of the limits of operation by PFADFINDERELEMENTE elements, R and D work for the improvement of fuel element materials is also necessary. (orig.) [de

  13. Method for increasing the activity of fuel cell electrodes containing tungsten carbide. Verfahren zur Steigerung der Aktivitaet von Brennstoffelektroden, die Wolframcarbid enthalten

    Energy Technology Data Exchange (ETDEWEB)

    Binder, H.; Koehling, A.; Kuhn, W.; Lindner, W.; Sandstede, G.

    1977-10-13

    An increase in the activity of electrodes containing tungsten carbide for a low-temperature fuel cell with sulfuric acid as electrolyte can be achieved, if one operates the electrodes for a few hours (5-20 h) in the presence of hydrogen and a means of reduction (formaldehyde, hydrazene) in a voltage range of between +500 and +800 mV (relative to the H/sub 2/ electrode). A corrosion resistant layer is formed, which is assumed to have the composition WC/sub X/O/sub y/H/sub z/.

  14. Improved CVD Coatings for Carbide Based Nuclear Thermal Propulsion Fuel Elements, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — One of the great hurdles to further development and evaluation of nuclear thermal propulsion systems is the issue surrounding the release of radioactive material...

  15. The industrial production of fuel elements; La fabrication en france des elements combustibles

    Energy Technology Data Exchange (ETDEWEB)

    Boussard, R [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires; Nadal, J [Societe Industrielle de Combustible Nucleaire (SICN), 75 - Paris (France); Pellen, A [Compagnie pour l' Etude et la Realisation de Combustibles Atomiques (CERCA), 75 - Paris (France)

    1964-07-01

    The authors deal successively with the industrial production of fuel elements for power reactors of the natural uranium-graphite-gas type, and more particularly for the EDF power stations, and with the industrial production of fuel elements containing enriched uranium designed for swimming-pool type reactors. 1. part: advanced fuel elements for the EDF reactors. After recalling the characteristics of the fuel elements now being produced industrially for the Marcoule and Chinon reactors, the authors give the various steps leading to the industrial production of a new type of fuel element both as concerns the can, and in certain cases the graphite sleeve, and the fuel itself. As for as the production of the fuel is concerned, they describe the various operations, stressing the original aspects of the production and of the equipment such as: - casting in hot moulds, - thermal treatments, of Uranium containing 1% in weight molybdenum, - welding of the pellets for closing the tubes of uranium, - canning, - controls in the various steps. As far as can production is concerned they show why the extruded can was replaced by a machined can and give a few characteristics of the equipment used as well as the controls effected. They give also some details concerning the production and machining of the sleeves. After recalling the state of the nuclear fuel industry in France in mid-1964 the authors stress the economic aspects of the production of fuel elements. They show the relative importance of capital costs on the cost price of the fuel itself and examine the various items involved. They analyse the cost price of a completed fuel element using present date knowledge. In conclusion the authors show the particular points which should be the subject of future efforts in order to decrease the cost of a production which is perhaps delicate but now will define, and review the development of this new industrial branch. 2. part: industrial production of fuel elements for swimming

  16. Block fuel element for gas-cooled high temperature reactors

    International Nuclear Information System (INIS)

    Hrovat, M.F.

    1978-01-01

    The invention concerns a block fuel element consisting of only one carbon matrix which is almost isotropic of high crystallinity into which the coated particles are incorporated by a pressing process. This block element is produced under isostatic pressure from graphite matrix powder and coated particles in a rubber die and is subsequently subjected to heat treatment. The main component of the graphite matrix powder consists of natural graphite powder to which artificial graphite powder and a small amount of a phenol resin binding agent are added

  17. 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

  18. Fuel element gamma scanning at the Oak Ridge Research Reactor

    International Nuclear Information System (INIS)

    Hobbs, R.W.

    1987-01-01

    In January 1986, a demonstration program was begun at the Oak Ridge Research Reactor (ORR) to convert operations from high-enrichment uranium fuel to the newly developed U 3 Si 2 low-enrichment fuel. A primary program objective is to validate neutronics calculations conducted by the Reduced Enrichment in Research and Test Reactors Program at Argonne National Laboratory. Accordingly, a new method for determining core-power distribution has been developed. The method is based on gamma-ray spectroscopy measurements to determine the relative levels of 140 La in the fuel elements after each operating cycle. The measurement and data analyses are described and a comparison of measured and diffusion theory calculated values of the core-power distribution is presented in this paper

  19. Fuel elements for pressurised-gas reactors; Elements combustibles des piles a gaz sous pression

    Energy Technology Data Exchange (ETDEWEB)

    Stohr, J A; Englander, M; Gauthron, M [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1958-07-01

    The design and fabrication of fuel elements for the first CO{sub 2} pressurized reactors have induced to investigate: various cladding materials, natural uranium base fuels, canning processes. The main analogical tests used in connection with the fuel element study are described. These various tests have enabled, among others, the fabrication of the fuel element for the EL2 reactor. Lastly, future solutions for electrical power producing reactors are foreseen. (author)Fren. [French] L'etude et la realisation d'elements combustibles pour les premieres piles a CO{sub 2} sous pression ont conduit a examiner: les divers materiaux de gaine, les combustibles a base d'uranium naturel, les modes de gainage. Les principaux essais analogiques ayant servi au cours de l'etude de la cartouche sont decrits. Ces divers essais ont notamment permis la realisation de la cartouche de la pile EL2. Enfin sont envisagees les solutions futures pour les piles productrices d'energie electrique. (auteur)

  20. CARA CVN: inherently safe fuel element for PHWR power plants

    International Nuclear Information System (INIS)

    Brasnarof, Daniel O.; Lestani, Hector A.; Agueda, Horacio C.; Marino, Armando C.; Florido, Pablo C.; Daverio, Hernando

    2007-01-01

    This paper presents design alternatives of the CARA fuel element with negative void reactivity coefficient (CVN) enhancing the PHWR safety for L-LOCA sequences. This design enhances the safety and the operation performance in Atucha and Embalse without changes in the operation conditions. This new design balances wide performance margins of CARA SEU 0.9% previous design, with new intrinsic safety requirements without economic penalties. (author) [es

  1. Hydraulic Design Criteria for Spacer Grids of Nuclear Fuel Element

    International Nuclear Information System (INIS)

    Juanico, Luis; Brasnarof, Daniel

    2000-01-01

    In this paper a hydraulic model for calculating the pressure drop on the CARA spacer grids is extended.This model is validated and feedback from experimental hydraulic test performed in a low pressure loop.The importance of the spacer grid geometric parameter (that is, its thickness and length, the number and kind of their fix spacer), developing hydraulic design criteria for spacer grid on fuel element

  2. Quality assurance in nuclear fuel element component supply

    International Nuclear Information System (INIS)

    Jenkins, B.P.

    1987-01-01

    The paper describes the application of Quality Assurance to nuclear fuel element component supply. The Quality Assurance programme includes integrated procurement, purchasing, surveillance and receipt inspection functions. Purchasing policy is based on a consistent preference for competitive tendering. Multiple sourcing is used to encourage competitive pricing and increase security of supply. A receipt inspection facility is maintained to ensure the high product quality levels demanded by the nuclear industry. (U.K.)

  3. Dry storage of spent fuel elements: interim facility

    International Nuclear Information System (INIS)

    Quihillalt, O.J.

    1993-01-01

    Apart from the existing facilities to storage nuclear fuel elements at Argentina's nuclear power stations, a new interim storage facility has been planned and projected by the Argentinean Atomic Energy Commission (CNEA) that will be constructed by private group. This article presents the developments and describes the activities undertaken until the national policy approach to the final decision for the most suitable alternative to be adopted. (B.C.A.). 09 refs, 01 fig, 09 tabs

  4. Design verification testing for fuel element type CAREM

    International Nuclear Information System (INIS)

    Martin Ghiselli, A.; Bonifacio Pulido, K.; Villabrille, G.; Rozembaum, I.

    2013-01-01

    The hydraulic and hydrodynamic characterization tests are part of the design verification process of a nuclear fuel element prototype and its components. These tests are performed in a low pressure and temperature facility. The tests requires the definition of the simulation parameters for setting the test conditions, the results evaluation to feedback mathematical models, extrapolated the results to reactor conditions and finally to decide the acceptability of the tested prototype. (author)

  5. Convective parameters in fuel elements for research nuclear reactors

    International Nuclear Information System (INIS)

    Lopez Martinez, C.D.

    1992-01-01

    The study of a prototype for the simulation of fuel elements for research nuclear reactors by natural convection in water is presented in this paper. This project is carry out in the thermofluids laboratory of National Institute of Nuclear Research. The fuel prototype has already been test for natural convection in air, and the first results in water are presented in this work. In chapter I, a general description of Triga Mark III is made, paying special atention to fuel-moderator components. In chapter II and III an approach to convection subject in its global aspects is made, since the intention is to give a general idea of the events occuring around fuel elements in a nuclear reactor. In chapter II, where an emphasis on forced convection is made, some basic concepts for forced convection as well as for natural convection are included. The subject of flow through cylinders is annotated only as a comparative reference with natural convection in vertical cylinders, noting the difference between used correlations and the involved variables. In chapter III a compilation of correlation found in the bibliography about natural convection in vertical cylinders is presented, since its geometry is the more suitable in the analysis of a fuel rod. Finally, in chapter IV performed experiments in the test bench are detailed, and the results are presented in form of tables and graphs, showing the used equations for the calculations and the restrictions used in each case. For the analysis of the prototypes used in the test bench, a constant and uniform flow of heat in the whole length of the fuel rod is considered. At the end of this chapter, the work conclusions and a brief explanation of the results are presented (Author)

  6. Cooling flow measurement in fuel elements of the RA-6

    International Nuclear Information System (INIS)

    Brollo, F; Silin, N

    2009-01-01

    Under the UBERA6 project for the core change and power increase of the RA-6 reactor, the total coolant flow was increased to meet the requirements imposed by the new operating conditions. The flow through the fuel elements is an important parameter and is difficult to determine due to the geometric complexity of the core. To ensure safe operation of the reactor, adequate safety margins must be kept for all operating conditions. In the present work we performed the direct measurement of the cooling flow rate of a fuel in the reactor core, for which we used a turbine flowmeter built specifically for this use. This helped to confirm previous results obtained during the launch, made by an indirect method based on measuring the pressure difference of the core. The turbine flowmeter was chosen due to its robustness, ease of operation and low disturbance of the input stream to the fuel. We describe the calibration of this instrument and the results of flow measurements made on some of the RA6 reactor fuel elements under conditions of zero power. [es

  7. Fabrication procedures for manufacturing High Flux Isotope Reactor fuel elements - 2

    International Nuclear Information System (INIS)

    Knight, R.W.; Morin, R.A.

    1999-01-01

    The original fabrication procedures written in 1968 delineated the manufacturing procedures at that time. Since 1968, there have been a number of procedural changes. This rewrite of the fabrication procedures incorporates these changes. The entire fuel core of this reactor is made up of two fuel elements. Each element consists of one annular array of fuel plates. These annuli are identified as the inner and outer fuel elements, since one fits inside the other. The inner element consists of 171 identical fuel plates, and the outer element contains 369 identical fuel plates differing slightly from those in the inner element. Both sets of fuel plates contain U 3 O 8 powder as the fuel, dispersed in an aluminum powder matrix and clad with aluminum. Procedures for manufacturing and inspection of the fuel elements are described and illustrated

  8. Fabrication procedures for manufacturing High Flux Isotope Reactor fuel elements - 2

    Energy Technology Data Exchange (ETDEWEB)

    Knight, R.W.; Morin, R.A.

    1999-12-01

    The original fabrication procedures written in 1968 delineated the manufacturing procedures at that time. Since 1968, there have been a number of procedural changes. This rewrite of the fabrication procedures incorporates these changes. The entire fuel core of this reactor is made up of two fuel elements. Each element consists of one annular array of fuel plates. These annuli are identified as the inner and outer fuel elements, since one fits inside the other. The inner element consists of 171 identical fuel plates, and the outer element contains 369 identical fuel plates differing slightly from those in the inner element. Both sets of fuel plates contain U{sub 3}O{sub 8} powder as the fuel, dispersed in an aluminum powder matrix and clad with aluminum. Procedures for manufacturing and inspection of the fuel elements are described and illustrated.

  9. Inspection of fuel elements in the cooling pond of a research reactor

    International Nuclear Information System (INIS)

    Pavlov, S.V.; Mestnikov, A.V.

    1992-01-01

    Nondestructive testing methods for fuel bundles and fuel elements in the cooling ponds of atomic power plants, using special inspection stands, have come into widespread use during the past decade. This paper describes a methodological stand that was built for the laboratory development of methods and individual units of inspection stands for fuel bundles of RBMK and VVER-1000 reactors. A complex of equipment was developed for the study of irradiated fuel elements, thus creating a methodological base for developing techniques for nondestructive testing of irradiated fuel elements and equipment to obtain information about the state of the fuel elements in a reactor expeditiously. The time required to inspect a fuel element can be shortened using some techniques simultaneously. The length of a fuel element can be measured simultaneously with visual inspection, eddy-current flaw detection can be preformed at the same time as the tranverse size of the fuel element is being determined. 6 refs., 5 figs

  10. 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)

  11. Behavior of LWR fuel elements under accident conditions

    International Nuclear Information System (INIS)

    Albrecht, H.; Bocek, M.; Erbacher, F.; Fiege, A.; Fischer, M.; Hagen, S.; Hofmann, P.; Holleck, H.; Karb, E.; Leistikow, S.; Melang, S.; Ondracek, G.; Thuemmler, F.; Wiehr, K.

    1977-01-01

    In the frame of the German reactor safety research program, the Kernforschungszentrum Karlsruhe is carrying out a comprehensive program on the behavior of LWR fuel elements under a variety of power cooling mismatch conditions in particular during loss-of-coolant accidents. The major objectives are to establish a detailed quantitative understanding of fuel rod failures mechanisms and their thresholds, to evaluate the safety margins of power reactor cores under accident conditions and to investigate the feedback of fuel rod failures on the efficiency of emergency core cooling systems. This detailed quantitative understanding is achieved through extensive basic and integral experiments and is incorporated in a fuel behavior code. On the basis of these results the design of power reactor fuel elements and of safety devices can be further improved. The results of investigations on the inelastic deformation (ballooning) behavior of Zircaloy 4 cladding at LOCA temperatures in oxidizing atmosphere are presented. Depending upon strain rate and temperature superplastic deformation behavior was observed. In the equation of state of Zry 4 the strain rate sensitivity index depends strongly upon strain and in the superplastic region upon sample anisotropy. Oxidation kinetics experiments with Zry-tubes at 900-1300 0 C showed that the Baker-Just correlation describes the reality quite conservative. Therefore a reduction of the amount of Zry oxidation can be assumed in the course of a LOCA. The external oxidation of Zry-cladding by steam as well as internal oxidation by the oxygen in oxide fuel and fission products (Cs, I, Te) have an influence on the strain and rupture behavior of Zry-cladding at LOCA temperatures. In out-of-pile and inpile experiments the mechanical and thermal behavior of fuel rods during the blowdown, the heatup and the reflood phases of a LOCA are investigated under representative and controlled thermohydraulic conditions. The task of the inpile experiments is

  12. Water reactor fuel element fabrication, with special emphasis on its effects on fuel performance

    International Nuclear Information System (INIS)

    1979-01-01

    Full text: The performance of nuclear fuel has improved over the years and is now a minor cause of outages and of power limitations in nuclear power plants. On the other hand, an increasing number of countries are in the process of developing or implementing their own capability for manufacturing fuel elements. In this context, the IAEA International Working Group on Water Reactor Fuel Performance and Technology (IWGFPT) advised that a symposium be organized devoted to the relationship between fuel fabrication and performance The Czechoslovak Atomic Energy Commission agreed to co-operate in the organization of this symposium and to host it in Prague. Those factors which influence fuel fabrication requirements are now well ascertained: as little reactor primary circuit contamination as possible, the tendency to increased burnups, reactor manoeuverability to match power grid demands, the desirability of an autonomous fabrication capability. It is the general experience of fuel element suppliers that fuel quality and performance has increased over the years, the importance of quality assurance and process monitoring has been decisive in this respect The ever increasing mass-production aspect of nuclear fuel leads to some processing steps being revised and alternatives being developed. The relation between fabrication processes and fuel performance characteristics, although generally well perceived, are still the subject of a large amount of experiment and assessment in most countries, both industrial and developing This evidence is most encouraging; it means indeed that nuclear power, which is already amongst the cheapest and safest sources of energy, will continue to be improved. The performance of Zircaloy fuel cladding - presently the material used in most water reactors - is under particular consideration. Better understanding of this quite recent alloy will pave the way for broader fuel utilization limits in the future. The panel discussion, which noted some

  13. A novel microbial fuel cell sensor with biocathode sensing element.

    Science.gov (United States)

    Jiang, Yong; Liang, Peng; Liu, Panpan; Wang, Donglin; Miao, Bo; Huang, Xia

    2017-08-15

    The traditional microbial fuel cell (MFC) sensor with bioanode as sensing element delivers limited sensitivity to toxicity monitoring, restricted application to only anaerobic and organic rich water body, and increased potential fault warning to the combined shock of organic matter/toxicity. In this study, the biocathode for oxygen reduction reaction was employed for the first time as the sensing element in MFC sensor for toxicity monitoring. The results shown that the sensitivity of MFC sensor with biocathode sensing element (7.4±2.0 to 67.5±4.0mA% -1 cm -2 ) was much greater than that showed by bioanode sensing element (3.4±1.5 to 5.5±0.7mA% -1 cm -2 ). The biocathode sensing element achieved the lowest detection limit reported to date using MFC sensor for formaldehyde detection (0.0005%), while the bioanode was more applicable for higher concentration (>0.0025%). There was a quicker response of biocathode sensing element with the increase of conductivity and dissolved oxygen (DO). The biocathode sensing element made the MFC sensor directly applied to clean water body monitoring, e.g., drinking water and reclaimed water, without the amending of background organic matter, and it also decreased the warning failure when challenged by a combined shock of organic matter/toxicity. Copyright © 2017 Elsevier B.V. All rights reserved.

  14. Nuclear reactor fuel element containing an end piece for maintaining the column of fuel pellets

    International Nuclear Information System (INIS)

    Pajot, Jacques; Rabellino, Jacques.

    1974-01-01

    The nuclear reactor fuel element described has an end piece for maintaining the column of fuel pellets in position inside the element cladding. This end piece has a central compression spring one end of which presses against the pellets and the other against a plug shaped piece fitted with a seat for the spring, a conical piece with an elastic ring around it diverging towards the end in contact with the spring and a head at the opposite end. The connection between the compression spring and the pellets is through an application piece. A central bore provided in the end piece helps balance the pressure inside the element. This element is particularly intended for liquid metal cooled fast neutron reactors [fr

  15. Fission product release profiles from spherical HTR fuel elements at accident temperatures

    International Nuclear Information System (INIS)

    Schenk, W.; Pitzer, D.; Nabielek, H.

    1986-10-01

    A total of 22 fuel elements with modern TRISO particles has been tested in the temperature range 1500-2500 0 C. Additionally, release profiles of iodine and other isotopes have been obtained with seven UO 2 samples at 1400-1800 0 C. For heating times up to 100 hours at the maximum temperature, the following results are pertinent to HTR accident conditions: Ag 110 m is the only fission products to be released at 1200-1600 0 C by diffusion through intact SiC, but it is of low significance in accident assessments; cesium, iodine, strontium, and noble gas releases up to 1600 0 C are solely due to various forms of contamination; at 1700-1800 0 C, corrosion induced SiC defects cause the release of Cs, Sr, I/Xe/Kr; above 2000 0 C, thermal decomposition of the silicon carbide layer sets in while pyrocarbons still remain intact. Around 1600 0 C, the accident specific contribution of cesium, strontium, iodine, and noble gases is negligible. (orig./HP) [de

  16. The element technology of clean fuel alcohol plant construction

    Energy Technology Data Exchange (ETDEWEB)

    Lee, D S; Lee, D S [Sam-Sung Engineering Technical Institute (Korea, Republic of); Choi, C Y [Seoul National University, Seoul (Korea, Republic of); and others

    1996-02-01

    The fuel alcohol has been highlighted as a clean energy among new renewable energy sources. However, the production of the fuel alcohol has following problems; (i)bulk distillate remains is generated and (ii) benzene to be used as a entertainer in the azeotropic distillation causes the environmental problem. Thus, we started this research on the ground of preserving the cleanness in the production of fuel alcohol, a clean energy. We examined the schemes of replacing the azotropic distillation column which causes the problems with MSDP(Molecular Sieve Dehydration Process) system using adsorption technology and of treating the bulk distillate remains to be generated as by-products. In addition, we need to develop the continuous yea station technology for the continuous operation of fuel alcohol plant as a side goal. Thus, we try to develop a continuous ethanol fermentation process by high-density cell culture from tapioca, a industrial substrate, using cohesive yeast. For this purpose, we intend to examine the problem of tapioca, a industrial substrate, where a solid is existed and develop a new process which can solve the problem. Ultimately, the object of this project is to develop each element technology for the construction of fuel alcohol plant and obtain the ability to design the whole plant. (author) 54 refs., 143 figs., 34 tabs.

  17. The fabrication of nuclear fuel elements in Mexico

    International Nuclear Information System (INIS)

    Guerrero Morillo, H.L.

    1977-01-01

    The situation of the nucleoelectrical generation in Mexico by 1976 is described: two nuclear reactors under construction but no defined program on the type and start-up dates for the next power plants. However the existence of a general plan on nuclear power plants is mentioned, which, according to the last estimates reaches to 10,000 MW installed by 1990. The national intension, definitely expressed in the Law, is to supply domestic nuclear fuel to the power reactors operating in the country, starting with the first reload for the two BWR's at the first national station in Laguna Verde, which will be required at the end of 1981 and of 1982, respectively. Before such circumstances and the relatively short amounts of fuel elements that should be produced for those two unique reactors, Mexico already has to adopt a strategy to follow in respect to fuel elements fabrication. The two main options are analyzed: 1. To delay the local fabrication until a National Nuclear Program may be defined, meanwhile purchasing abroad the necessary reloads and initial cores; and 2. To start as soon as possible the local fuel elements fabrication in order to supply fuel for the first reload of the first unit of Laguna Verde, confronting the economical risks of such posture with the advantages of an immediate action. Both options are analyzed in detail comparing them specially under the economic point of view, standing out immediately the big effect of some factors which are economically imponderable, as experience and independance that would be gained with the second option. Emphasis is made on the advantages and risks of any case. According to the first option and once a National Program is defined, the work would be heavy but of simple strategy. On the contrary, the second option requires the adoption of a more complicated strategy, as either the project of the factory as its initial operation should be made under transient conditions, in view of the expected future expansion still

  18. Remote real time x-ray examination of fuel elements in a hot cell environment

    International Nuclear Information System (INIS)

    Yapuncich, F.L.

    1993-01-01

    This report discusses the Remote Real Time X-ray System which will allow for detailed examination of fuel elements. This task will be accomplished in a highly radioactive hot cell environment. Two remote handling systems win be utilized at the examination station. One handling system will transfer the fuel element to and from the shielded x-ray system. A second handling system will allow for vertical and rotational inspection of the fuel elements. The process win include removing a single nuclear fuel element from a element fabrication magazine(EFM), positioning the fuel element within the shielding envelope of the x-ray system and transferring the fuel element from the station manipulator to the x-ray system manipulator, performing the x-ray inspection, and then transferring the fuel element to either the element storage magazine(ESM) or a reject bin

  19. 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

  20. Fuel Element Experience at the Halden Boiling Water Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Aas, S. [OECD Halden Reactor Project, Halden (Norway); Videm, K.; Hanevik, A. [Institutt for Atomenergi, Kjeller (Norway)

    1968-04-15

    The penalty for neutron absorbing materials is higher for a reactor moderated with heavy water than one with light water. As Zircaloy and enriched uranium were not readily available in 1954 when the design of the first fuel charge for HBWR was frozen, fuel elements of natural uranium metal clad in a specially developed aluminium alloy (A 1 0.3% Fe, 0.03% Si) were used. The temperature was limited to 150 Degree-Sign C and with this limitation the general behaviour of the elements was good. In I960, in another effort to maintain a good neutron economy, a couple of elements with as thin cladding as 0.25 mm A1S1 316, stainless steel with an unsegmented length of 2 m supported by wire grid spacers were tested. These elements with 1.5% enriched UO{sub 2} behaved satisfactorily at 150'C. Elements of a rather similar construction failed due to stress corrosion during the later operation at 230 'C. The reason for the different behaviour is probably the higher stresses in the cladding, due to the increased pressure, possibly combined with a short period with a high chloride content in the heavy water. The second fuel core with 1.5% enriched UO{sub 2} clad in Zircaloy-2 was installed in order to permit an increase in temperature to 230 Degree-Sign C and in power from 5 to 20 MW(th). The maximum burnup obtained is 11000 MWd/t and the maximum heat rating 375 W/cm with no fracture failure and practically no change in appearance according to the post-irradiation examination. One element was deliberately taken to burn-out conditions by throttling the water flow. After a series of burn-outs, the element finally failed because of over-temperature. The successful use of aluminium cladding at 150 Degree-Sign C mitiated an effort for making aluminium alloys suitable for normal power reactor operation. Promising properties were found for an alloy (designated IFA 3 aluminium) with A1 10% Si, 1% Ni, 1% Mg, 0.3% Fe + Ti. Despite increase in corrosion rate under heat transfer conditions

  1. Fuel elements assembling for the DON project exponential experience; Montaje de los elementos combustibles para la experiencia exponencial del proyecto DON

    Energy Technology Data Exchange (ETDEWEB)

    Anca Abati, R de

    1966-07-01

    It is described the fuel unit used in the DON exponential experience, the manufacturing installments and tools as well as the stages in the fabrication.These 74 elements contain each 19 cartridges loaded with synterized urania, uranium carbide and indium, gold, and manganese probes. They were arranged in calandria-like tubes and the process-tube. This last one containing a cooling liquid simulating the reactor organic. Besides being used in the DON reactor exponential experience they were used in critic essays by the substitution method in the French reactor AQUILON II. (Author) 6 refs.

  2. Consequences of metallic fuel-cladding liquid phase attack during over-temperature transient on fuel element lifetime

    International Nuclear Information System (INIS)

    Lahm, C.E.; Koenig, J.F.; Seidel, B.R.

    1990-01-01

    Metallic fuel elements irradiated in EBR-II at temperatures significantly higher than design, causing liquid phase attack of the cladding, were subsequently irradiated at normal operating temperatures to first breach. The fuel element lifetime was compared to that for elements not subjected to the over-temperature transient and found to be equivalent. 1 ref., 3 figs

  3. Fuel-to-cladding heat transfer coefficient into reactor fuel element

    International Nuclear Information System (INIS)

    Lassmann, K.

    1979-01-01

    Models describing the fuel-to-cladding heat transfer coefficient in a reactor fuel element are reviewed critically. A new model is developed with contributions from solid, fluid and radiation heat transfer components. It provides a consistent description of the transition from an open gap to the contact case. Model parameters are easily available and highly independent of different combinations of material surfaces. There are no restrictions for fast transients. The model parameters are fitted to 388 data points under reactor conditions. For model verification another 274 data points of steel-steel and aluminium-aluminium interfaces, respectively, were used. The fluid component takes into account peak-to-peak surface roughnesses and, approximatively, also the wavelengths of surface roughnesses. For minor surface roughnesses normally prevailing in reactor fuel elements the model asymptotically yields Ross' and Stoute's model for the open gap, which is thus confirmed. Experimental contact data can be interpreted in very different ways. The new model differs greatly from Ross' and Stoute's contact term and results in better correlation coefficients. The numerical algorithm provides an adequate representation for calculating the fuel-to-cladding heat transfer coefficient in large fuel element structural analysis computer systems. (orig.) [de

  4. Design of the Fuel Element for the RRR Reactor (Australia)

    International Nuclear Information System (INIS)

    Estevez, E.A.; Markiewicz, M.E.; Gerding, R.

    2003-01-01

    The supply to the Replacement Research Reactor ( RRR ) to Australia represents a technological goal for our country, as much for the designers and manufacturers of this irradiation facility ( Invap SE ), as well for the responsibles of the fuel elements ( FE ) design and the suppliers of the first core ( CNEA ).In relation with the FE, although the conceptual design and fabrication technology of the FE are similar to the just developed and qualified by CNEA ( plane plates MTR fuel type ), the characteristics of this new reactor imposes most severe operation conditions on them than in previous supplies.In that sense, two distinguishing characteristics deserve to be shown: a) The magnitude of the hydrodynamics loads acting on the FE due to the coolant ascendent flow direction, and mainly, the very high flow velocities between the fuel plates ( aproximately five times higher than which presents in others Argentine FE actually in operation. b) The use of U3Si2 as fuel material.CNEA has started a programme to qualify this type of fuel.As result of these higher loads under irradiations and with the objective to maintain the high reliability level reached by our FE ( very low failure rates ), it was necessary to introduce FE mechanical-structural design modifications respect to the ECBE or standard design version, and to verify these changes through hydrodynamics tests on a 1:1 scale prototype.In this paper it is described the mechanical-structural FE design with special emphasis in the innovatives aspects incorporated.The design criteria established in function of the solicitations and limitating effects present under irradiation conditions.Also, a brief description of the proposed programme to verify and evaluate this design is presented, including analytical and numerical calculus of stresses acting on the fuel plates and others FE components, pressure loss hydrodynamics tests and endurance essays

  5. Behavior of mixed-oxide fuel elements during the TOPI-1E transient overpower test

    International Nuclear Information System (INIS)

    Tsai, H.; Neimark, L.A.; Yamamoto, K.; Hirai, K.; Shikakura, S.

    1993-12-01

    A slow-ramp, extended overpower transient test was conducted on a group of nineteen preirradiated mixed-oxide fuel elements in EBR-II. During the transient two of the test elements with high-density fuel and tempered martensitic cladding (PNC-FMS) breached at an overpower of ∼75%. Fuel elements with austenitic claddings (D9, PNC316, and PNC150), many with aggressive design features and high burnups, survived the overpower transient and incurred little or no cladding strain. Fuel elements with annual fuel or heterogeneous fuel columns also behaved well

  6. Determination of heterogeneous medium parameters by single fuel element method

    International Nuclear Information System (INIS)

    Veloso, M.A.F.

    1985-01-01

    The neutron pulse propagation technique was employed to study an heterogeneous system consisting of a single fuel element placed at the symmetry axis of a large cylindrical D 2 O tank. The response of system for the pulse propagation technique is related to the inverse complex relaxation length of the neutron waves also known as the system dispersion law ρ (ω). Experimental values of ρ (ω) were compared with the ones derived from Fermi age - Diffusion theory. The main purpose of the experiment was to obtain the Feinberg-Galanin thermal constant (γ), which is the logaritmic derivative of the neutron flux at the fuel-moderator interface and a such a main input data for heterogeneous reactor theory calculations. The γ thermal constant was determined as the number giving the best agreement between the theoretical and experimental values of ρ (ω). The simultaneous determination of two among four parameters η,ρ,τ and L s is possible through the intersection of dispersion laws of the pure moderator system and the fuel moderator system. The parameters τ and η were termined by this method. It was shown that the thermal constant γ and the product η ρ can be computed from the real and imaginary parts of the fuel-moderator dispersion law. The results for this evaluation scheme showns a not stable behavior of γ as a function of frequency, a result not foreseen by the theoretical model. (Author) [pt

  7. The prediction problems of VVER fuel element cladding failure theory

    International Nuclear Information System (INIS)

    Pelykh, S.N.; Maksimov, M.V.; Ryabchikov, S.D.

    2016-01-01

    Highlights: • Fuel cladding failure forecasting is based on the fuel load history and the damage distribution. • The limit damage parameter is exceeded, though limit stresses are not reached. • The damage parameter plays a significant role in predicting the cladding failure. • The proposed failure probability criterion can be used to control the cladding tightness. - Abstract: A method for forecasting of VVER fuel element (FE) cladding failure due to accumulation of deformation damage parameter, taking into account the fuel assembly (FA) loading history and the damage parameter distribution among FEs included in the FA, has been developed. Using the concept of conservative FE groups, it is shown that the safety limit for damage parameter is exceeded for some FA rearrangement, though the limits for circumferential and equivalent stresses are not reached. This new result contradicts the wide-spread idea that the damage parameter value plays a minor role when estimating the limiting state of cladding. The necessary condition of rearrangement algorithm admissibility and the criterion for minimization of the probability of cladding failure due to damage parameter accumulation have been derived, for using in automated systems controlling the cladding tightness.

  8. The properties of spherical fuel elements and its behavior in the modular HTR

    International Nuclear Information System (INIS)

    Lohnert, G.H.; Ragoss, H.

    1985-01-01

    The reference fuel element for all future HTR applications in the Federal Republic of Germany as developed by NUKEM/HOBEG in the framework of the 'High temperature Fuel-Cycle Project' had to be scrutinised for its compatibility with all the other design principles of the modular HTR, or possibly for restrictions forced upon reactor layout. This reference fuel element can be characterized by the following features: moulded spherical fuel element of 60 mm in diameter with fuel free shell of 5 mm thickness, based on carbon matrix; low enriched uranium (U/Pu fuel cycle); UO 2 fuel kernels; TRISO coating (pyrocarbon and additional SiC layers)

  9. Improvement of Reactor Fuel Element Heat Transfer by Surface Roughness

    International Nuclear Information System (INIS)

    Kjellstroem, B.; Larsson, A.E.

    1967-04-01

    In heat exchangers with a limited surface temperature such as reactor fuel elements, rough heat transfer surfaces may give lower pumping power than smooth. To obtain data for choice of the most advantageous roughness for the superheater elements in the Marviken reactor, measurements were made of heat transfer and pressure drop in an annular channel with a smooth or rough test rod in a smooth adiabatic shroud. 24 different roughness geometries were tested. The results were transformed to rod cluster geometry by the method of W B Hall, and correlated by the friction and heat transfer similarity laws as suggested by D F Dipprey and R H Sabersky with RMS errors of 12.5 % in the friction factor and 8.1 % in the Stanton number. The relation between the Stanton number and the friction factor could be described by a relation of the type suggested by W Nunner, with a mean error of 3.1 % and an RMS error of 11.6 %. Application of the results to fuel element calculations is discussed, and the great gains in economy which can be obtained with rough surfaces are demonstrated by two examples

  10. Improvement of Reactor Fuel Element Heat Transfer by Surface Roughness

    Energy Technology Data Exchange (ETDEWEB)

    Kjellstroem, B; Larsson, A E

    1967-04-15

    In heat exchangers with a limited surface temperature such as reactor fuel elements, rough heat transfer surfaces may give lower pumping power than smooth. To obtain data for choice of the most advantageous roughness for the superheater elements in the Marviken reactor, measurements were made of heat transfer and pressure drop in an annular channel with a smooth or rough test rod in a smooth adiabatic shroud. 24 different roughness geometries were tested. The results were transformed to rod cluster geometry by the method of W B Hall, and correlated by the friction and heat transfer similarity laws as suggested by D F Dipprey and R H Sabersky with RMS errors of 12.5 % in the friction factor and 8.1 % in the Stanton number. The relation between the Stanton number and the friction factor could be described by a relation of the type suggested by W Nunner, with a mean error of 3.1 % and an RMS error of 11.6 %. Application of the results to fuel element calculations is discussed, and the great gains in economy which can be obtained with rough surfaces are demonstrated by two examples.

  11. The modeling experience of fuel element units operation under MSC.MARC and MENTAT 2008R1

    International Nuclear Information System (INIS)

    Kulakov, G.; Kashirin, B.; Kosaurov, A.; Konovalov, Y.; Kuznetsov, A.; Medvedev, A.; Novikov, V.; Vatulin, A.

    2009-01-01

    MSC Software is leading developer of CAE-software in the world, so behaviour of fuel elements modeling with MSC.MARC use is of great practical importance. Behaviour of fuel elements usually is modeled in the elastic-viscous-plastic statement with account on fuel swelling during irradiation. For container type fuel elements contact interaction between fuel pellets and cladding or other parts of fuel element in top and bottom plugs must be in account. Results of simulated behaviour of various type fuel elements - container type fuel elements for PWR and RBMK reactors, dispersion type fuel elements for research reactors are presented. (authors)

  12. Graphite behaviour in relation to the fuel element design

    Energy Technology Data Exchange (ETDEWEB)

    Everett, M. R. [OECD High Temperature Reactor Project Dragon, Winfrith (United Kingdom); Manzel, R. [OECD High Temperature Reactor Project Dragon, Winfrith (United Kingdom); Blackstone, R. [Reactor Centrum, Petten (Netherlands); Delle, W. [Kernforschungsanlage, Juelich (Germany); Lungagnani, V. [Joint Nuclear Research Centre, Euratom, Petten (Netherlands); Krefeld, R. [Joint Nuclear Research Centre, Euratom, Petten (Netherlands)

    1969-09-01

    The first designs of H.T.R. power reactors will probably use a Gilsocarbon based graphite for both the moderator/carrier blocks and for the fuel tubes. The initial physical properties and changes of dimensions, thermal expansion coefficient, Young*s modulus, and thermal conductivity on irradiation of Gilsocarbon graphites to typical reactor dwell-time fast neutron doses of 4 * 1021 cm -2 Ni dose Dido equivalent are given and values for the irradiation creep constant are presented. The influence of these property changes and those of chemical corrosion are considered briefly in relation to the present fuel element designs. The selection of an eventual less costly replacement graphite for Gilsocarbon graphite is discussed in terms of materials properties.

  13. Surface coating Zr or Zr alloy nuclear fuel elements

    International Nuclear Information System (INIS)

    Donaghy, R.E.; Sherman, A.H.

    1980-01-01

    A method is disclosed for preventing stress corrosion cracking or metal embrittlement of a zirconium or zirconium alloy container that is to be coated on the inside surface with a layer of a metal such as copper, a copper alloy, nickel, or iron and used for holding nuclear fuel material as a nuclear fuel element. The zirconium material is etched in an etchant solution, desmutted mechanically or ultrasonically, oxidized to form an oxide coating on the zirconium, cleaned in an aqueous alkaline cleaning solution, activated for electroless deposition of a metal layer and contacted with an electroless metal plating solution. This method provides a boundary layer of zirconium oxide between the zirconium container and the metal layer. (author)

  14. Actual Status of CAREM-25 Fuel Element Development

    International Nuclear Information System (INIS)

    Perez, Edmundo

    2000-01-01

    In the frame of the CAREM Project, under Cnea s Reactor and Nuclear Plants Program, the Nuclear Fuel Thematic Area is one among others on which the project is organized. In this area, the primary objective to reach is to actualize the mechanical fuel element and reactivity control designs, taking in account the recents conceptual and engineering modifications introduced in the reactor, and ending with a consolidated conceptual and basic development.In order to reach these objectives, it is presented the way on which the area was organized, the participating working groups, the task required, the personnel involucrated, the grade of global development reached in the areas of engineering, developments, fabrication and essays of design verification, and the found difficulties, the tasks under ejecution, just finished and necessaries to fulfill completely the objectives. Finally, it is possible to say that due to the work realized, the conceptual design of both components is finished and the basic design is under development

  15. Control and repair system for radioactive nuclear fuel elements

    International Nuclear Information System (INIS)

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

    1975-01-01

    Irradiated fuel, especially such containing Pu-239, are put in a shielding container for inspection or repair. This container consists of an inner and outer tube of, for example, stainless steel, between which there is a gap filled with water, mineral oil, or polyethylene. At the upper end of the shielding container a rotating sleeve is positioned, by means of a bearing. It contains, for instance, an access opening and an inspection opening which are shielded by means of plexiglass. The access hole is opened only for repair work. In oder to prevent radiation from escaping to the environment during withdrawal and inspection of the fuel elements a second shielding container or shielding tube may be put over the sleeve. (DG/PB) [de

  16. Store for radioactive waste and burnt-up fuel elements

    International Nuclear Information System (INIS)

    Spilker, H.; Rox, R.; Peschl, H.W.

    1985-01-01

    The invention concerns a concrete storage block in which there are several vertical storage and cooling ducts for radioactive waste and burnt-up fuel elements. The storage block is assembled from several square concrete blocks. Several vertical ducts are made in these. The square blocks are placed on a concrete baseplate. The aligned ducts of several square blocks placed above each other form storage and cooling ducts for tubular storage containers. An annular gap is left for cooling air between the outside wall of the storage containers and the inside wall of the storage and cooling ducts. (orig./HP) [de

  17. Determination of the hydrogen content of fuel elements

    International Nuclear Information System (INIS)

    Soare, M.; Petriu, F.; Toma, V.

    1995-01-01

    A new method and apparatus are reported for determination of the total hydrogen content by measurements on as-manufactured fuel elements, heated at prescribed temperature values between 200 degrees C and 600 degrees C. The method is based on the catalytic oxidation of the organic compounds and transformation of the hydrogen in the equivalent water quantity which is analysed by a special infrared detector. Different types of measurements for determination of the hydrogen content from graphite coating, UO 2 pellets and filling gas are presented. Also, experimental observation regarding water release and graphite thermal decomposition kinetic are discussed. (author)

  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. Storage ponds for fuel elements of nuclear reactors

    International Nuclear Information System (INIS)

    Kumpf, H.

    1981-01-01

    Heat exchangers are inserted in storage ponds for fuel elements of nuclear reactors, so that the heat to be removed is given up to an external coolant, without any radio-activity being emitted. The heat exchanger is a hollow body, which is connected to an air cooler, which works with a cooling circuit with natural circulation. A cooling pipe is enclosed in the hollow body, which forms a cooling circuit with forced flow with an open pond. One therefore obtains two successive separating walls for the external coolant. (orig.) [de

  20. Improved moulding material for addition to nuclear fuel particles to produce nuclear fuel elements

    International Nuclear Information System (INIS)

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

    1976-01-01

    A suggestion is made to improve the moulding materials used to produce carbon-contained nuclear fuel particles by a coke-reducing added substance. The nuclear fuel particles are meant for the formation of fuel elements for gas-cooled high-temperature nuclear reactors. The moulding materials are above all for the formation of coated particles which are burnt in situ in nuclear fuel element chambers out of 'green' nuclear fuel bodies. The added substance improves the shape stability of the particles forming and prevents a stiding or bridge formation between the particles or with the surrounding walls. The following are named as added substances: 1) Polystyrene and styrene-butadiene-Co polymers (mol. wt. between 5oo and 1,000,000), 2) aromatic compounds (mol. wt. 75 to 300), 3) saturated hydrocarbon polymers (mol. wt. 5,000 to 1,000,000). Additional release agents further improve the properties in the same direction (e.g. alcohols, fatty acids, amines). (orig.) [de