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Sample records for zircaloy-2 clad uranium

  1. Plastic deformation and fracture behavior of zircaloy-2 fuel cladding tubes under biaxial stress

    International Nuclear Information System (INIS)

    Maki, Hideo; Ooyama, Masatosi

    1975-01-01

    Various combinations of biaxial stress were applied on five batches of recrystallized zircaloy-2 fuel cladding tubes with different textures; elongation in both axial and circumferential directions of the specimen was measured continuously up to 5% plastic deformation. The anisotropic theory of plasticity proposed by Hill was applied to the resulting data, and anisotropy constants were obtained through the two media of plastic strain loci and plastic strain ratios. Comparison of the results obtained with the two methods proved that the plastic strain loci provide data that are more effective in predicting quantitatively the plastic deformation behavior of the zircaloy-2 tubes. The anisotropy constants change their value with progress of plastic deformation, and judicious application of the effective stress and effective strain obtained on anisotropic materials will permit the relationship between stress and strain under various biaxialities of stresses to be approximated by the work hardening law. The test specimens used in the plastic deformation experiments were then stressed to fracture under the same combination of biaxial stress as in the proceeding experiments, and the deformation in the fractured part was measured. The result proved that the tilt angle of the c-axis which serves as the index of texture is related to fracture ductility under biaxial stress. Based on this relationship, it was concluded that material with a tilt angle ranging from 10 0 to 15 0 is the most suitable for fuel cladding tubes, from the viewpoint of fracture ductility, at least in the case of unirradiated material. (auth.)

  2. Cladding the inside surface of a 3 1/4 in. ID Zircaloy-2 pressure tube with 1S aluminum

    International Nuclear Information System (INIS)

    Watson, R.D.

    1966-09-01

    A hot-press sizing technique has been developed for cladding the inside surface of Zircaloy-2 pressure tubes with 1S aluminum. The process is performed in air with the Zircaloy-2 and aluminum at a temperature of approximately 950 o F. A controlled atmosphere is not required, either during preheating or while the cladding is being applied. Tubes 30 inches long and 3 1/4 inches ID have been coated with 1S aluminum in thicknesses ranging from 0.005 inches to more than 0.02 inches; tubes longer than 30 inches have not been attempted. The lining of aluminum is firmly attached to the Zircaloy-2 at all points in the tube but the bond strength varies considerably - from. 6500 to 28000 lbf/in 2 . This work is the subject of Canadian Patent Application No. 955,358 filed March 21, 1966. (author)

  3. Influence of manufacturing process on the in-reactor creep anisotropy of stress-relieved Zircaloy-2 cladding

    International Nuclear Information System (INIS)

    Shann, S.H.; Van Swam, L.F.

    1995-01-01

    A procedure to determine the axial/radial and circumferential/radial contractile strain ratios (the R and P factors respectively in the Backofen-modified von Mises-Hill yield criterion) from post-irradiation dimensional measurements of Zircaloy-2 cladding of BWR fuel rods, tie rods and water rods was developed and has been described previously (S.H. Shann and L.F. van Swam, Creep anisotropy of Zircaloy-2 cladding during irradiation, Trans. SMiRT-11, Vol. C, 1991). The present study employs the procedure to determine the anisotropy factors R and P for textured cold-worked stress-relieved (CWSR) Zircaloy-2 cladding fabricated by various manufacturing processes. The analysis indicates that the cladding manufacturing process can have a pronounced effect on the anisotropy of irradiation-induced creep. Cladding types with identical yield and ultimate tensile strengths but fabricated by different manufacturing processes have different values of R and P during in-reactor creep. ((orig.))

  4. Measurements of delayed hydride cracking propagation rate in the radial direction of Zircaloy-2 cladding tubes

    Energy Technology Data Exchange (ETDEWEB)

    Kubo, T., E-mail: kubo@nfd.co.jp [Nippon Nuclear Fuel Development Co., Ltd., 2163 Narita-cho, Oarai-machi, Ibaraki 311-1313 (Japan); Kobayashi, Y. [M.O.X. Co., Ltd., 1828-520 Hirasu-cho, Mito, Ibaraki 311-0853 (Japan); Uchikoshi, H. [Nippon Nuclear Fuel Development Co., Ltd., 2163 Narita-cho, Oarai-machi, Ibaraki 311-1313 (Japan)

    2012-08-15

    Highlights: Black-Right-Pointing-Pointer The delayed hydride cracking (DHC) velocity of Zircaloy-2 was measured. Black-Right-Pointing-Pointer The velocity followed the Arrhenius law up to 270 Degree-Sign C. Activation energy was 49 kJ/mol. Black-Right-Pointing-Pointer The threshold stress intensity factor for the DHC was from 4 to 6 MPa m{sup 1/2}. Black-Right-Pointing-Pointer An increase in material strength accelerated the DHC. Black-Right-Pointing-Pointer Precipitation and fracture of hydrides at a crack tip is responsible for the DHC. - Abstract: Delayed hydride cracking (DHC) tests of Zircaloy-2 cladding tubes were performed in the chamber of a scanning electron microscope (SEM) to directly observe the crack propagation and measure the crack velocity in the radial direction of the tubes. Pre-cracks were produced at the outer surfaces of the tubes. Hydrogen contents of the tubes were from 90 ppm to 130 ppm and test temperatures were from 225 Degree-Sign C to 300 Degree-Sign C. The crack velocity followed the Arrhenius law at temperatures lower than about 270 Degree-Sign C with apparent activation energy of about 49 kJ/mol. The upper temperature limit for DHC, above which DHC did not occur, was about 280 Degree-Sign C. The threshold stress intensity factor for the initiation of the crack propagation, K{sub IH}, was from about 4 MPa m{sup 1/2} to 6 MPa m{sup 1/2}, almost independent of temperature. An increase in 0.2% offset yield stress of the material accelerated the crack velocity and slightly decreased K{sub IH}. Detailed observations of crack tip movement showed that cracks propagated in an intermittent fashion and the propagation gradually approached the steady state as the crack depth increased. The SEM observations also showed that hydrides were formed at a crack tip and a number of micro-cracks were found in the hydrides. It was presumed from these observations that the repetition of precipitation and fracture of hydrides at the crack tip would be

  5. Annealing studies of zircaloy-2 cladding at 580-8500C

    International Nuclear Information System (INIS)

    Hindle, E.D.

    1978-05-01

    For fuel element cladding it is important to determine if prior metallurgical condition combined with irradiation damage can influence high temperature deformation, because studies of such deformation are required to produce data for the cladding ballooning models which are used in analysing loss-of-coolant accidents (LOCA). If the behaviour of all cladding conditions during a LOCA can be represented by, say, the annealed condition, then much experimental work on a multiplicity of cladding conditions can be avoided. By examining the metallographic structure and hardness, the present study determines the time required in the range 580 to 850 0 C for returning Zircaloy cladding to the annealed condition, so that for any transient, a point can be specified where the material should have annealed. An equation has been derived to give this information. (author)

  6. Annealing studies of Zircaloy-2 cladding at 580-850 deg C

    International Nuclear Information System (INIS)

    Hindle, E.D.

    1983-01-01

    For fuel rod cladding it is important to determine if prior metallurgical condition combined with irradiation damage can influence high temperature deformation, because studies of such deformation are required to produce data for the cladding ballooning models which are used in analysing loss-of-coolant (LOCA). If the behaviour of all cladding conditions during a LOCA can be represented by, say, the annealed condition, then a great deal of experimental work on a multiplicity of cladding conditions can be avoided. By examining the metallographic structure and hardness, the present study determines the time required in the range 580 to 850 deg C for returning Zircaloy cladding to the annealed condition, so that for any transient a point can be specified where the material should have annealed. An equation has been derived to give this information. (author)

  7. Characterization of the Microstructure in Recrystallized Zircaloy-2 Cladding Irradiated to a High Neutron Dose

    International Nuclear Information System (INIS)

    Pettersson, Kjell

    2003-04-01

    The objectives of the present project were to determine if there is anything in the microstructure of highly irradiated Zircaloy-2 which may make the material fracture in a brittle manner. Samples were taken from three different locations on a fuel rod which had been irradiated for 12 years. The displacement doses were estimated to be 1.4, 9 and 28 dpa. Specimens for electron microscopy were prepared with two different orientation called axial and radial. In the axial orientation the electron beam goes parallel with the basal plane and diffraction conditions can be arranged so that dislocations with a Burgers' vectors become invisible. In the low dose specimen only a-component damage was present and all second phase particles were crystalline. In both the high and intermediate dose samples there was c-component damage present with a slightly higher amount in the high dose sample. The particles of the Zr(Cr,Fe) 2 type were generally amorphous in these samples and the Fe-content of the particles was highly reduced. The hydride structures were similar in all samples. The hydrides were often precipitated in parallel in the same grain and chains of hydrides were seen which ran from grain to grain. No population of small hydrides were observed except from surface hydrides formed during specimen preparation. It was concluded from the investigation that there is nothing in the microstructure which may make the material in the high dose state subject to a purely mechanically induced fast brittle cracking

  8. The maximum allowable temperature of zircaloy-2 fuel cladding under dry storage conditions

    International Nuclear Information System (INIS)

    Mayuzumi, M.; Yoshiki, S.; Yasuda, T.; Nakatsuka, M.

    1990-09-01

    Japan plans to reprocess and reutilise the spent nuclear fuel from nuclear power generation. However, the temporary storage of spent fuel is assuming increasing importance as a means of ensuring flexibility in the nuclear fuel cycle. Our investigations of various methods of storage have shown that casks are the most suitable means of storing small quantities of spent fuel of around 500 t, and research and development are in progress to establish dry storage technology for such casks. The soundness of fuel cladding is being investigated. The most important factor in evaluating soundness in storage under inert gas as currently envisaged is creep deformation and rupture, and a number of investigations have been made of the creep behaviour of cladding. The present study was conducted on the basis of existing in-house results in collaboration with Nippon Kakunenryo Kaihatsu KK (Nippon Nuclear Fuel Department Co.), which has hot lab facilities. Tests were run on the creep deformation behaviour of irradiated cladding, and the maximum allowable temperature during dry storage was investigated. (author)

  9. Fabrication and post-irradiation examination of a zircaloy-2 clad UO2-1.5 wt% PuO2 fuel pin irradiated in PWL, CIRUS

    International Nuclear Information System (INIS)

    Sah, D.N.; Sahoo, K.C.; Chatterjee, S.; Majumdar, S.; Kamath, H.S.; Ramachandran, R.; Bahl, J.K.; Purushottam, D.S.C.; Ramakumar, M.S.; Sivaramakrishnan, K.S.; Roy, P.R.

    1977-01-01

    A zircaloy-2 clad UO 2 -1.5 wt% PuO 2 fuel pin was fabricated at the Radiometallurgy Section of the Bhabha Atomic Research Centre, Bombay, for irradiation in the pressurised water loop in CIRUS. Requisite development work related to powder conditioning, blending, pressing and sintering parameters was carried out to meet the exacting fuel pellet specifications of CANDU fuel. The fuel pin ruptured while being irradiated in the pressurised water loop in CIRUS, after experiencing a low burn-up of 507 MWD/MTM and was subsequently examined at the Radiometallurgy Hot Cells Facility. The results showed that internal clad hydriding led to primary failure of the fuel pin. Subsequent ingress of the coolant water caused excessive swelling of the thermal insulating magnesia pellets located at the ends of the fuel column. The swelling of magnesia pellets caused severe rupturing of the fuel pin at the two ends. The delayed rupturing of the fuel pin at the upper end, caused the fuel column to be displaced downwards by 5.85mm. (author)

  10. Analysis of atomic distribution in as-fabricated Zircaloy-2 claddings by atom probe tomography under high-energy pulsed laser

    Energy Technology Data Exchange (ETDEWEB)

    Sawabe, T., E-mail: sawabe@criepi.denken.or.jp [Central Research Institute of Electric Power Industry (CRIEPI), Iwado Kita 2-11-1, Komae, Tokyo 201-8511 (Japan); Sonoda, T.; Kitajima, S. [Central Research Institute of Electric Power Industry (CRIEPI), Iwado Kita 2-11-1, Komae, Tokyo 201-8511 (Japan); Kameyama, T. [Tokai University, Department of Nuclear Engineering, Kitakaname 4-1-1, Hiratsuka, Kanagawa 259-1292 (Japan)

    2013-11-15

    The properties of second-phase particles (SPPs) in Zircaloy-2 claddings are key factors influencing the corrosion resistance of the alloy. The chemical compositions of Zr (Fe, Cr){sub 2} and Zr{sub 2}(Fe, Ni) SPPs were investigated by means of pulsed laser atom probe tomography. In order to prevent specimen fracture and to analyse wide regions of the specimen, the pulsed laser energy was increased to 2.0 nJ. This gave a high yield of average of 3 × 10{sup 7} ions per specimen. The Zr (Fe, Cr){sub 2} SPPs contained small amounts of Ni and Si atoms, while in Zr{sub 2}(Fe, Ni) SPPs almost all the Si was concentrated and the ratio of Zr: (Fe + Ni + Si) was 2:1. Atomic concentrations of the Zr-matrix and the SPPs were identified by two approaches: the first by using all the visible peaks of the mass spectrum and the second using the representative peaks with the natural abundance of the corresponding atoms. It was found that the change in the concentration between the Zr-matrix and the SPPs can be estimated more accurately by the second method, although Sn concentration in the Zr{sub 2}(Fe, Ni) SPPs is slightly overestimated.

  11. Effect of zinc injection on BWR fuel cladding corrosion. Pt. 1. Study on an accelerated corrosion condition to evaluate corrosion resistance of zircaloy-2 fuel cladding

    International Nuclear Information System (INIS)

    Kawamura, Hirotaka; Kanbe, Hiromu; Furuya, Masahiro

    2002-01-01

    Japanese BWR utilities have a plan to apply zinc injection to the primary coolant in order to reduce radioactivity accumulation on the structure. Prior to applying the zinc injection to BWR plants, it is necessary to evaluate the effect of zinc injection on corrosion resistance of fuel cladding. The objective of this report was to examine the accelerated corrosion condition for evaluation of BWR fuel cladding corrosion resistance under non-irradiated conditions, as the first step of a zinc injection evaluation study. A heat transfer corrosion test facility, in which a two phase flow condition could be achieved, was designed and constructed. The effects of heat flux, void fraction and solution temperature on BWR fuel cladding corrosion resistance were quantitatively investigated. The main findings were as follows. (1) In situ measurements using high speed camera and a void sensor together with one dimensional two phase flow analysis results showed that a two phase flow simulated BWR core condition can be obtained in the corrosion test facility. (2) The heat transfer corrosion test results showed that the thickness of the zirconium oxide layer increased with increasing solution temperature and was independent of heat flux and void fraction. The corrosion accelerating factor was about 2.5 times in the case of a temperature increase from 288degC to 350degC. (author)

  12. Plating on Zircaloy-2

    International Nuclear Information System (INIS)

    Dini, J.W.; Johnson, H.R.; Jones, A.

    1979-03-01

    Zircaloy-2 is a difficult alloy to coat with an adherent electroplate because it easily forms a tenacious oxide film in air and aqueous solutions. Procedures reported in the literature and those developed at SLL for surmounting this problem were investigated. The best results were obtained when specimens were first etched in either an ammonium bifluoride/sulfuric acid or an ammonium bifluoride solution, plated, and then heated at 700 0 C for 1 hour in a constrained condition. Machining threads in the Zircaloy-2 for the purpose of providing sites for mechanical interlocking of the plating also proved satisfactory

  13. Investigation of Zircaloy-2 oxidation model for SFP accident analysis

    Energy Technology Data Exchange (ETDEWEB)

    Nemoto, Yoshiyuki, E-mail: nemoto.yoshiyuki@jaea.go.jp [Japan Atomic Energy Agency, 2-4 Shirakata, Ohaza, Tokai-mura, Naka-gun, Ibaraki, 319-1195 (Japan); Kaji, Yoshiyuki; Ogawa, Chihiro; Kondo, Keietsu [Japan Atomic Energy Agency, 2-4 Shirakata, Ohaza, Tokai-mura, Naka-gun, Ibaraki, 319-1195 (Japan); Nakashima, Kazuo; Kanazawa, Toru; Tojo, Masayuki [Global Nuclear Fuel – Japan Co., Ltd., 2-3-1, Uchikawa, Yokosuka-shi, Kanagawa, 239-0836 (Japan)

    2017-05-15

    The authors previously conducted thermogravimetric analyses on Zircaloy-2 in air. By using the thermogravimetric data, an oxidation model was constructed in this study so that it can be applied for the modeling of cladding degradation in spent fuel pool (SFP) severe accident condition. For its validation, oxidation tests of long cladding tube were conducted, and computational fluid dynamics analyses using the constructed oxidation model were proceeded to simulate the experiments. In the oxidation tests, high temperature thermal gradient along the cladding axis was applied and air flow rates in testing chamber were controlled to simulate hypothetical SFP accidents. The analytical outputs successfully reproduced the growth of oxide film and porous oxide layer on the claddings in oxidation tests, and validity of the oxidation model was proved. Influence of air flow rate for the oxidation behavior was thought negligible in the conditions investigated in this study. - Highlights: •An oxidation model of Zircaloy-2 in air environment was developed. •The oxidation model was validated by the comparison with oxidation tests using long cladding tubes in hypothetical spent fuel pool accident condition. •The oxidation model successfully reproduced the typical oxidation behavior in air.

  14. Metallography of pitted aluminum-clad, depleted uranium fuel

    International Nuclear Information System (INIS)

    Nelson, D.Z.; Howell, J.P.

    1994-01-01

    The storage of aluminum-clad fuel and target materials in the L-Disassembly Basin at the Savannah River Site for more than 5 years has resulted in extensive pitting corrosion of these materials. In many cases the pitting corrosion of the aluminum clad has penetrated in the uranium metal core, resulting in the release of plutonium, uranium, cesium-137, and other fission product activity to the basin water. In an effort to characterize the extent of corrosion of the Mark 31A target slugs, two unirradiated slug assemblies were removed from basin storage and sent to the Savannah River Technology Center for evaluation. This paper presents the results of the metallography and photographic documentation of this evaluation. The metallography confirmed that pitting depths varied, with the deepest pit found to be about 0.12 inches (3.05 nun). Less than 2% of the aluminum cladding was found to be breached resulting in less than 5% of the uranium surface area being affected by corrosion. The overall integrity of the target slug remained intact

  15. Oxidation of zircaloy-2 in high temperature steam

    International Nuclear Information System (INIS)

    Ikeda, Seiichi; Ito, Goro; Ohashi, Shigeo

    1975-01-01

    Oxidation tests were conducted for zircaloy-2 in steam at temperature ranging from 900 to 1300 0 C to clarify its oxidation kinetics as a nuclear fuel cladding materials in case of a loss-of-coolant accident. The influence of maximum temperature and heating rate of the specimen on its oxidation rate in steam was investigated. The changes in mechanical properties of the specimens after oxidation tests are also studied. The results obtained were summarized as follows: (1) The weight of the specimen after oxidation in steam increased two times as the time required to reach the maximum temperature increased from 1 to 10 mins. (2) The kinetics of oxidation of zircaloy-2 in steam were not affected by the difference in the surface condition before test such as chemical polishing or pre-oxidation in steam. (3) The dominant growth of oxide film on the surface of zircaloy-2 was observed at the initial stage of oxidation in steam. However, the thickness of oxygen-rich solid solution layer under the film increased gradually with the progress of oxidation and the ratio of oxygen in oxide to that in solid solution has a constant value of 8:2. (4) The breakaway took place only in the specimen subjected to 900 0 C repeated heating. This penomenon was caused by the local growth of the oxide below a crack of the oxide film resulting from the reheating of the specimen. (5) The results of bending tests showed that the deflection until fracture of the specimen was smaller for the one heated at a higher temperature even if the weight increase was of the same order of magnitude for both specimens. (6) It was concluded that the ductility of zircaloy-2 decreased remarkably at a heating temperature in excess of 1100 0 C for more than 5 min. (auth.)

  16. Penetrate-leach dissolution of zirconium-clad uranium and uranium dioxide fuels

    International Nuclear Information System (INIS)

    Harmon, H.D.

    1975-01-01

    A new decladding-dissolution process was developed for zirconium-clad uranium metal and UO 2 fuels. The proposed penetrate-leach process consists of penetrating the zirconium cladding with Alniflex solution (2M HF--1M HNO 3 --1M Al(NO 3 ) 3 --0.1M K 2 Cr 2 O 7 ) and of leaching the exposed core with 10M HNO 3 . Undissolved cladding pieces are discarded as solid waste. Periodic HF and HNO 3 additions, efficient agitation, and in-line zirconium analyses are required for successful control of ZrF 4 and/or AlF 3 precipitation during the cladding-penetration step. Preliminary solvent extraction studies indicated complete recovery of uranium with 30 vol. percent tributyl phosphate (TBP) from both Alniflex solution and blended Alniflex-HNO 3 leach solutions. With 7.5 vol. percent TBP, high extractant/feed flow ratios and low scrub flows are required for satisfactory uranium recovery from Alniflex solution. Modified waste-handling procedures may be required for Alniflex waste, because it cannot be evaporated before neutralization and large quantities of solids are generated on neutralization. The effect of unstable UZr 3 (epsilon phase of uranium-zirconium system) on the safety of penetrate-leach dissolution was investigated

  17. Ductility loss of ion-irradiated zircaloy-2 in iodine

    International Nuclear Information System (INIS)

    Shimada, M.; Terasawa, M.; Yamamoto, S.; Kamei, H.; Koizumi, K.

    1981-01-01

    An ion bombardment simulation technique for neutron irradiation was applied to 'thick' materials to study the effect of radiation damage on the ductility change in Zircaloy-2 in an iodine environment. Specimens were prepared from actual cladding tubes and, prior to the irradiation, they were heat-treated in vacuo at 450, 580, and 700/degree/C for 2 h. Irradiation was performed by 52-MeV alpha particles up to the 0.32 displacements per atom (dpa) at 340/degree/C. Ductility loss begins to appear after 0.03 dpa irradiation, both in iodine and argon gas environments. The iodine presence resulted in ductility reduction, compared with the argon result in all irradiation dose ranges examined. The stress applied during irradiation caused ductility loss to commence at lower dosage than in the case of stress-free irradiation. These results are discussed in relation to the existing stress corrosion cracking models

  18. SSMS near surface analysis of B in irradiated Zircaloy-2: ion implantation standards as a calibration technique

    International Nuclear Information System (INIS)

    Christie, W.H.; Carter, J.A.; Eby, R.E.; Landau, L.; Musick, W.R.

    1980-01-01

    Purpose of this study was to determine the amount of 10 B contamination on the surface of Zircaloy-2 clad irradiated fuel elements that had been stored in an aqueous solution containing 5000 wt. ppM enriched B. SMSS indicated that the contamination was less than 0.06 μg/cm 2

  19. Tensile creep of beta phase zircaloy-2

    International Nuclear Information System (INIS)

    Burton, B.; Reynolds, G.L.; Barnes, J.P.

    1977-08-01

    The tensile creep and creep rupture properties of beta-phase zircaloy-2 are studied under vacuum in the temperature and stress range 1300-1550 K and 0.5-2 MN/m 2 . The new results are compared with previously reported uniaxial and biaxial data. A small but systematic difference is noted between the uniaxial and biaxial creep data and reasons for this discrepancy are discussed. (author)

  20. Examination of disks from the IPNS depleted uranium target

    International Nuclear Information System (INIS)

    Strain, R.V.; Carpenter, J.M.

    1995-10-01

    This report describes the results of examining the Zircaloy-2 clad depleted uranium disks from the Intense Pulse Neutron Source (IPNS) Target. That target operated from August, 1981 to June, 1988 and from September, 1991 to September, 1992 at 450 MeV, pulsing at 30 Hz with a time average proton current of about 15 microA. The target was removed from service when the presence of fission products ( 135 Xe) in the coolant cover gas indicated a failure in the Zircaloy-2 cladding. Altogether, the target had absorbed about 240 mA hours of proton current, and endured between 50,000 and 100,000 thermal cycles. The purpose of the examination was to assess the condition of the disks and determine the cause of the cladding failure. The results of visual, gamma ray scanning, and destructive metallurgical examination of two disks are described

  1. Electromigration of hydrogen in zircaloy-2

    International Nuclear Information System (INIS)

    Parmeswaran, P.; Kamachi Mudali, U.; Raghunathan, V.S.; Govinda Rajan, K.

    1989-01-01

    Electromigration is a purification technique for removing interstitial impurities from metals like Zr, Ti and Nb. It uses an electric field to induce migration of atoms from one end to other. This paper describes an attempt to purify zircaloy-2 of its hydrogen content by this technique. Resistivity measurement has been used to evaluate the change in impurity concentration that occurs during the process. Results indicate the movement of hydrogen atoms towards the cathode end. The value of the effective charge number, Z * , calculated from the results confirms hydrogen migration to the cathode aided by a positive wind force. (author). 6 refs., 5 figs

  2. Study of the Zircaloy-2 welding

    International Nuclear Information System (INIS)

    Rodriguez-Solano, R.; Jimenez Moreno, J. M.

    1968-01-01

    After a bibliographical analysis of the Zircaloy-2 properties its welding was approached. The selected procedure is the TIG (Tungsten Inert Gas) d.c. arc-welding under an inert atmosphere vessel. A detailed description of the equipment and characteristics is given. During the tests two types of argon were used: one with 96 ppm. Impurities, the other with 7 ppm- impurities. It is al so mentioned the welding in helium atmosphere. The contamination of the welding was evaluated through hardness testing. (Author) 3 refs

  3. Plastic strain accumulation during asymmetric cyclic loading of Zircaloy-2 at room temperature

    International Nuclear Information System (INIS)

    Rajpurohit, R.S.; Santhi Srinivas, N.C.; Singh, Vakil

    2016-01-01

    Asymmetric cyclic loading leads to accumulation of cyclic plastic strain and reduces the fatigue life of components. This phenomenon is known as ratcheting fatigue. Zircaloy-2 is a important structural material in nuclear reactors and used as pressure tubes and fuel cladding in pressurized light and heavy water nuclear reactors. Due to power fluctuations, these components experience plastic strain cycles in the reactor and their life is reduced due to strain cycles. Power fluctuations also cause asymmetric straining of the material and leads to accumulation of plastic strain. The present investigation deals with the effect of the magnitude of mean stress, stress amplitude and stress rate on hardening/softening behavior of Zircaloy-2 under asymmetric cyclic loading, at room temperature. It was observed that plastic strain accumulation increased with mean stress and stress amplitude; however, it decreased with stress rate. (author)

  4. Diffusionless bonding of aluminum to Zircaloy-2

    International Nuclear Information System (INIS)

    Watson, R.D.

    1965-04-01

    Aluminum can be bonded to zirconium without difficulty even when a thin layer of oxide is present on the surface of the zirconium . No detectable diffusion takes place during the bonding process. The bond layer can be stretched as much. as 8% without affecting the bond. The bond can be heated for 1000 hours at 260 o C (500 o F), and can be water quenched from 260 o C (500 o F) without any noticeable change in the bond strength. An extrusion technique has been devised for making transition sections of aluminum bonded to zirconium which can then be used to join these metals by conventional welding. Welding can be done close to the bond zone without seriously affecting the integrity of the bond. This method of bonding aluminum to Zircaloy-2 is covered by Canadian patent 702,438 January 26, 1965. (author)

  5. Thermal expansion studies on zircaloy-2

    International Nuclear Information System (INIS)

    Sivabharathy, M.; Senthilkumar, A.; Palanichamy, P.; Ramachandran, K.

    2016-01-01

    Zircaloy-2 and Zr-2.5% Nb alloys are widely used in the pressurized heavy water reactors (PHWR) as the material for the pressure tubes. The pressure tube operates at 573 K, 11 MPa internal pressures and is subjected to neutron flux of the order of 1013 n/cm 2 /s. These conditions lead to degradations in the pressure tube with respect to dimensional changes, deterioration in mechanical properties due to irradiation embrittlement, thereby reducing its flaw tolerance, the growth of existing flaws, which were too small or 'insignificant' at the time of installation. Physical and chemical properties of materials are also very essential in nuclear industry and the relations among them is of interest in the selection of materials when they are used in the design and manufacturing of devices particularly for atomic reactors.Studies on the relations between mechanical and thermal properties are of interest to the steel and metal industries as these would give useful information on the relation between hardness and thermal diffusivity (α) of steel. Jayakumar et al have already carried out the ultrasonic and metallographic investigations to see that all the heat-treated specimens retained essentially the martensite structure. In this present work, thermal expansion measurements on useful reactor material, Zircaloy-2 with different sample. Given a β-quenching treatment by heating to 1223 K and holding for 2 h, followed by water quenching. These specimens were then thermally aged for 1 h in the temperature range 473 to 973 K and air-cooled. For all samples, the thermal expansion was carried out and the results are correlated with ultrasonic measurements, metallographic and photoacoustic studies. (author)

  6. The anisotropic mechanical behaviour of zircaloy-2

    International Nuclear Information System (INIS)

    Ballinger, R.; Pelloux, R.M.

    1980-01-01

    Zirconium alloys used in the LWR industry crystallize in the hexagonal crystal structure below approximately 1136 K and many of the fabrication steps are performed below this temperature. The hexagonal structure possesses a limited number of slip systems and normal deformation processes result in extensive twinning. The twinning process results in the development of a fabrication texture, the type and extent of which is a function of the strain path used in the fabrication process. The texture which develops is important for two reasons. First, the texture at a given point in the fabrication process will determine the ease with which the next strain increment may be taken. Second, the texture of the completed part will have a significant effect on its in service performance because properties such as yield strength, creep strength, and fatigue and stress corrosion cracking resistance are a strong function of texture. Currently there is little data available concerning the evolution of textures as a function of strain path during the fabrication process of Zircaloy. Consequently this experimental investigation was conducted to determine the effect of textures on the mechanical behaviour of Zircaloy-2 with a primary emphasis on the evolution of texture during plastic deformation. (author)

  7. Fuel Retention Improvement at High Temperatures in Tungsten-Uranium Dioxide Dispersion Fuel Elements by Plasma-Spray Cladding

    Science.gov (United States)

    Grisaffe, Salvatore J.; Caves, Robert M.

    1964-01-01

    An investigation was undertaken to determine the feasibility of depositing integrally bonded plasma-sprayed tungsten coatings onto 80-volume-percent tungsten - 20-volume-percent uranium dioxide composites. These composites were face clad with thin tungsten foil to inhibit uranium dioxide loss at elevated temperatures, but loss at the unclad edges was still significant. By preheating the composite substrates to approximately 3700 degrees F in a nitrogen environment, metallurgically bonded tungsten coatings could be obtained directly by plasma spraying. Furthermore, even though these coatings were thin and somewhat porous, they greatly inhibited the loss of uranium dioxide. For example, a specimen that was face clad but had no edge cladding lost 5.8 percent uranium dioxide after 2 hours at 4750 dgrees F in flowing hydrogen. A similar specimen with plasma-spray-coated edges, however, lost only 0.75 percent uranium dioxide under the same testing conditions.

  8. Development and fabrication of seamless Aluminium finned clad tubes for metallic uranium fuel rods for research reactor

    International Nuclear Information System (INIS)

    Singh, A.K.; Hussain, M.M.; Jayachandran, N.K.; Abdulla, K.K.

    2012-01-01

    Natural uranium metal or its alloy is used as fuel in nuclear reactors. Usually fuel is clad with compatible material to prevent its direct contact with coolant which prevents spread of activity. One of the methods of producing fuel for nuclear reactor is by co-drawing finished uranium rods with aluminum clad tube to develop intimate contact for effective heat removal during reactor operation. Presently seam welded Aluminium tubes are used as clad for Research Reactor fuel. The paper will highlight entire fabrication process followed for the fabrication of seamless Aluminium finned tubes along with relevant characterisation results

  9. Study of the Zircaloy-2 welding; Estudio de la soldadura de Zircaloy-2

    Energy Technology Data Exchange (ETDEWEB)

    Rodriguez-Solano, R; Jimenez Moreno, J M

    1968-07-01

    After a bibliographical analysis of the Zircaloy-2 properties its welding was approached. The selected procedure is the TIG (Tungsten Inert Gas) d.c. arc-welding under an inert atmosphere vessel. A detailed description of the equipment and characteristics is given. During the tests two types of argon were used: one with 96 ppm. Impurities, the other with 7 ppm- impurities. It is al so mentioned the welding in helium atmosphere. The contamination of the welding was evaluated through hardness testing. (Author) 3 refs.

  10. Recent observations on the evolution of secondary-phase particles in zircaloy-2 under irradiation in a BWR to high burn-up

    International Nuclear Information System (INIS)

    Abolhassani, S.; Graber, T.; Gavillet, D.; Groeschel, F.

    2000-01-01

    The influence of radiation on the corrosion of the fuel claddings in a Light Water Reactor (LWR) has been the subject of many investigations, and different aspects of the overall phenomena have been studied by different techniques. Analysis of the evolution of Secondary-Phase Particles (SPPs) for different periods of immersion of the cladding in the reactor enables the rate of corrosion to the structure of the material to be correlated. In the case of Zircaloy-2 in a Boiling Water Reactor (BWR), SPPs are dissolved under irradiation, and their dissolution affects the rate of oxidation and other correlated phenomena. In recent studies, the Zircaloy-2 in claddings loaded in the Leibstadt BWR are analysed after one, three and five cycles. Results are presented, and give an account of the changes which occurred in the materials under irradiation. (authors)

  11. Recent observations on the evolution of secondary-phase particles in zircaloy-2 under irradiation in a BWR to high burn-up

    Energy Technology Data Exchange (ETDEWEB)

    Abolhassani, S.; Graber, T.; Gavillet, D.; Groeschel, F

    2000-07-01

    The influence of radiation on the corrosion of the fuel claddings in a Light Water Reactor (LWR) has been the subject of many investigations, and different aspects of the overall phenomena have been studied by different techniques. Analysis of the evolution of Secondary-Phase Particles (SPPs) for different periods of immersion of the cladding in the reactor enables the rate of corrosion to the structure of the material to be correlated. In the case of Zircaloy-2 in a Boiling Water Reactor (BWR), SPPs are dissolved under irradiation, and their dissolution affects the rate of oxidation and other correlated phenomena. In recent studies, the Zircaloy-2 in claddings loaded in the Leibstadt BWR are analysed after one, three and five cycles. Results are presented, and give an account of the changes which occurred in the materials under irradiation. (authors)

  12. A pneumatic bellows-driven setup for controlled-distance electrochemical impedance measurements of Zircaloy-2 in simulated BWR conditions

    International Nuclear Information System (INIS)

    Arilahti, E.; Bojinov, M.; Hansson-Lyyra, L.

    2004-01-01

    This paper describes a novel pneumatic bellows-driven arrangement designed for controlled distance electrochemistry (CDE) measurements. The feasibility of the new arrangement has been verified by performing contact electric impedance measurements to study corrosion of Zircaloy-2 in a re-circulation loop simulating the BWR conditions. Until now, the measurements have been carried out using a step-motor driven controlled-distance electrochemistry (CDE) arrangement. The electrical and electrochemical properties of the pre transition oxide on Zircaloy-2 determined from these measurements were in good agreement with those estimated from measurements with a step-motor driven CDE. Furthermore, the results indicate that the bellows-driven CDE device is less sensitive to the contact pressure variation than the step-motor driven arrangement. This property combined with the bellows driven displacement mechanism provides a clear advantage for future in-core corrosion studies of fuel cladding materials. (Author)

  13. Vapor corrosion of aluminum cladding alloys and aluminum-uranium fuel materials in storage environments

    International Nuclear Information System (INIS)

    Lam, P.; Sindelar, R.L.; Peacock, H.B. Jr.

    1997-04-01

    An experimental investigation of the effects of vapor environments on the corrosion of aluminum spent nuclear fuel (A1 SNF) has been performed. Aluminum cladding alloys and aluminum-uranium fuel alloys have been exposed to environments of air/water vapor/ionizing radiation and characterized for applications to degradation mode analysis for interim dry and repository storage systems. Models have been developed to allow predictions of the corrosion response under conditions of unlimited corrodant species. Threshold levels of water vapor under which corrosion does not occur have been identified through tests under conditions of limited corrodant species. Coupons of aluminum 1100, 5052, and 6061, the US equivalent of cladding alloys used to manufacture foreign research reactor fuels, and several aluminum-uranium alloys (aluminum-10, 18, and 33 wt% uranium) were exposed to various controlled vapor environments in air within the following ranges of conditions: Temperature -- 80 to 200 C; Relative Humidity -- 0 to 100% using atmospheric condensate water and using added nitric acid to simulate radiolysis effects; and Gamma Radiation -- none and 1.8 x 10 6 R/hr. The results of this work are part of the body of information needed for understanding the degradation of the A1 SNF waste form in a direct disposal system in the federal repository. It will provide the basis for data input to the ongoing performance assessment and criticality safety analyses. Additional testing of uranium-aluminum fuel materials at uranium contents typical of high enriched and low enriched fuels is being initiated to provide the data needed for the development of empirical models

  14. A tem investigation on intermetallic particles in zircaloy-2

    International Nuclear Information System (INIS)

    Sudarminto, Harini Sosiati; Kuwano, Noriyuki; Oki, Kensuke

    1996-01-01

    Tem investigation were conducted on the heat treated zircaloy-2 having the composition of Zr containing 1.6% Sn, 0.2% Fe, 0.1% Cr and 0.05% Ni (%wt) in order tostudy the characteristics of intermetallic particles related to the microstructural basis on the corrosion effect. Forged zircaloy-2 was annealed in the β-phase at 1050 C degrees for various isothermally in the α-phase region at 650 and 750 C degrees, followed by water quenching. The size precipates, the lower became their number. By increasing the annealing temperature, the growth of precipitates formed in this zircaloy-2 were of the Zr(Cr,Fe) 2 and Zr 2 (Fe,Cr,Ni) types. These kinds of precipitates and the ratios of Fe/Cr were independent of size and shape of precipitates and annealing time and temperature. (author), 16 refs, 2 tabs, 5 figs

  15. Evaluation of refractory-metal-clad uranium nitride and uranium dioxide fuel pins after irradiation for times up to 10 450 hours at 990 C

    Science.gov (United States)

    Bowles, K. J.; Gluyas, R. E.

    1975-01-01

    The effects of some materials variables on the irradiation performance of fuel pins for a lithium-cooled space power reactor design concept were examined. The variables studied were UN fuel density, fuel composition, and cladding alloy. All pins were irradiated at about 990 C in a thermal neutron environment to the design fuel burnup. An 85-percent dense UN fuel gave the best overall results in meeting the operational goals. The T-111 cladding on all specimens was embrittled, possibly by hydrogen in the case of the UN fuel and by uranium and oxygen in the case of the UO2 fuel. Tests with Cb-1Zr cladding indicate potential use of this cladding material. The UO2 fueled specimens met the operational goals of less than 1 percent cladding strain, but other factors make UO2 less attractive than low-density UN for the contemplated space power reactor use.

  16. Repository emplacement costs for Al-clad high enriched uranium spent fuel

    International Nuclear Information System (INIS)

    McDonell, W.R.; Parks, P.B.

    1994-01-01

    A range of strategies for treatment and packaging of Al-clad high-enriched uranium (HEU) spent fuels to prevent or delay the onset of criticality in a geologic repository was evaluated in terms of the number of canisters produced and associated repository costs incurred. The results indicated that strategies in which neutron poisons were added to consolidated forms of the U-Al alloy fuel generally produced the lowest number of canisters and associated repository costs. Chemical processing whereby the HEU was removed from the waste form was also a low cost option. The repository costs generally increased for isotopic dilution strategies, because of the substantial depleted uranium added. Chemical dissolution strategies without HEU removal were also penalized because of the inert constituents in the final waste glass form. Avoiding repository criticality by limiting the fissile mass content of each canister incurred the highest repository costs

  17. Evolution of deformation velocity in narrowing for Zircaloy 2

    Energy Technology Data Exchange (ETDEWEB)

    Cetlin, P R [Minas Gerais Univ., Belo Horizonte (Brazil). Dept. de Engenharia Metalurgica; Okuda, M Y [Goias Univ., Goiania (Brazil). Inst. de Matematica e Fisica

    1980-09-01

    Some studies on the deformation instability in strain shows that the differences in this instability may lead to localized narrowing or elongated narrowing, for Zircaloy-2. The variation of velocity deformation with the narrowing evolution is expected to be different for these two cases. The mentioned variation is discussed, a great difference in behavior having been observed for the case of localized narrowing.

  18. Heat transfer coefficient between UO2 and Zircaloy-2

    International Nuclear Information System (INIS)

    Ross, A.M.; Stoute, R.L.

    1962-06-01

    This paper provides some experimental values of the heat-transfer coefficient between UO 2 and Zircaloy-2 surfaces in contact under conditions of interfacial pressure, temperature, surface roughness and interface atmosphere, that are relevant to UO 2 /Zircaloy-2 fuel elements operating in pressurized-water power reactors. Coefficients were obtained from eight UO 2 / Zircaloy-2 pairs in atmospheres of helium, argon, krypton or xenon, at atmosphere pressure and in vacuum. Interfacial pressures were varied from 50 to 550 kgf/cm 2 while surface roughness heights were in the range 0.2 x 10 -4 to 3.5 x 10 -4 cm. The effect on the coefficients of cycling the interfacial pressure, of interface gas pressure and of temperature were examined. The experimental values of the coefficients were used to test the predictions of expressions for the heat-transfer between two solids in contact. For the particular UO 2 / Zircaloy-2 pairs examined, numerical values were assigned to several parameters that related the surface roughnesses to either the radius of solid/solid contact spots or to the mean thickness of the interface voids and that accounted for the imperfect accommodation of the void gas on the test surfaces. (author)

  19. Clad Treatment in KARMA Code and Library

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jeong-yeup; Lee, Hae-chan; Woo, Hae-seuk [KEPCO Nuclear Fuel Co., Daejeon (Korea, Republic of)

    2016-05-15

    Zirconium is the main components in clad materials. The subgroup parameters of zirconium were generated with effective cross section which obtained by using flux distribution in clad region. It decreases absorption reaction rate differences with reference MCNP results. Use of composite nuclide is acceptable to increase efficiency but should be limited to specific target composition. Therefore, the use of the composite nuclide of Zircaloy-2 should be limited when HANA clad material is used for clad. Either using explicit components or generating composite nuclide for HANA is suggested. This paper investigates the clad analysis model for KARMA whether current method is applicable to HANA clad material.

  20. Fracture of Zircaloy cladding by interactions with uranium dioxide pellets in LWR fuel rods. Technical report 10

    International Nuclear Information System (INIS)

    Smith, E.; Ranjan, G.V.; Cipolla, R.C.

    1976-11-01

    Power reactor fuel rod failures can be caused by uranium dioxide fuel pellet-Zircaloy cladding interactions. The report summarizes the current position attained in a detailed theoretical study of Zircaloy cladding fracture caused by the growth of stress corrosion cracks which form near fuel pellet cracks as a consequence of a power increase after a sufficiently high burn-up. It is shown that stress corrosion crack growth in irradiated Zircaloy must be able to proceed at very low stress intensifications if uniform friction effects are operative at the fuel-cladding interface, when the interfacial friction coefficient is less than unity, when a symmetric distribution of fuel cracks exists, and when symmetric interfacial slippage occurs (i.e., ''uniform'' conditions). Otherwise, the observed fuel rod failures must be due to departures from ''uniform'' conditions, and a very high interfacial friction coefficient and particularly fuel-cladding bonding, are means of providing sufficient stess intensification at a cladding crack tip to explain the occurrence of cladding fractures. The results of the investigation focus attention on the necessity for reliable experimental data on the stress corrosion crack growth behavior of irradiated Zircaloy, and for further investigations on the correlation between local fuel-cladding bonding and stress corrosion cracking

  1. TEM examination of irradiated zircaloy-2 pressure tube material

    International Nuclear Information System (INIS)

    Srivastava, D.; Tewari, R.; Dey, G.K.; Sharma, B.P.; Sah, D.N.; Banerjee, Suparna; Sahoo, K.C.

    2005-09-01

    In the present work, microstructure of the zircaloy-2 pressure tube material irradiated in the Indian Pressurized Heavy Water RAPP-1. Reactor (PHWR) has been examined for the first time using transmission electron microscope (TEM). The samples were obtained from a zircaloy-2 pressure tube, which had been in operation in the high flux region of Rajasthan Atomic Power Station Unit -1, for a period for 6.77 effective full power years (EFPYs) and expected to have a cumulative radiation damage of about 3 dpa. In this study irradiated microstructure has been characterized and compared it with the microstructure of the unirradiated pressure tube samples. The effect of irradiation on the hydriding behaviour is also studied. (author)

  2. Brittle-fracture potential of irradiated Zircaloy-2 pressure tubes

    Science.gov (United States)

    Huang, F. H.

    1993-12-01

    Neutron irradiation can degrade the fracture toughness of Zircaloy-2 and may cause highly irradiated reactor components of this material to fail in a brittle manner. The effects of radiation embrittlement on the structural integrity of N Reactor pressure tubes are studied by performing KIc and JIc fracture toughness testing on samples cut from the Zircaloy-2 tubes periodically removed from the reactor. A fluence of 6 × 10 25n/ m2 ( E > 1 MeV) reduced the fracture toughness of the material by 40 to 50%. The fracture toughness values appear to saturate at 260°C with fluences above 3 × 10 25n/ m2 ( E > 1 MeV), but continue to decline with increasing fluence at temperatures below 177°C. Present and previous results obtained from irradiated pressure tubes indicate that the brittle-fracture potential of Zircaloy-2 increases with decreasing temperature and increasing fluence. Fractographic examinations of the fracture surfaces of irradiated samples reveal that circumferential hydride formation significantly influenced fracture morphology by providing sites for easy crack nucleation and leaving deep cracks. However, the deep cracks created at the hydride platelets in specimens containing less than 220 ppm hydrogen are not believed to be the major cause of degradation in postirradiation fracture toughness.

  3. Diffusion in cladding materials

    International Nuclear Information System (INIS)

    Anand, M.S.; Pande, B.M.; Agarwala, R.P.

    1992-01-01

    Aluminium has been used as a cladding material in most research reactors because its low neutron absorption cross section and ease of fabrication. However, it is not suitable for cladding in power reactors and as such zircaloy-2 is normally used as a clad because it can withstand high temperature. It has low neutron absorption cross section, good oxidation, corrosion, creep properties and possesses good mechanical strength. With the passage of time, further development in this branch of science took place and designers started looking for better neutron economy and less hydrogen pickup in PHW reactors. The motion of fission products in the cladding material could pose a problem after long operation. In order to understand their behaviour under reactor environment, it is essential to study first the diffusion under normal conditions. These studies will throw light on the interaction of defects with impurities which would in turn help in understanding the mechanism of diffusion. In this article, it is intended to discuss the diffusion behaviour of impurities in cladding materials.(i.e. aluminium, zircaloy-2, zirconium-niobium alloy etc.). (author). 94 refs., 4 figs., 3 tabs

  4. Microstructure and crystallographic texture evolution during TIG welding of zircaloy-2 material

    International Nuclear Information System (INIS)

    Jha, S.K.; Singh, R.P.; Singh, V.K.; Ramanathan, R.; Samjdar, I.; Srivastava, D.; Tewari, R.; Dey, G.K.

    2005-01-01

    Zirconium and its alloys are extensively used as structural materials in nuclear reactors, because of better neutron economy, good corrosion resistance in water and good mechanical properties at operating temperature. Zircaloy-2 and zircaloy-4 are widely used in both pressurized water reactors (PWR) and boiling water reactors (BWR) as fuel cladding materials and as calandria tube and pressure tube materials in pressurized heavy water reactors (PHWR). The satisfactory performance and the life of the reactor components depend mainly upon their mechanical properties, corrosion properties and dimensional stability in the reactor condition, which are strong function of metallurgical parameters such as microstructure and texture. Therefore, for best performance of the reactor components these parameters are optimized during their fabrication. The microstructure and texture of the zircaloy-2 components are expected to get modified during the welding of the components. In this study the evolution of the microstructure and texture has been investigated as a function of the welding parameters. Heat input was varied the current and welding time. A variety of analytical techniques have been applied for the study on microstructure and texture of the welds. Optical microscopy and electron microscopy were used to evaluate the detailed microstructure. X-ray diffraction (XRD) was used investigate the crystallographic textures among the base metal, heat affected zone and fusion zone. Particular attention was focused on the determination of microtexture in weld by using electron backscatter diffraction (EBSD) technique. After that, an effort was put to compare the results of X-ray macro-texture and EBS-microtexture. (author)

  5. Thermal Characteristic Of AIMg2 Cladding And Fuel Plates Of U3Si2-Al With Various Uranium Loading

    International Nuclear Information System (INIS)

    Aslina, Br. G.; Suparjo; Aggraini, D.; Hasbullah, N.

    1998-01-01

    Thermal characteristic analyzed in this paper included linear expansion value, coefficient expansion, and enthalpy of cladding material fuel core and fuel plate of U 3 Si 2 -AI. Before analyzing, the fresh cladding of AIMg2 (without treatment) and the rolled AIMg2 were annealed at temperature of 425 o C for 1 hour, and the fuel plates of U 3 Si 2 -AI was prepared for various uranium loading of 0.9 - 3.6 - 4.2 - 4.8 and 5.2 g/cm 3 . Linear expansion nominal value and expansion coefficient were analyzed by using Dilatometer whereas enthalpy determination used Differential Thermal Analysis (DTA). The linear expansion and expansion coefficient analysis was performed to study the dimension cladding and of fuel plates during their stay in the reactor core, whereas determination of enthalpy was carried out to estimate the energy absorbed and released by fuel meat of U 3 Si 2 -AI to the cooling water through AlMg2 as a cladding. The result showed that the linear expansion and expansion coefficient of fresh AIMg2 cladding, rolled AIMg2 and fuel plates of U 3 Si 2 -AI are increased with the increase of temperature as well as the increase of uranium loading. The enthalpy measure showed that the enthalpy of fresh AIMg2 is smaller than that of rolled AIMg2 but melting temperature of fresh AIMg2 is greater than that of rolled AIMg2. The enthalpy of fuel plates and meat of U 3 Si 2 -AI is less than that of plates of U 3 Si 2 -AI. The enthalpy of fuel platers and meat of U 3 Si 2 -AI decrease with the increase of uranium loading. It is concluded that the fuel meat more reactive than fuel plates of U 3 Si 2 -AI

  6. Corrosion performance of new Zircaloy-2-based alloys

    International Nuclear Information System (INIS)

    Rudling, P.; Mikes-Lindbaeck, M.; Lethinen, B.; Andren, H.O.; Stiller, K.

    1994-01-01

    A material development project was initiated to develop a new zirconium alloy, outside the ASTM specifications for Zircaloy-2 and Zircaloy-4, with optimized hydriding and corrosion properties for both boiling water reactors and pressurized water reactors. A number of different alloys were manufactured. These alloys were long-term corrosion tested in autoclaves at 400 C in steam. Also, a 520 C/24 h steam test was carried out. The zirconium metal microstructure and the chemistry of precipitates were characterized by analytical electron microscopy. The metal matrix chemistry was determined by atom probe analysis. The paper describes the correlations between corrosion material performance and zirconium alloy microstructure

  7. Identification of the zirconium hydrides metallography in zircaloy-2

    International Nuclear Information System (INIS)

    Garcia Gonzalez, F.

    1968-01-01

    Technique for the Identification of the zirconium hydrides in metallographic specimens have been developed. Microhardness, quantitative estimation and relative orientation of the present hydrides as well as grain size determination of the different Zircaloy-2 tube specimens have also been made. The specimens used were corrosion- tested in water during various periods of time at 300 degree castrating, prior to the metallographic examination. Reference specimens, as received, and heavily hydride specimens in a hydrogen atmosphere at 800 degree centigrees, have been used in the previous stages of the work. No difficulties have been met in this early stage of acquaintanceship with the zirconium hydrides. (Author) 5 refs

  8. Effect of current density on the anodization of zircaloy-2

    International Nuclear Information System (INIS)

    Bhaskar Reddy, P.; Panasa Reddy, A.

    2005-01-01

    The effect of current density on the kinetics of anodization of Zircaloy-2 in 0.1 M potassium tartarate have been studied at various constant current densities ranging from 2 to 10 mA.cm -2 and at room temperature to investigate the exponential dependence of ionic current density on the field across the oxide. The rate of anodic film formation (dV/dt), the current efficiency the differential field of formation (F) and the ionic current density (i i ) were calculated. It was found that all these parameters were increased with increase of current density. The induction period was decreased with the increase of current density. It was also found that the plot of log (ionic current density) vs differential field gave fairly a linear relationship. The kinetic parameters, half jump distance (a) and height of the energy barrier (W) were calculated. (author)

  9. Deformation texture and microtexture development in zircaloy-2

    International Nuclear Information System (INIS)

    Vanitha, C.; Kiran Kumar, M.; Samajdar, I.; Vishvanathan, N.N.; Dey, G.K.; Tewari, R.; Srivastava, D.; Banerjee, S.

    2002-01-01

    In the present study, two starting materials used were as-cast Zircaloy-2 with random texture and the finished tube with relatively stronger starting texture. Specimens of the alloys were hot rolled to various strains at different temperature. The texture measurement was carried out and was represented in the form of Orientation Distribution Function which showed a sluggish texture development on high temperature deformation. In the case of as cast alloy with increase in strain at a constant deformation temperature, development in the texture was significant. Upon increasing the working temperature, rate of the overall texture development has been found to reduce. This could be due to reduced slip-twin activities, recovery or due to recrystallization. Microstructural and relative hardening studies were carried out for understanding the mechanisms of deformation texture developments at warm and hot working stages. In the case of finished tube having initially strong texture exhibited slower development in texture on warm and hot rolling. (author)

  10. Spectrochemical determination of impurities in zircaloy 2 and 4

    International Nuclear Information System (INIS)

    Paula Reino, L.C. de; Lordello, A.R.

    1987-06-01

    A method has been developed for the determination of Hf,Co,Mo,Pb,Ti,V,Al,Si,W,Cu,Mg,Mn,B and Cd in zircaloy 2 and 4. For hafnium determination 10% CuF 2 is added as spectrographic buffer on a previously oxidized zircaloy; the samples are loaded in a shallow cup electrode of Scribner Mullins type and excited in a direct current arc. The carrier distillation technique has been used for the other elements. Better results were obtained with 25% AgCl as carrier. The precision of the method varies from 4% for copper to 29% for boron but it does not exceed 17% for most elements. (Author) [pt

  11. Reaction diffusion in chromium-zircaloy-2 system

    International Nuclear Information System (INIS)

    Xiang Wenxin; Ying Shihao

    2001-01-01

    Reaction diffusion in the chromium-zircaloy-2 diffusion couples is investigated in the temperature range of 1023 - 1123 K. Scanning electron microscope (SEM) and energy dispersive spectrum (EDS) were used to measure the thickness of the reaction layer and to determine the Zr, Fe and Cr concentration penetrate profile in reaction layer, respectively. The growth kinetics of reaction layer has been studied and the results show that the growth of intermetallic compound is controlled by the process of volume diffusion as the layer growth approximately obeys the parabolic law. Interdiffusion coefficients were calculated using Boltzmann-Matano-Heumann model. Calculated interdiffusion coefficients were compared with those obtained on the condition that Cr dissolves in Zr and merely forms dilute solid solution. The comparison indicates that Cr diffuses in dilute solid solution is five orders of magnitude faster than in Zr(Fe, Cr) 2 intermetallic compound

  12. Thermal diffusion of hydrogen in zircaloy-2 containing hydrogen beyond terminal solid solubility

    International Nuclear Information System (INIS)

    Maki, Hideo; Sato, Masao.

    1975-01-01

    The thermal diffusion of hydrogen is one of causes of uneven hydride precipitation in zircaloy fuel cladding tubes that are used in water reactors. In the diffusion model of hydrogen in zircaloy, the effects of the hydride on the diffusibility of hydrogen has been regarded as negligibly small in comparison with that of hydrogen dissolved in the matrix. Contrary to the indications given by this model, phenomena are often encountered that cannot be explained unless hydride platelets have considerable ostensible diffusibility in zircaloy. In order to determine quantitatively the diffusion characteristics of hydrogen in zircaloy, a thermal diffusion experiment was performed with zircaloy-2 fuel cladding tubes containing hydrogen beyond the terminal solid solubility. In this experiment, a temperature difference of 20 0 --30 0 C was applied between the inside and outside surfaces of the specimen in a thermal simulator. To explain the experimental results, a modified diffusion model is presented, in which the effects of stress are introduced into Markowitz's model with the diffusion of hydrogen in the hydride taken into account. The diffusion equation derived from this model can be written in a form that ostensibly represents direct diffusion of hydride in zircaloy. The apparent diffusion characteristics of the hydride at around 300 0 C are Dsub(p)=2.3x10 5 exp(-32,000/RT), (where R:gas constant, T:temperature) and the apparent heat of transport Qsub(p) =-60,000 cal/mol. The modified diffusion model well explains the experimental results in such respects as reaches a steady state after several hours. (auth.)

  13. Microstructural examination of irradiated zircaloy-2 pressure tube material

    International Nuclear Information System (INIS)

    Srivastava, D.; Tewari, R.; Dey, G.K.; Sah, D.N.; Banerjee, S.

    2005-01-01

    Irradiation induced microstructural changes in Zr alloys strongly influence the creep, growth and mechanical properties of pressure tube material. Since dimensional changes and mechanical property degradation can limit the life of pressure tube, it is essential to study and develop an understanding of the microstructure produced by neutron irradiation, by examining samples taken from the irradiated components. In the present work, an effort has been made to examine, microstructure of the Zircaloy-2 pressure tube material irradiated in the Indian Pressurized Heavy Water Reactor (PHWR). The present work is a first step towards a comprehensive program of characterization of microstructure of reactor materials after irradiation to different fluence levels in power reactors. In this study, samples from a Zircaloy-2 pressure tube, which had been in operation in the high flux region of Rajasthan Atomic Power Station Unit 1, for a period for 6.77 effective full power years (EFPYs), have been prepared and examined. The samples selected from the tube are expected to have a cumulative radiation damage of about 3 dpa. Samples prepared from the off cuts of RAPS-1 pressure tubes were also studied for examining the unirradiated microstructure of the material. The samples were examined in a 200kV JEOL 2000 FX microscope. This paper presents the distinct features observed in irradiated sample and a comprehensive comparison of the microstructures of the unirradiated and irradiated material. The effect of annealing on the annihilation of the defects generated during irradiation has been also studied. The bright field micrographs revealed that microstructure of the irradiated samples was different in many respects from the microstructure of the unirradiated samples. The presence of defect structure in the form of loops etc could be seen in the irradiated sample. These loops were mostly c-type loops lying in the basal plane. The dissolution and redistribution of the precipitates were

  14. Effect of the aluminum flow pattern on the bonding of aluminum to oxidized Zircaloy-2

    International Nuclear Information System (INIS)

    Watson, R.D.; Lambert, J.P.

    1965-04-01

    The bonds produced when hot aluminum is allowed to flow smoothly from an extrusion die to the oxidized surface of a heated tube of Zircaloy-2 are consistently inferior to those produced with back-extruded flow. The difference is believed to be due to the reduction in, or elimination of, the oxide layer on the aluminum that comes in contact with the surface of the Zircaloy-2. This method of bonding aluminum to Zircaloy-2 is covered by Canadian patent 702,438 January 1965. (author)

  15. Delayed hydride cracking behavior for zircaloy-2 plate

    International Nuclear Information System (INIS)

    Mills, J.W.; Huang, F.H.

    1991-01-01

    The delayed hydride cracking (DHC) behaviour for Zircaloy-2 plate was characterized at temperatures ranging from 300 to 550 o F. Specimens with a longitudinal (T-L) orientation exhibited a classic two-stage DHC response. At K values slightly above the threshold level (K th ), crack-growth rates increased dramatically with increasing K values (stage I). The K th value was found to be 11 and 14 ksi√ in at 400 and 500 o F. At high K values (stage II), cracking rates were relatively insensitive to applied K levels. Stage II crack growth was a thermally activated process described by an Arrhenius-type relationship with an activation energy of 65 kJ/mol. This energy level agreed with the theoretical activation energy for hydrogen diffusion into the triaxial stress field ahead of a crack. Above a critical temperature (300 o F), an overtemperature cycle was required to initiate DHC. The magnitude of the thermal excursion required to initiate cracking was found to increase at higher test temperatures. Specimens with a transverse(L-T) orientation showed a very low sensitivity to DHC because of an unfavorable crystallographic orientation for hydride reorientation. Metallographic and fractographic examinations were performed to understand the DHC mechanism. (author)

  16. Ratcheting fatigue behavior of Zircaloy-2 at room temperature

    Energy Technology Data Exchange (ETDEWEB)

    Rajpurohit, R.S., E-mail: rsrajpurohit.rs.met13@iitbhu.ac.in [Department of Metallurgical Engineering, Indian Institute of Technology, Banaras Hindu University, Varanasi, 221005 (India); Sudhakar Rao, G. [Nuclear Energy and Safety Department, Paul Scherrer Institute, Villigen, CH-5232 (Switzerland); Chattopadhyay, K.; Santhi Srinivas, N.C.; Singh, Vakil [Department of Metallurgical Engineering, Indian Institute of Technology, Banaras Hindu University, Varanasi, 221005 (India)

    2016-08-15

    Nuclear core components of zirconium alloys experience asymmetric stress or strain cycling during service which leads to plastic strain accumulation and drastic reduction in fatigue life as well as dimensional instability of the component. Variables like loading rate, mean stress, and stress amplitude affect the influence of asymmetric loading. In the present investigation asymmetric stress controlled fatigue tests were conducted with mean stress from 80 to 150 MPa, stress amplitude from 270 to 340 MPa and stress rate from 30 to 750 MPa/s to study the process of plastic strain accumulation and its effect on fatigue life of Zircaloy-2 at room temperature. It was observed that with increase in mean stress and stress amplitude accumulation of ratcheting strain was increased and fatigue life was reduced. However, increase in stress rate led to improvement in fatigue life due to less accumulation of ratcheting strain. - Highlights: • Ratcheting strain accumulation occurred due to asymmetric cyclic loading. • Accumulation of ratcheting strain increased with mean stress and stress amplitude. • Ratcheting strain accumulation decreased with increase in stress rate. • With increase in mean stress and stress amplitude there was reduction in fatigue life. • Fatigue life is improved with increase in stress rate.

  17. Stress corrosion cracking behavior of zircaloy-2 in iodine environment

    International Nuclear Information System (INIS)

    Ikeda, Seiichi

    1983-01-01

    The effects of strain rates, iodine partial pressure and testing temperature on SCC behavior of zircaloy-2 in iodine environment were studied by means of slow strain rate technique (SSRT). SCC behavior of recrystallized specimens in iodine environment was remarkably influenced by the testing temperatures, and the susceptibility to SCC of specimens tested at 623 K was higher than that at 573 K. The susceptibility to SCC of recrystallized specimens increased with increasing iodine partial pressure at the lower strain rates of 4.2 x 10 -6 s -1 and 8.3 x 10 -7 s -1 . Cold worked specimens indicate no SCC failure in iodine environment regardless of strain rates, although those were tested only at 573 K. Fractographic observation revealed that SCC features of recrystallized specimens can be classified into two groups. One group, mostly specimens tested at 573 K, are characterized by the fact that cracks are initiated from corrosion pits. The other group are characterized by transgranuler SCC in the absence of pitting. This type of crack is found on specimens tested in environments containing more than 570 Pa iodine and seems to be produced by iodine embrittlement. (author)

  18. Crystal plasticity modeling of irradiation growth in Zircaloy-2

    Science.gov (United States)

    Patra, Anirban; Tomé, Carlos N.; Golubov, Stanislav I.

    2017-08-01

    A physically based reaction-diffusion model is implemented in the visco-plastic self-consistent (VPSC) crystal plasticity framework to simulate irradiation growth in hcp Zr and its alloys. The reaction-diffusion model accounts for the defects produced by the cascade of displaced atoms, their diffusion to lattice sinks and the contribution to crystallographic strain at the level of single crystals. The VPSC framework accounts for intergranular interactions and irradiation creep, and calculates the strain in the polycrystalline ensemble. A novel scheme is proposed to model the simultaneous evolution of both, number density and radius, of irradiation-induced dislocation loops directly from experimental data of dislocation density evolution during irradiation. This framework is used to predict the irradiation growth behaviour of cold-worked Zircaloy-2 and trends compared to available experimental data. The role of internal stresses in inducing irradiation creep is discussed. Effects of grain size, texture and external stress on the coupled irradiation growth and creep behaviour are also studied and compared with available experimental data.

  19. Characteristics and properties of cladding tubes for VVER-1000 higher Uranium content fuel rods

    International Nuclear Information System (INIS)

    Peregud, M.; Markelov, A.; Novikov, V.; Gusev, A.; Konkov, V.; Pimenov, Y.; Agapitov, V.; Shtutsa, M.

    2009-01-01

    To improve the fuel cycle economics and to further increase the VVER fuel usability the work programme is under way to design novel improved fuel, fuel rods and fuel assemblies. Longer FA operation time that is needed to increase the fuel burnup and the related design developments of novel fuel assemblies resulted not only in changing types and sizes of Zirconium items and fuel assembly components but also altered the requirements placed on their technical characteristics. To use fuel rods having a larger charge of fuel, to improve their behaviour in LOCA, to reduce fuel rod damage ability during assembling the work was carried out to perfect the characteristics of both the cladding (reduced wall thickness and more rigid tolerances for geometry) and its material. To meet the more rigid requirements for the geometry dimensions of cladding tubes an improved process flow sheet has been designed and employed for their fabrication and also the finishing treatment of tube surfaces has been improved. The higher and stable properties of the cladding materials were managed through using the special purity in terms of Hafnium Zirconium (not higher than 100 ppm Hf) as a base of the E110 alloy and maintaining within the valid specifications for the alloy the optimized contents of Oxygen and Iron at the levels of (600 - 990) ppm and (250 - 700) ppm, respectively. The work was under way in 2004 - 2008 years; during this period the technology and materials science solutions were mastered that were phased-in introduced into the production of the cladding tubes for the fuels loaded into the of the Kalinin NPP Unit 1

  20. Degradation resistant fuel cladding materials and manufacturing

    Energy Technology Data Exchange (ETDEWEB)

    Marlowe, M.O. [GE Nuclear Energy, Wilmington, NC (United States); Montes, J. [ENUSA, Madrid (Spain)

    1995-12-31

    GE has been producing the degradation resistant cladding (zirconium liner and zircaloy-2 surface larger) described here with the cooperation of its primary zirconium vendors since the beginning of 1994. Approximately 24 fuel reloads, or in excess of 250,000 fuel rods, have been produced using this material by GE. GE has also produced tubing for one reload of fuel that is currently being produced by its technology affiliate ENUSA. (orig./HP)

  1. Observations on deformation systems in zircaloy-2 deformed at room temperature

    International Nuclear Information System (INIS)

    Pettersson, K.; Bergqvist, H.

    1975-08-01

    Different polycrystalline samples of Zircaloy-2 with textures such that the c-axis of most of the grains are oriented near the sheet normal were subjected to loading conditions such that sheet thinning was accomplished. Metallography showed that no twinning was involved. Electron microscopy showed the presence of dislocations which were usually confined to deformation bands. With the help of stereo micrographs the most likely plane of slip was determined to be (1011). The possibility of slip as a means of breaking the oxide film in iodine induced stress corrosion cracking of Zircaloy-2 is briefly discussed. (author)

  2. The effect of oxide microstructure on kinetic transition in out-of-pile steam corrosion test for Zircaloy-2 and Nb-added Zircaloy-2

    Energy Technology Data Exchange (ETDEWEB)

    Nanikawa, Shuichi [Japan Nuclear Fuel Co. Ltd., Yokosuka, Kanagawa (Japan); Etoh, Yoshinori [Japan Nuclear Fuel Co. Ltd., Yokohama, Kanagawa (Japan)

    2001-06-01

    In order to study the mechanism of kinetic transition of corrosion rate for zirconium alloys, oxide films formed on Zircaloy-2 (Zry-2) and Nb-added Zircaloy-2 (0.5Nb/Zry-2) in steam at 673 K and 10.3 MPa were examined with TEM and SIMS. Kinetic transition occurred at almost the same oxide thicknesses for both Zry-2 and 0.5Nb/Zry-2, but the corrosion rate after the transitions were quite different for the two alloys. Zircaloy-2 showed cyclical oxidation, while the weight gain of 0.5Nb/Zry-2 increased linearly. The morphology and crystal structure were similar for the oxides of the two alloys and both the oxide films still mainly consisted of columnar grains even after the transition. Interface layers which mainly consisted of {alpha}-Zr crystallites were observed for both alloys and the oxygen content in the interface layers increased after the transition. The solute concentrations of Fe, Cr and Ni became higher, accompanying the increase of oxygen concentrations at columnar grain boundaries in the oxide films after the transition for 0.5Nb/Zry-2. It was thought that the properties of grain boundaries of the 0.5Nb/Zry-2 oxide films changed after the transition, and the increase in oxygen diffusivity at grain boundaries caused the linear increase in weight gain. (author)

  3. Influence of texture on fracture toughness of zircaloy cladding

    International Nuclear Information System (INIS)

    Grigoriev, V.; Andersson, Stefan

    1997-06-01

    The correlation between texture and fracture toughness of Zircaloy 2 cladding has been investigated in connection with axial cracks in fuel rods. The texture of the cladding determines the anisotropy of plasticity of the cladding which, in turn, should influence the strain conditions at the crack-tip. Plastic strains in the cladding under uniaxial tension were characterised by means of the anisotropy constants F, G and H calculated according to Hill's theory. Test temperatures between 20 and 300 deg C do not influence the F, G and H values. Any significant effect of hydrogen (about 500 wtppm) on the anisotropy constants F, G and H has not been revealed at a test temperature of 300 deg C. The results, obtained for stress-relieved and recrystallized cladding with different texture, show an obvious influence of texture on the fracture toughness of Zircaloy cladding. A higher fracture toughness has been found for cladding with more radial texture

  4. Extrusion and drawing of zircaloy 2. Production of pressure tubes for EL-4; Filage et etirage du zircaloy 2. Realisation des tubes de force pour EL-4

    Energy Technology Data Exchange (ETDEWEB)

    Thevenet, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Buffet, J [Cefilac (France)

    1964-07-01

    The authors give briefly the physical mechanical and chemical properties of zircaloy 2, as far as the transformation of this alloy is concerned. Extrusion: After a few general remarks concerning the extrusion and co-extrusion, including a comparison of the deformation resistance of canning metals and of zircaloy 2, the following points are considered: - the difficulties occurring because of the use of this alloy: - atmosphere protection - adjustment on to the machine tools - low thermal conductivity - economy of the metal (price) - the factors affecting the quality of the extruded products extrusion under a copper can and under lubricant glass - fine grain structure - temperature homogeneity - working temperature The transformation cycle - '550 kg ingot - preliminary shape 'for drawing of EL-4 tubes (112 x 120 L 12 m)' - is described in detail (extrusion or forging of the {phi} = 340 ingot into {phi} = 220 billets, cutting into lengths and hot drilling at {phi} = 125, fixing into a copper can and rough extrusion). Drawing: The main difficulties are due to seizing of the tools and to the necessity of protecting the alloy from the atmosphere during annealings. A brief description is given of drawing out on a short mandrel, on a long mandrel, of laminating on a reducing machine and of the carrying out of an annealing, as well as of the production of EL-4 tubes ({phi} =107 x 113 L 430 m) by drawing out shapes having a size of 112 x 120 on long mandrels. Conclusion: It is possible by extrusion and drawing to produce zircaloy 2 tubes similar to those which may be obtained normally using stainless steel. (authors) [French] Les auteurs donnent un resume succint des proprietes physiques mecaniques et chimiques du zircaloy 2 en ce qui concerne la transformation de cet alliage. Filage: Apres quelques generalites sur le filage et le cofilage, dont une comparaison entre les resistances a la deformation des metaux de gainage et du zircaloy 2, on etudie successivement: - les

  5. Identification of the zirconium hydrides metallography in zircaloy-2; Contribucion al estudio por metalografia de los hidruros de circonio en Zircaloy-2

    Energy Technology Data Exchange (ETDEWEB)

    Garcia Gonzalez, F

    1968-07-01

    Technique for the Identification of the zirconium hydrides in metallographic specimens have been developed. Microhardness, quantitative estimation and relative orientation of the present hydrides as well as grain size determination of the different Zircaloy-2 tube specimens have also been made. The specimens used were corrosion- tested in water during various periods of time at 300 degree castrating, prior to the metallographic examination. Reference specimens, as received, and heavily hydride specimens in a hydrogen atmosphere at 800 degree centigrees, have been used in the previous stages of the work. No difficulties have been met in this early stage of acquaintanceship with the zirconium hydrides. (Author) 5 refs.

  6. The Determination of Composite Elements in Zircaloy-2 by X-Ray Fluorescence and Emission Spectrometry Method

    International Nuclear Information System (INIS)

    Dian Anggraini; Rosika Kriswarini; Yusuf N

    2007-01-01

    Analysis of composing elements in zircaloy-2 has been done by Emission Spectrometry method and X-Ray Fluorescence (XRF). The aim of the analysis is to verify conformity between composing elements in zircaloy-2 and the material certificate. Spectrometry Emission method has higher sensitivity in element determination of a material than that of XRF method, so can be estimated that emission spectrometry method has higher accuracy than that of XRF method. The result of qualitative analysis by Emission Spectrometry indicate that the composing elements in zircaloy-2 were Sn, Cr and Ni. However, the qualitative analysis result by XRF method indicated that the composing elements in zircaloy 2 were Sn, Cr, Ni and Fe. Fe element can not be analysed by Emission Spectrometry method because Emission Spectrometer did not equipped with Fe detector. The quantitative analysis result of the composing elements in the material with both methods showed that Sn, Cr and Ni concentration of zircaloy 2 existed in concentration ranges of the material certificate. Result of statistical test (F and t-test) of analysis result of both methods can be used for analyzing composing elements in zircaloy 2. Emission Spectrometry method was more sensitive and accurate for determining Cr and Ni element in zircaloy 2 than that of emission Spectrometry method but both methods had same accuracy. The precision of measurement of Sn, Cr and Ni element using XRF method was better than that of Emission spectrometry method. (author)

  7. Fuel cladding behavior under rapid loading conditions

    Science.gov (United States)

    Yueh, K.; Karlsson, J.; Stjärnsäter, J.; Schrire, D.; Ledergerber, G.; Munoz-Reja, C.; Hallstadius, L.

    2016-02-01

    A modified burst test (MBT) was used in an extensive test program to characterize fuel cladding failure behavior under rapid loading conditions. The MBT differs from a normal burst test with the use of a driver tube to simulate the expansion of a fuel pellet, thereby producing a partial strain driven deformation condition similar to that of a fuel pellet expansion in a reactivity insertion accident (RIA). A piston/cylinder assembly was used to pressurize the driver tube. By controlling the speed and distance the piston travels the loading rate and degree of sample deformation could be controlled. The use of a driver tube with a machined gauge section localizes deformation and allows for continuous monitoring of the test sample diameter change at the location of maximum hoop strain, during each test. Cladding samples from five irradiated fuel rods were tested between 296 and 553 K and loading rates from 1.5 to 3.5/s. The test rods included variations of Zircaloy-2 with different liners and ZIRLO, ranging in burn-up from 41 to 74 GWd/MTU. The test results show cladding ductility is strongly temperature and loading rate dependent. Zircaloy-2 cladding ductility degradation due to operational hydrogen pickup started to recover at approximately 358 K for test condition used in the study. This recovery temperature is strongly loading rate dependent. At 373 K, ductility recovery was small for loading rates less than 8 ms equivalent RIA pulse width, but longer than 8 ms the ductility recovery increased exponentially with increasing pulse width, consistent with literature observations of loading rate dependent brittle-to-ductile (BTD) transition temperature. The cladding ductility was also observed to be strongly loading rate/pulse width dependent for BWR cladding below the BTD temperature and Pressurized Water Reactor (PWR) cladding at both 296 and 553 K.

  8. Measurements of the effective total and resonance absorption cross sections for zircaloy-2 and zirconium

    Energy Technology Data Exchange (ETDEWEB)

    Kocic, A; Markovic, V [Institute of nuclear sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1967-04-15

    Zirconium and zircaloy-2 alloy, as constructive materials, have found wide application in reactor technology, especially in heavy water systems for two reasons: a) low neutron absorption cross section, b) good mechanical properties. The thickness of the zirconium and zircaloy-2 for different applications varies from several tenths of a millimeter to about ten millimeters. Therefore, to calculate reactor systems it is desirable to know the effective neutron absorption cross section for the range of thicknesses mention above. The thermal neutron cross sections for these materials are low and no appreciable variation of the effective neutron cross section occurs even for the largest thicknesses. However, this is not true for effective resonance absorption. On the other hand, due to the lack of detailed knowledge of the zirconium resonances, calculations of the effective resonance integrals cannot be performed. Therefore it is necessary to measure the effective total and resonance absorption cross section for zirconium (author)

  9. Investigation of microstructure and mechanical properties of proton irradiated Zircaloy 2

    Energy Technology Data Exchange (ETDEWEB)

    Sarkar, Apu, E-mail: asarkar@barc.gov.in [Mechanical Metallurgy Division, Bhabha Atomic Reserch Centre, Mumbai, 400 085 (India); Kumar, Ajay [Nuclear Physics Division, Bhabha Atomic Reserch Centre, Mumbai, 400 085 (India); Mukherjee, S.; Sharma, S.K.; Dutta, D.; Pujari, P.K. [Radiochemistry Division, Bhabha Atomic Research Centre, Mumbai, 400 085 (India); Agarwal, A.; Gupta, S.K.; Singh, P. [Ion Accelerator Development Division, Bhabha Atomic Research Centre, Mumbai, 400 085 (India); Chakravartty, J.K. [Mechanical Metallurgy Division, Bhabha Atomic Reserch Centre, Mumbai, 400 085 (India)

    2016-10-15

    Samples of Zircaloy 2 have been irradiated with 4 MeV protons to two different doses. Microstructures of the unirradiated and irradiated samples have been characterized by Electron Back Scatter Diffraction (EBSD), X-ray diffraction line profile analysis (XRDLPA), Positron Annihilation Lifetime Spectroscopy (PALS) and Coincident Doppler Broadening (CDB) Spectroscopy. Tensile tests and micro hardness measurements have been carried out at room temperature to assess the changes in mechanical properties of Zircaloy 2 due to proton irradiation. The correlation of dislocation density, grain size and yield stress of the irradiated samples indicated that an increase in dislocation density due to irradiation is responsible for the change in mechanical behavior of irradiated Zircaloy.

  10. The formation and characteristics of hydride blisters in c.w. Zircaloy-2 pressure tubes

    Energy Technology Data Exchange (ETDEWEB)

    Price, E G [ed.

    1994-09-01

    Under the auspices of the IAEA, a consultants` meeting was arranged in Vienna, 1994 July 25-29, at which a Canadian delegation, consisting of AECL and Ontario Hydro Technologies personnel, presented information on their knowledge of the behaviour of hydride blisters in Zircaloy-2 pressure tubes. This document contains the 10 papers presented by the Canadian delegation to the meeting. It is believed that they represent a good reference document on hydride blister phenomena.

  11. Experimental determination of resonance absorption cross sections for Zircaloy-2 and zirconium

    Energy Technology Data Exchange (ETDEWEB)

    Kocic, A; Markovic, V [Boris Kidric Institute of Nuclear Sciences, Vinca, Beograd (Yugoslavia)

    1968-05-15

    The integral absorption cross section for the neutron spectrum and the thermal absorption cross section for zircaloy-2 have been determined using the pile oscillator technique. Using both values and a measured ratio of the epithermal to the thermal flux, the effective resonance integrals were obtained. After subtraction of the contributions for alloy and impurity elements, the effective resonance integrals for zirconium were evaluated. An extrapolated value of 0.91{+-}0.10 was obtained for the dilute integral. (author)

  12. Observations on the ductility of zircaloy-2 under simultaneous tension and bending

    International Nuclear Information System (INIS)

    Pettersson, K.

    1975-01-01

    The ductility of Zircaloy-2 in creep-fatigue interaction tests has been found to exceed the ductility in separate tensile tests. It was shown that the increase of ductility was due to either the suppression of the localized shear band instability causing final failure in a tensile test, or because the hydrostatic tension-shear stress ratio in the creep-fatigue test is lower than in the tensile test. Possible applications of the ductility increase in forming operations are suggested. (author)

  13. Dislocation Arrangements in Deformed and Neutron Irradiated Zirconium and Zircaloy-2

    International Nuclear Information System (INIS)

    Roy, R.B.

    1963-12-01

    Dislocation arrangements in deformed and neutron irradiated Zr and Zircaloy-2 have been studied by thin film transmission electron microscopy. Results indicate that the prominent slip system, in both Zr and Zircaloy-2, is the {1010} 1/3 type; no evidence for basal slip was observed. Attractive and repulsive dislocation interactions seem to be more important than the intersection jog reactions. Elongated loops and dipoles were seen at higher deformations and it is suspected that such loops or dipoles are formed due to interactions between dislocations lying in parallel planes. Stacking fault ribbons lying in {1010} plane have been found in 15% cold rolled Zircaloy-2: a rough estimate of stacking fault energy indicates that it is ∼ 65 ergs/cm 2 . Calculations show that the equilibrium separation of partials is ∼ 60 A and a stress as high as 19x10 -3 μ acting along {0010} direction is needed to separate them. It has been suggested that O 2 and N 2 in addition to their solid solution hardening effect may also cause a lowering of the stacking fault energy and Suzuki hardening

  14. Dislocation Arrangements in Deformed and Neutron Irradiated Zirconium and Zircaloy-2

    Energy Technology Data Exchange (ETDEWEB)

    Roy, R B

    1963-12-15

    Dislocation arrangements in deformed and neutron irradiated Zr and Zircaloy-2 have been studied by thin film transmission electron microscopy. Results indicate that the prominent slip system, in both Zr and Zircaloy-2, is the {l_brace}1010{r_brace} 1/3 <1210> type; no evidence for basal slip was observed. Attractive and repulsive dislocation interactions seem to be more important than the intersection jog reactions. Elongated loops and dipoles were seen at higher deformations and it is suspected that such loops or dipoles are formed due to interactions between dislocations lying in parallel planes. Stacking fault ribbons lying in {l_brace}1010{r_brace} plane have been found in 15% cold rolled Zircaloy-2: a rough estimate of stacking fault energy indicates that it is {approx} 65 ergs/cm{sup 2}. Calculations show that the equilibrium separation of partials is {approx} 60 A and a stress as high as 19x10{sup -3} {mu} acting along {l_brace}0010{r_brace} direction is needed to separate them. It has been suggested that O{sub 2} and N{sub 2} in addition to their solid solution hardening effect may also cause a lowering of the stacking fault energy and Suzuki hardening.

  15. Stress corrosion testing of irradiated cladding tubes

    International Nuclear Information System (INIS)

    Lunde, L.; Olshausen, K.D.

    1980-01-01

    Samples from two fuel rods with different cladding have been stress corrosion tested by closed-end argon-iodine pressurization at 320 0 C. The fuel rods with stress relieved and recrystallized Zircaloy-2 had received burnups of 10.000 and 20.000 MWd/ton UO 2 , respectively. It was found that the SCC failure stress was unchanged or slightly higher for the irradiated than for the unirradiated control tubes. The tubes failed consistently in the end with the lowest irradiation dose. The diameter increase of the irradiated cladding during the test was 1.1% for the stress-relieved samples and 0.24% for the recrystallized samples. SEM examination revealed no major differences between irradiated and unirradiated cladding. A ''semi-ductile'' fracture zone in recrystallized material is described in some detail. (author)

  16. Superficial characterization and zircaloy-2 electrochemistry with hydrothermal deposit of platinum; Caracterizacion superficial y electroquimica de zircaloy-2 con deposito hidrotermal de platino

    Energy Technology Data Exchange (ETDEWEB)

    Contreras R, A.; Arganis J, C. R.; Medina A, A. L. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Gris C, M. M., E-mail: aida.contreras@inin.gob.mx [Comision Federal de Electricidad, Central Nucleoelectrica Laguna Verde, Carretera Cardel-Nautla Km 42.5, Alto Lucero, Veracruz (Mexico)

    2011-11-15

    The combustible elements of the boiling water nuclear reactors (BWR) are formed by zircaloy-2 tubes that contain in their interior UO{sub 2} pellets. With the objective of mitigating the speed of crack growth by IGSCC to a minimum negative impact on the BWR operation, General Electric developed the noble metals chemical addition (NMCA), in where noble metals particles as Pt, Pd, and Rh, are deposited on the surface of the metal to catalyze the recombination of H{sub 2} and O{sub 2}. Hydrogen is also injected to have it in excess and to favor this recombination (HWC) and zinc to reduce dose. In this work was oxidized zircaloy-2 low similar conditions to the HWC, platinum was deposited starting from a solution of Na{sub 2}Pt(OH){sub 6} with 30 ppm of Pt, in refined samples and without polishing, they were characterized by scanning electron microscopy, energy dispersed spectroscopy, XPS and electrochemistry, by means of Tafel curves and cyclical polarization. On the zircaloy surface was found a ZrO{sub 2} layer that remains under the different study conditions. Under HWC conditions is the oxides formation, possibly complex oxides of zirconium, iron and tin. After the platinum deposit these oxides decrease forming the sub-oxides: Zr{sub 2}O, Zr O, Zr{sub 2}O{sub 3}. The Tafel curves indicates the reduction of the oxygen of the sample with platinum and the cyclical polarization curves show that the reactions that happen on the zircaloy electrodes are not dur to located corrosion. (Author)

  17. Corrosion performance of optimised and advanced fuel rod cladding in PWRs at high burnups

    International Nuclear Information System (INIS)

    Jourdain, P.; Hallstadius, L.; Pati, S.R.; Smith, G.P.; Garde, A.M.

    1997-01-01

    The corrosion behaviour both in-pile and out-of-pile for a number of cladding alloys developed by ABB to meet the current and future needs for fuel rod cladding with improved corrosion resistance is presented. The cladding materials include: 1) Zircaloy-4 (OPTIN) with optimised composition and processing and Zircaloy-2 optimised for Pressurised Water Reactors (PWR), (Zircaloy-2P), and 2) several alternative zirconium-based alloys with compositions outside the composition range for Zircaloys. The data presented originate from fuel rods irradiated in six PWRs to burnups up to about 66 MWd/kgU and from tests conducted in 360 o water autoclave. Also included are in-pile fuel rod growth measurements on some of the alloys. (UK)

  18. Extrusion and drawing of zircaloy 2. Production of pressure tubes for EL-4

    International Nuclear Information System (INIS)

    Thevenet, J.

    1964-01-01

    The authors give briefly the physical mechanical and chemical properties of zircaloy 2, as far as the transformation of this alloy is concerned. Extrusion: After a few general remarks concerning the extrusion and co-extrusion, including a comparison of the deformation resistance of canning metals and of zircaloy 2, the following points are considered: - the difficulties occurring because of the use of this alloy: - atmosphere protection - adjustment on to the machine tools - low thermal conductivity - economy of the metal (price) - the factors affecting the quality of the extruded products extrusion under a copper can and under lubricant glass - fine grain structure - temperature homogeneity - working temperature The transformation cycle - '550 kg ingot - preliminary shape 'for drawing of EL-4 tubes (112 x 120 L 12 m)' - is described in detail (extrusion or forging of the φ = 340 ingot into φ = 220 billets, cutting into lengths and hot drilling at φ = 125, fixing into a copper can and rough extrusion). Drawing: The main difficulties are due to seizing of the tools and to the necessity of protecting the alloy from the atmosphere during annealings. A brief description is given of drawing out on a short mandrel, on a long mandrel, of laminating on a reducing machine and of the carrying out of an annealing, as well as of the production of EL-4 tubes (φ =107 x 113 L 430 m) by drawing out shapes having a size of 112 x 120 on long mandrels. Conclusion: It is possible by extrusion and drawing to produce zircaloy 2 tubes similar to those which may be obtained normally using stainless steel. (authors) [fr

  19. Cyclic softening in annealed Zircaloy-2: Role of edge dislocation dipoles and vacancies

    Science.gov (United States)

    Sudhakar Rao, G.; Singh, S. R.; Krsjak, Vladimir; Singh, Vakil

    2018-04-01

    The mechanism of cyclic softening in annealed Zircaloy-2 at low strain amplitudes under strain controlled fatigue at room temperature is rationalized. The unusual softening due to continuous decrease in the phenomenological friction stress is found to be associated with decrease in the resistance against movement of dislocations because of the formation and easy glide of pure edge dislocation dipoles and consequent decrease in friction stress from reduction in the shear modulus. Positron annihilation spectroscopy data strongly support the increase in edge dislocation density containing jogs, from increased positron trapping and increase in annihilation lifetime.

  20. Metallographic Study of the Isothermal Transformation of Beta Phase in Zircaloy-2

    Energy Technology Data Exchange (ETDEWEB)

    Oestberg, G

    1960-06-15

    Observations of the structure of commercial zircaloy-2 have been made in the microscope showing that the high temperature beta phase is transformed isothermally at lower temperatures into alpha plus secondary precipitate. The alpha occurs mainly as Widmanstaetten plates developed by a shear mechanism. The secondary precipitate is formed from the beta - alpha structure at the phase boundary between these phases. This precipitation of particles of secondary phase occurs on account of a eutectoid reaction, alpha also being formed. A time-temperature transformation diagram has been constructed from the observations.

  1. Fatigue limit of Zircaloy-2 under variable one-directional tension and temperature 300 deg C

    International Nuclear Information System (INIS)

    Spasic, Z.; Simic, G.

    1968-11-01

    A vacuum chamber wad designed and constructed. It was suitable for study of materials at higher temperatures in vacuum or controlled atmospheres. Zircaloy-2 fatigue at 300 deg C in argon atmosphere was measured. Character of strain is variable one directional (A=1) tension. Obtained results are presented in tables and in the form of Veler's curve. The obtained fatigue limit was σ - 15 kp/mm 2 . The Locati method was allied as well and fatigue limit value obtained was 15,75 kp/mm 2 . Error calculated in reference to the previous value obtained by classical methods was 5% [sr

  2. Process for reliewing stresses in a zircaloy 2 or zircaloy 4 strip

    International Nuclear Information System (INIS)

    Charquet, D.; Dombre, M.

    1986-01-01

    Fabrication process of a zircaloy 2 or zircaloy 4 strip with an oxygen content between 900 and 1600 ppm with the following mechanical properties: E0.2≥250MPa at 315 deg C, parallel and perpendicular A% ≥4 at 20 deg C. The strip is rolled and stabilized by heat treatment between 490 and 580 deg C for 1 to 10 minutes and partially recrystallized for 0.5 to 5 vol.%. It is used for spacers of nuclear fuels [fr

  3. Method to determine the thermal conductivity of uranium dioxide and the surface conductance at the cladding-core interface from internal reactions

    Energy Technology Data Exchange (ETDEWEB)

    Tsykanov, V A; Samsonov, B F; Spiridonov, Yu G; Fomin, N A

    1975-01-01

    A method is given for determining the temperature-dependent thermal conductivity of uranium dioxide and the contact conductance of the gas gap between the core and cladding of a fuel element. These quantities should be determined on various samples with different diameters. A method is described for determining the heat-production rate of a fuel element to within 1.5 to 2.5 percent. The method is based on using a calibrated electric heater and a sensor to measure the specific energy evolution from reactor gamma-radiation. The total errors in determining the thermal conductivity and the contact conductance do not exceed 4.5 and 8 percent, respectively.

  4. A regression approach for Zircaloy-2 in-reactor creep constitutive equations

    International Nuclear Information System (INIS)

    Yung Liu, Y.; Bement, A.L.

    1977-01-01

    In this paper the methodology of multiple regressions as applied to Zircaloy-2 in-reactor creep data analysis and construction of constitutive equation are illustrated. While the resulting constitutive equation can be used in creep analysis of in-reactor Zircaloy structural components, the methodology itself is entirely general and can be applied to any creep data analysis. The promising aspects of multiple regression creep data analysis are briefly outlined as follows: (1) When there are more than one variable involved, there is no need to make the assumption that each variable affects the response independently. No separate normalizations are required either and the estimation of parameters is obtained by solving many simultaneous equations. The number of simultaneous equations is equal to the number of data sets. (2) Regression statistics such as R 2 - and F-statistics provide measures of the significance of regression creep equation in correlating the overall data. The relative weights of each variable on the response can also be obtained. (3) Special regression techniques such as step-wise, ridge, and robust regressions and residual plots, etc., provide diagnostic tools for model selections. Multiple regression analysis performed on a set of carefully selected Zircaloy-2 in-reactor creep data leads to a model which provides excellent correlations for the data. (Auth.)

  5. Parametric studies of cutting zircaloy-2 sheets with a laser beam

    International Nuclear Information System (INIS)

    Ghosh, S.; Badgujar, B.P.; Goswami, G.L.

    1996-01-01

    The highly reactive and pyrophoric nature of zirconium alloys limits the use of conventional thermal sources (e.g., plasma arc cutting, oxygen flame cutting, etc.) for the cutting and drilling of these alloys. In this context, a highly coherent laser beam provides a good alternative for the cutting and drilling. In the present paper, laser beam cutting of zircaloy-2 sheets of 1.1 mm and 0.74 mm thickness is performed using a 300 W average power pulsed Nd:YAG laser. Pulse energy, pulse repetition rate, nozzle gap, gas pressure and cutting speed were varied to give different laser cutting conditions. Metallographic study of the cut surfaces showed the presence of transformed beta phase in the heat affected zone (HAZ) near the cut surface. The microhardness value across the cut surface was also measured. It showed a gradual increase in microhardness from the base metal (160 VHN) towards the HAZ having a maximum value of 365 VHN. The results of parametric studies of the cutting indicated that, with proper selection of process parameters, very narrow cuts can be easily made in zircaloy-2 using a pulsed Nd:YAG laser with a saving in material and at a much faster rate than alternative processes such as plasma arc cutting and oxygen flame cutting

  6. Operating envelope to minimize probability of fractures in Zircaloy-2 pressure tubes

    International Nuclear Information System (INIS)

    Azer, N.; Wong, H.

    1994-01-01

    The failure mode of primary concern with Candu pressure tubes is fast fracture of a through-wall axial crack, resulting from delayed hydride crack growth. The application of operating envelopes is demonstrated to minimize the probability of fracture in Zircaloy-2 pressure tubes based on Zr-2.5%Nb pressure tube experience. The technical basis for the development of the operating envelopes is also summarized. The operating envelope represents an area on the pressure versus temperature diagram within which the reactor may be operated without undue concern for pressure tube fracture. The envelopes presented address both normal operating conditions and the condition where a pressure tube leak has been detected. The examples in this paper are prepared to illustrate the methodology, and are not intended to be directly applicable to the operation of any specific reactor. The application of operating envelopes to minimized the probability of fracture in 80 mm diameter Zircaloy-2 pressure tubes has been discussed. Both normal operating and leaking pressure tube conditions have been considered. 3 refs., 4 figs

  7. Electrochemical corrosion of Zircaloy-2 under PWR water chemistry but at room temperature

    International Nuclear Information System (INIS)

    Waheed, Abdel-Aziz Fahmy; Kandil, Abdel-Hakim Taha; Hamed, Hani M.

    2016-01-01

    Highlights: • There is no simple relation between the corrosion rate and LiOH concentration. • At low concentration, 100 ppm Li, an increase of the rate is due to the pH impact. • LiOH in concentrated solution led to accelerated corrosion by pH effect and porosity. • Boron abates the lithium effect by pH neutralizing and participation in the corrosion. - Abstract: Electrochemical corrosion of Zircaloy-2 was tested at room temperature in lithium hydroxide (LiOH) concentrations that ranged from 2.2 to 7000 ppm and boric acid (H 3 BO 3 ) concentrations that ranged from 50 to 4000 ppm. Following the corrosion experiments, the oxide films of specimens were examined by SEM to examine the oxide existence. LiOH concentrations as high as 1 M (7000-ppm lithium) can lead to significantly increased electrochemical corrosion rate. It is suggested that the accelerated corrosion in concentrated solution is caused by the synergetic effect of LiOH, pH and porosity generation. In solutions containing 100 ppm of lithium, the presence of boron had an ameliorating effect on the corrosion rates of Zircaloy-2. Similar to acceleration of corrosion by lithium, the inhibition by boron is due to a combined effect of pH neutralizing and its participation in the corrosion process.

  8. Cumulative damage fatigue tests on nuclear reactor Zircaloy-2 fuel tubes at room temperature and 3000C

    International Nuclear Information System (INIS)

    Pandarinathan, P.R.; Vasudevan, P.

    1980-01-01

    Cumulative damage fatigue tests were conducted on the Zircaloy-2 fuel tubes at room temperature and 300 0 C on the modified Moore type, four-point-loaded, deflection-controlled, rotating bending fatigue testing machine. The cumulative cycle ratio at fracture for the Zircaloy-2 fuel tubes was found to depend on the sequence of loading, stress history, number of cycles of application of the pre-stress and the test temperature. A Hi-Lo type fatigue loading was found to be very much damaging at room temperature and this feature was not observed in the tests at 300 0 C. Results indicate significant differences in damage interaction and damage propagation under cumulative damage tests at room temperature and at 300 0 C. Block-loading fatigue tests are suggested as the best method to determine the life-time of Zircaloy-2 fuel tubes under random fatigue loading during their service in the reactor. (orig.)

  9. Structural cladding /clad structures

    DEFF Research Database (Denmark)

    Beim, Anne

    2012-01-01

    Structural Cladding /Clad Structures: Studies in Tectonic Building Practice A. Beim CINARK – Centre for Industrialized Architecture, Institute of Architectural Technology, The Royal Danish Academy of Fine Arts School of Architecture, Copenhagen, Denmark ABSTRACT: With point of departure in the pr......Structural Cladding /Clad Structures: Studies in Tectonic Building Practice A. Beim CINARK – Centre for Industrialized Architecture, Institute of Architectural Technology, The Royal Danish Academy of Fine Arts School of Architecture, Copenhagen, Denmark ABSTRACT: With point of departure...... to analyze, compare, and discuss how these various construction solutions point out strategies for development based on fundamentally different mindsets. The research questions address the following issues: How to learn from traditional construction principles: When do we see limitations of tectonic maneuver......, to ask for more restrictive building codes. As an example, in Denmark there are series of increasing demands in the current building legislations that are focused at enhancing the energy performance of buildings, which consequently foster rigid insulation standards and ask for improvement of air...

  10. Critical stability conditions of the fuel element cladding; Kriticni uslovi stabilnosti kosuljice G.E

    Energy Technology Data Exchange (ETDEWEB)

    Pavlovic, M; Savic, D [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1968-12-15

    The role of the fuel element cladding being the first safety barrier, is to prevent contamination by the fission products. Construction of the fuel element cladding depends on the reactor type, coolant type, fuel type, technology of material fabrication, influence of the material on the neutron economy, thermal conditions, etc. That is why an optimum solution has to be found. This paper deals with mechanical properties of ceramic natural UO{sub 2} sintered fuel pellets in the zircaloy-2 cladding. This type of fuel is used in heavy water reactors.

  11. The effect of neutron irradiation on the mechanical properties of welded zircaloy-2

    Energy Technology Data Exchange (ETDEWEB)

    Evans, D G

    1962-07-15

    Zircaloy-2 tensile specimens, subsize impact bars and representative spigot welds were subjected to three NRX cycles in the X-5 loop. Average loop temperature was 260{sup o}C over the three cycles. One group of tensile specimens was heat-treated in vacuum at 900{sup o}C for 40 minutes, another group contained welded areas in the centre of the gauge length and a third group was hydrided after welding. Notches of the impact specimens were located in the fusion zone of the weld, Spigot welds were made on autoclaved and unautoclaved simulated production assemblies. The transition temperature of Zircaloy-2 increased appreciably upon welding. This was accompanied by a decrease in absorbed energy values for all temperatures between 0{sup o} and 300{sup o}C. Neutron irradiation had no effect on the impact properties of welded. Zircaloy-2. Welding decreased the uniform and total elongation at room temperature and at 260{sup o}C, and increased the 260{sup o}C PL, YS and UTS. Hydriding to a nominal 100 ppm hydrogen had no effect on the unirradiated tensile properties at either test temperature. The heat treatment decreased the strength properties but did not affect the ductility. Neutron irradiation increased the YS of the welded and hydrided material by 20% and the heat treated YS by 40%. Irradiation also increased the 260{sup o}C strength properties of the as-welded material. It was found that the unautoclaved spigot welds had a generally higher tensile strength than the autoclaved and welded specimens. For specimens welded in either condition, the outer welds of the 19-element bundle had a lower average breaking load than the inner welds. Neutron irradiation had no effect on the tensile strength of these welds. It was also demonstrated that a cup-and-cone type of fracture could be produced in a bend test. These fractures were similar to those observed in irradiated fuel bundles which had been damaged during transfer operations. A large amount of scatter rendered some

  12. Effect of dynamic strain aging on cyclic stress response and deformation behavior of Zircaloy-2

    International Nuclear Information System (INIS)

    Sudhakar Rao, G.; Verma, Preeti; Mahobia, G.S.; Santhi Srinivasa, N.C.; Singh, Vakil; Chakravartty, J.K.; Nudurupatic, Saibaba

    2016-01-01

    The effect of strain rate and temperature was studied on cyclic stress response and deformation behavior of annealed Zircaloy-2. Dynamic strain aging was exhibited under some test conditions. The cyclic stress response was found to be dependent on temperature and strain rate. At 300 °C, with decrease in strain rate, there was decrease in the rate as well as the degree of cyclic hardening. However, at 400°C, there was opposite trend and with decrease in strain rate both the rate as well as the degree of hardening increased. The deformation substructure showed dislocation bands, dislocation vein structure, PSB wall structure at both the temperatures. Irrespective of the temperature, there was dislocation loop structure, known as corduroy structure, at both the test temperatures. Based on the dislocation structure, the initial linear hardening is attributed to development of veins and PSB wall structure and the secondary hardening to the Corduroy structure. (author)

  13. Irradiation effects on Fe distributions in zircaloy-2 and Zr-2.5Nb

    International Nuclear Information System (INIS)

    Zou, H.; Hood, G.M.; Roy, J.A.

    1995-03-01

    Irradiation of large-grained Zr-2.5Nb (ZN) and Zircaloy-2 (Zy) with 1.5 MeV Ar ions to a fluence of ∼ 10 20 /m 2 (≡ 10 dpa) at 50, 300 and 420 deg C leads to enhanced α-phase Fe levels of 250-1500 ppma, compared to equivalent non-irradiated state values of ∼ 70 ppma. In ZN the β-phase Fe levels fell from about 6000 to 3500 ppma: this result accords, qualitatively, with the loss of Fe from the β-phase following in-service neutron irradiation. Measurements on Zy showed that the Fe concentrations were higher near the specimen surfaces. Limited data for Ni distributions in Zy show similar (to Fe) behaviour. (author). 18 refs., 2 tabs

  14. Role of internal stresses in the transient of irradiation growth of zircaloy-2

    International Nuclear Information System (INIS)

    Tome, C.N.; Christodoulou, N.; Turner, P.A.; Miller, M.A.; Woo, C.H.; Root, J.; Holden, T.M.

    1995-07-01

    A 'self-consistent' polycrystalline model is used to simulate irradiation growth of Zircaloy-2 samples irradiated at about 330 K. The predictions of the model are compared with experimental measurements obtained from specimens irradiated in the Advanced Test Reactor (ATR) at Idaho Falls. Three types of material are studied here: annealed, cold worked in tension and cold worked by rolling. In general, the growth rate attains a steady-state value after it goes through a transient that depends on the initial state of the material. The transient growth behaviour is explained in terms of the evolution of intergranular residual stresses that are present in the sample, and in terms of the dislocation structure. From this study, information regarding irradiation creep and growth mechanisms occurring at the single crystal level is obtained. (author). 28 refs., 1 tab., 4 figs

  15. Hot hardness studies on zircaloy 2 pressure tube along three orientations

    International Nuclear Information System (INIS)

    Kutty, T.R.G.; Ravi, K.; Jarvis, T.; Sengupta, A.K.; Majumdar, S.; Tewari, R.; Shrivastava, D.; Dey, G.K.

    2002-01-01

    Zirconium based alloys are the natural choice for both the fuel element cans and in-core structural components in water cooled nuclear reactors. In this paper, the hot hardness behaviour of zircaloy 2 pressure tubes has been examined from room temperature to 400 degC using a hot hardness tester. For the purpose of comparison, the hardness of the as cast and room temperature rolled specimens has also been carried out. For this, the samples were cut along three orientations and hardness was measured in each of these directions using Vickers diamond pyramid indenter. The variation in hardness of the pressure tube samples show that the hardness was highest along circumferential direction and least along the axial direction. The room temperature rolled samples showed highest hardness along the rolling planes. These variations in hardness could be explained in terms of development of texture during working on the material. (author)

  16. Stochastic model of texture dependence of iodine SCC susceptibility of a zircaloy-2 alloy

    International Nuclear Information System (INIS)

    Hirao, Keiichi; Yamane, Toshimi; Nakajima, Shinichi; Node, Shunsaku; Fujisawa, Takashi; Minamino, Yoritoshi

    1991-01-01

    Effects of textures on statistical parameters of tensile elongations in stress corrosion cracking (SCC) of zircaloy-2 using a slow strain rate test (SSRT) method have been investigated by Weibull distribution method based on stochastic process theory. The SCC is analyzed by assuming a probabilistic state transition model. Tensile directions of test pieces were prepared parallel, 45deg and perpendicular to rolling direction of the sheet. The test pieces in evacuated silica tubes were annealed at 1073K for 7.2x10 3 s, and then quenched into ice water. The annealed pieces with tilt angle α between tensile direction and a basal plane {0001} were 0, 18 and 25deg respectively. The tensile elongations of zircaloy-2 in SCC using the SSRT method are found to obey the single Weibull distribution with location parameters, and the SCC phenomena can be described by the Weibull distribution based on the stochastic process. The values of scale parameter η decrease with the tilt angle α, and the SCC susceptibility can be indicated by the values of scale parameter η. The texture dependence of the values of shape parameters m shows the changes of corrosion process in iodine solution and deformation system in air which are observed in the SSRT. The mechanism of decrement in the SCC susceptibility changes with the tilt angle α. The SCC under SSRT method is found to obey the model of probabilistic state transition. The constant load SCC process which obey the model of probabilistic state transition, is found to be effective for estimation of accelerated SCC condition. (author)

  17. The corrosion of zircaloy 2 in anaerobic synthetic cement pore solution

    International Nuclear Information System (INIS)

    Hansson, C.M.

    1984-12-01

    Measurements have been made of the corrosion rates of Zircaloy 2 tubes in anaerobic synthetic cement pore solution of pH 12.0-13.8. The samples were tested in the as-received condition by the polarization resistance technique using a Tafal constant of 52 mV/decade and, for all pH values, corrosion rates of 3.10 -5 A/m 2 (0.03 μm/yr) were determined. These corrosion currents are at the lower limit of the experimental detection range of the technique used. Some samples were then held at a low electrochemical potential, namely -1850 mV SCE, for several days but this treatment had only a minor effect on the behaviour of the Zircaloy: the value of corrosion rate was increased by a factor of 3 and the free potential was temporarily lowered but drifted towards more positive values after the applied potential was removed. Attempts were made to remove the passive film from the surface of the samples by electrochemical reduction. For practical, experimental reasons, this was not successful and, instead, the effect of removing the film by scratching the surface was investigated. At both the free potential and at applied cathodic potentials, an anodic current was detected immediately and the surface was scratched but, in all cases, the scratched area repassivated within a few seconds and the anodic corrosion current fell accordingly. Thus, it may be concluded that active corrosion of Zircaloy 2 in anaerobic concrete will not occur and, by comparison with measurements on steel, it is likely that the passive corrosion rates will be even lower in concrete than those measured in the synthetic pore solution. (Author)

  18. Uranium

    International Nuclear Information System (INIS)

    Hamdoun, N.A.

    2007-01-01

    The article includes a historical preface about uranium, discovery of portability of sequential fission of uranium, uranium existence, basic raw materials, secondary raw materials, uranium's physical and chemical properties, uranium extraction, nuclear fuel cycle, logistics and estimation of the amount of uranium reserves, producing countries of concentrated uranium oxides and percentage of the world's total production, civilian and military uses of uranium. The use of depleted uranium in the Gulf War, the Balkans and Iraq has caused political and environmental effects which are complex, raising problems and questions about the effects that nuclear compounds left on human health and environment.

  19. Influence of Zircaloy cladding composition on hydride formation during aqueous hydrogen charging

    Energy Technology Data Exchange (ETDEWEB)

    Rajasekhara, S. [Intel Corporation, 2501 NW 229th Av., Hillsboro, OR 97124 (United States); Kotula, P.G.; Enos, D.G.; Doyle, B.L. [Sandia National Laboratories, Albuquerque, NM, 87185 (United States); Clark, B.G., E-mail: blyclar@sandia.gov [Sandia National Laboratories, Albuquerque, NM, 87185 (United States)

    2017-06-15

    Although hydrogen uptake in Zirconium (Zr) based claddings has been a topic of many studies, hydrogen uptake as a function of alloy composition has received little attention. In this work, commercial Zr-based cladding alloys (Zircaloy-2, Zircaloy-4 and ZIRLO™), differing in composition but with similar initial textures, grain sizes, and surface roughness, were aqueously charged with hydrogen for 100, 300, and 1000 s at nominally 90 °C to produce hydride layers of varying thicknesses. Transmission electron microscope characterization following aqueous charging showed hydride phase and orientation relationship were identical in all three alloys. However, elastic recoil detection measurements confirmed that surface hydride layers in Zircaloy-2 and Zircaloy-4 were an order of magnitude thicker relative to ZIRLO™. - Highlights: •Aqueous charging was performed to produce a layer of zirconium hydride for three different Zr-alloy claddings. •Hydride thicknesses were analyzed by elastic recoil detection and transmission electron microscopy. •Zircaloy-2 and Zircaloy-4 formed thicker hydride layers than ZIRLO™ for the same charging durations.

  20. Corrosion of electron-irradiated Zr-2.5Nb and Zircaloy-2

    Energy Technology Data Exchange (ETDEWEB)

    Woo, O.-T.; McDougall, G.M.; Hutcheon, R.M.; Urbanic, V.F.; Griffiths, M.; Coleman, C.E

    2000-07-01

    We used 10-MeV electrons to rapidly produce radiation damage in zirconium alloys, investigated whether electrons produced the same microstructural changes as neutrons, then performed post-irradiation corrosion tests to determine whether electron-irradiated materials displayed similar corrosion behavior to neutron-irradiated materials. Two irradiations were completed using 10-MeV electrons with the beam normal to thin disks of material of 4 diameter slightly larger than the beam. The beam distribution. and disk cooling were designed to produce radial temperature and dose distributions having maxima at the disk center. A high-temperature irradiation was performed on annealed Zr-2.5Nb disks, achieving a central dose of 1.3 dpa and at a central temperature of {approx}450 deg C. After irradiation, the samples contained needle-like {beta}-Nb precipitates in the {alpha}-Zr matrix similar to those produced by neutrons. A low-temperature irradiation was performed on half-moon disks of Zr-2.5Nb and Zircaloy-2 pressure tube materials at 310 deg C central temperature and 1.3-dpa central dose. Dislocation loops were observed, again similar to those produced in neutron-irradiated materials. Some of the high-temperature electron-irradiated disks were exposed to 300 deg C moist air (saturated with D{sub 2}O), and in separate tests, high- and low-temperature irradiated disks were corroded in 300 deg C D{sub 2}0 (11.0 pD at room temperature) in an autoclave. Measurements of oxide thickness by Fourier Transform Infrared Reflectance (FTIR) spectroscopy showed that electron irradiation reduced the corrosion rate of Zr-2.5Nb compared with that of unirradiated material, as observed for neutron irradiation. For exposures to moist air and to D{sub 2}O, the theoretical deuterium uptakes for the electron-irradiated materials were, respectively, about 4 times and 1.5 to 2 times those for the unirradiated materials. This is also in good agreement with results for neutron-irradiated pressure

  1. Corrosion of electron-irradiated Zr-2.5Nb and Zircaloy-2

    International Nuclear Information System (INIS)

    Woo, O.-T.; McDougall, G.M.; Hutcheon, R.M.; Urbanic, V.F.; Griffiths, M.; Coleman, C.E.

    2000-01-01

    We used 10-MeV electrons to rapidly produce radiation damage in zirconium alloys, investigated whether electrons produced the same microstructural changes as neutrons, then performed post-irradiation corrosion tests to determine whether electron-irradiated materials displayed similar corrosion behavior to neutron-irradiated materials. Two irradiations were completed using 10-MeV electrons with the beam normal to thin disks of material of 4 diameter slightly larger than the beam. The beam distribution. and disk cooling were designed to produce radial temperature and dose distributions having maxima at the disk center. A high-temperature irradiation was performed on annealed Zr-2.5Nb disks, achieving a central dose of 1.3 dpa and at a central temperature of ∼450 deg C. After irradiation, the samples contained needle-like β-Nb precipitates in the α-Zr matrix similar to those produced by neutrons. A low-temperature irradiation was performed on half-moon disks of Zr-2.5Nb and Zircaloy-2 pressure tube materials at 310 deg C central temperature and 1.3-dpa central dose. Dislocation loops were observed, again similar to those produced in neutron-irradiated materials. Some of the high-temperature electron-irradiated disks were exposed to 300 deg C moist air (saturated with D 2 O), and in separate tests, high- and low-temperature irradiated disks were corroded in 300 deg C D 2 0 (11.0 pD at room temperature) in an autoclave. Measurements of oxide thickness by Fourier Transform Infrared Reflectance (FTIR) spectroscopy showed that electron irradiation reduced the corrosion rate of Zr-2.5Nb compared with that of unirradiated material, as observed for neutron irradiation. For exposures to moist air and to D 2 O, the theoretical deuterium uptakes for the electron-irradiated materials were, respectively, about 4 times and 1.5 to 2 times those for the unirradiated materials. This is also in good agreement with results for neutron-irradiated pressure tube materials. Thus, 10-Me

  2. Creep and stress rupture behaviour of zircaloy-2 and Zr-2.5% Nb alloy tubes at 573 K

    International Nuclear Information System (INIS)

    Laha, K.; Bhanu Sankara Rao, K.; Chandravathi, K.S.; Mannan, S.L.

    1992-01-01

    Zirconium alloys are extensively used for coolant tubes of pressurised heavy water reactors. The choice of these materials is based on their good corrosion resistance in water, low capture cross section for thermal neutrons and good mechanical properties. In this paper the results of an investigation performed on the creep and rupture behaviour of indigenously produced zircaloy-2 and Zr-2.5% Nb alloy are presented. Samples for creep testing were cut longitudinally from finished pressure tubes. Creep rupture tests were carried out in air under constant load conditions at 300 C employing five stress levels in the range 300-360 MPa. Zr-2.5% Nb alloy displayed higher rupture lives at all stress levels compared to zircaloy-2. Steady state creep rate of Zr-2.5%Nb was lower than that zircaloy-2 at identical stress levels. In the stress range of the experiments, the dependence of the steady state creep rate (ε s ) on applied stress (σ) for both the alloys could be represented by a power law, ε s =A σ n The stress sensitivity (n) for Zr-2.5% Nb was lower than that of zircaloy-2. For both the alloys the time to creep rupture t r was found related to the steady state creep rate through the modified Monkman-Grant relation (ε s ) α . t r = constant. Similar value of α was obtained for both the materials. Zr-2.5%Nb exhibited higher ductility (% elongation to rupture) compared to zircaloy-2 at stress levels ≥ 320 MPa. At lower stresses significant difference in ductility was not noticed. Percentage reduction in area was lower in Zr-2.5%Nb at all stress levels indicating better resistance for necking. The time for onset of tertiary was longer for Zr-2.5% Nb alloy. The proportion of life spent by Zr-2.5% Nb in steady state creep regime was higher compared to that of zircaloy-2. Metallographic investigations on longitudinal sections in both the alloys showed large number of intragranular pores close to the fracture surface. A few number of cracks which are characteristic of

  3. Uranium

    International Nuclear Information System (INIS)

    Cuney, M.; Pagel, M.; Leroy, J.

    1992-01-01

    First, this book presents the physico-chemical properties of Uranium and the consequences which can be deduced from the study of numerous geological process. The authors describe natural distribution of Uranium at different scales and on different supports, and main Uranium minerals. A great place in the book is assigned to description and classification of uranium deposits. The book gives also notions on prospection and exploitation of uranium deposits. Historical aspects of Uranium economical development (Uranium resources, production, supply and demand, operating costs) are given in the last chapter. 7 refs., 17 figs

  4. Effect of impurity elements Al, Mn, and N2 on the corrosion resistance of zircaloy-2 in high temperature water and steam

    International Nuclear Information System (INIS)

    Gadiyar, H.S.

    1978-01-01

    Although the impurity limits are specified in standard zircaloy-2, it is possible that during its manufacture some of the impurities may exceed by a few ppm than the normally set values. It is necessary to understand the corrosion behaviour of such zircaloy-2 which contain a small amount of excessive impurities. This report summarizes some such data of the impurities aluminium, manganese and nitrogen. It is seen that the common impurities which can affect the corrosion of zircaloy-2 significantly are Al and N 2 and to a lesser extent Mn. (author)

  5. Influence of texture on fracture toughness of zircaloy cladding

    Energy Technology Data Exchange (ETDEWEB)

    Grigoriev, V. [Studsvik Material AB, Nykoeping (Sweden); Andersson, Stefan [Royal Inst. of Tech., Stockholm (Sweden)

    1997-06-01

    The correlation between texture and fracture toughness of Zircaloy 2 cladding has been investigated in connection with axial cracks in fuel rods. The texture of the cladding determines the anisotropy of plasticity of the cladding which, in turn, should influence the strain conditions at the crack-tip. Plastic strains in the cladding under uniaxial tension were characterised by means of the anisotropy constants F, G and H calculated according to Hill`s theory. Test temperatures between 20 and 300 deg C do not influence the F, G and H values. Any significant effect of hydrogen (about 500 wtppm) on the anisotropy constants F, G and H has not been revealed at a test temperature of 300 deg C. The results, obtained for stress-relieved and recrystallized cladding with different texture, show an obvious influence of texture on the fracture toughness of Zircaloy cladding. A higher fracture toughness has been found for cladding with more radial texture. With a 2 page summary in Swedish. 32 refs, 18 figs.

  6. Inverse strain rate effect on cyclic stress response in annealed Zircaloy-2

    Energy Technology Data Exchange (ETDEWEB)

    Sudhakar Rao, G.; Verma, Preeti [Center of Advanced Study, Department of Metallurgical Engineering, Indian Institute of Technology (Banaras Hindu University), Varanasi 221005 (India); Chakravartty, J.K. [Mechanical Metallurgy Group, Bhabha Atomic Research Center, Trombay 400 085, Mumbai (India); Nudurupati, Saibaba [Nuclear Fuel Complex, Hyderabad 500 062 (India); Mahobia, G.S.; Santhi Srinivas, N.C. [Center of Advanced Study, Department of Metallurgical Engineering, Indian Institute of Technology (Banaras Hindu University), Varanasi 221005 (India); Singh, Vakil, E-mail: vsingh.met@itbhu.ac.in [Center of Advanced Study, Department of Metallurgical Engineering, Indian Institute of Technology (Banaras Hindu University), Varanasi 221005 (India)

    2015-02-15

    Low cycle fatigue behavior of annealed Zircaloy-2 was investigated at 300 and 400 °C at different strain amplitudes and strain rates of 10{sup −2}, 10{sup −3}, and 10{sup −4} s{sup −1}. Cyclic stress response showed initial hardening with decreasing rate of hardening, followed by linear cyclic hardening and finally secondary hardening with increasing rate of hardening for low strain amplitudes at both the temperatures. The rate as well the degree of linear hardening and secondary hardening decreased with decrease in strain rate at 300 °C, however, there was inverse effect of strain rate on cyclic stress response at 400 °C and cyclic stress was increased with decrease in strain rate. The fatigue life decreased with decrease in strain rate at both the temperatures. The occurrence of linear cyclic hardening, inverse effect of strain rate on cyclic stress response and deterioration in fatigue life with decrease in strain rate may be attributed to dynamic strain aging phenomena resulting from enhanced interaction of dislocations with solutes. Fracture surfaces revealed distinct striations, secondary cracking, and oxidation with decrease in strain rate. Deformation substructure showed parallel dislocation lines and dislocation band structure at 300 °C. Persistent slip band wall structure and development of fine Corduroy structure was observed at 400 °C.

  7. Factors affecting in-core dimensional stability of Zircaloy-2 calandria tubes

    International Nuclear Information System (INIS)

    Fidleris, V.; Causey, A.R.; Holt, R.A.

    1985-01-01

    In CANDU PHW reactors, the heavy water moderator is contained in a cylindrical vessel (calandria) which is penetrated by 380 horizontal fuel channel assemblies. The outer Zircaloy-2 tube of each assembly (the calandria tube) is rolled into the end shields to seal the calandria. The calandria tubes operate at ≅340 K with axial stresses that range from -10 to +40 MPa and experience fast neutron fluxes as large as 3 x 10 17 n m -2 s -1 , E > 1.0 MeV. In this environment tubes elongate and sag due to irradiation-induced creep and growth. Our understanding of these irradiation effects is based on creep, stress relaxation and irradiation growth experiments on calandria tube materials irradiated to neutron fluences of 7 x 10 25 n m -2 , E > 1.0 MeV. Both creep and growth strains decrease with the proportion of grains that have basal plane normals in the direction of testing. Cold work increases the creep rate but appears to introduce a negative component of growth in the working direction due to neutron induced stress relief that persists up to at least 7 x 10 25 n m -2 . Thermal stress relief restores the positive growth rate in the working direction. There is little effect of grain size in the range 10 TO 30 μm. This information can be used to select fabrication routes that will minimize dimensional changes of tubes during service

  8. Studies of the Effective Total and Resonance Absorption Cross Sections for Zircaloy 2 and Zirconium

    Energy Technology Data Exchange (ETDEWEB)

    Hellstrand, E; Lindahl, G; Lundgren, G

    1961-06-15

    Using pile oscillator technique, the total absorption cross section for zircaloy 2 plates has been determined in the neutron spectrum of the reactor R1. The plate thickness was varied in six steps from 0. 2 mm to 6. 4 mm. The thermal cross section for the alloy was calculated from cross section data and the known composition of the alloy. By subtracting this value from the measured cross sections and dividing by the factor {alpha}=2/{radical}({pi}) x r x {radical}(T/T{sub 0}) the effective resonance integrals were obtained. After subtraction of a constant amount for resonance contributions from hafnium, tin etc., effective resonance integrals for zirconium could be evaluated. An extrapolated value of 0.85 {+-} 0.15 b was obtained for the infinitely dilute integral (l/v part excluded). The ratio of the resonance integral at plate thicknesses 0.2 and 6.4 mm came out as 1.65 {+-} 0.25.

  9. Influence of deformation history on texture change and subsequent yield locus of zircaloy-2 tubing

    International Nuclear Information System (INIS)

    Nagai, Nobuyuki; Kakuma, Tsutomu; Miyamoto, Yoshiyuki

    1981-01-01

    Fully-annealed Zircaloy-2 tubing was strained by balanced axial stress σsub(z) and circumferential stress σsub(theta) (stress ratio: α = σsub(z)/σsub(theta)). Then, texture and subsequent yield loci of these prestrained materials were measured. Results of texture measurement after prestraining showed that (0002) poles tend to move toward the radial tube direction under α = 0, 0.5 and 1, but toward the circumferential tube direction under α = 2 and infinity. Specimens highly prestrained under α = 0 and 0.5 have extremely concentrated texture. Such texture changes can be explained by a deformation model in which type slip system was assumed as one of the deformation system. The yield strength of most prestrained materials is higher than that of starting material, however, the material prestrained under α = infinity shows lower yield strength than starting material under test condition of α = 0. It was observed that the texture change had an important influence on subsequent yield behavior. Typically, the material highly prestrained under α = 0.5, which had concentrated basal poles, gave the yield locus characterized by remarkable ''texture hardening''. (author)

  10. The examination of the ruptured Zircaloy-2 pressure tube from Pickering NGS Unit 2

    International Nuclear Information System (INIS)

    Cheadle, B.A.; Smith, A.D.; Baskin, C.C.

    1985-07-01

    On 1983 August 01 a Zircaloy-2 pressure tube in Pickering NGS Unit 2 ruptured. All the fuel channel components, the fuel bundles, pressure tube, end fittings, garter springs and calandria tubes were shipped to Chalk River Nuclear Laboratories for examination to determine the cause of the rupture. The examination showed that the rupture initiated at a series of hydride blisters on the outside surface of the pressure tube. The blisters formed because of the garter spring spacers between the pressure tube and calandria tube was about one metre out of position. This allowed the horizontal pressure tube to sag by creep and touch the cool calandria tube. The resulting thermal gradients in the pressure tube concentrated the hydrogen and deuterium at the cool zones and blisters of solid hydride formed. Cracks initiated at several of the blisters and linked together to form a partial through wall critical crack which initiated the final rupture. The video presentation shows how the examination of the fuel channel components was conducted in underwater bays and shielded cells and explains the sequence of events that caused the rupture

  11. Characterization of Zircaloy-2 and Zircaloy-4 by X-Ray fluorescence

    International Nuclear Information System (INIS)

    Sato, I.M.; Imakuma, K.; Salvador, V.L.R.

    1981-03-01

    The analytical characterization of zircaloy-2 and zircaloy-4 is intimataly connected with the determination of Sn, Fe, Cr, Ni, O, N, H, and Hf. An analytical method developed in this laboratory is discribed for the determination of metallic elements like Sn, Fe, Cr and Ni using the technique of X-ray fluorescence. The samples are prepared in the form of double-layer pellets using boric acid as a binding agent. The zircaloy-4 is dissolved in hydrofluoric acid and the metallic elements are converted to fluorides. The standard samples used for calibration are prepared from synthetic materials. The elements are determined by measuring the characteristic first order K α lines. A Zircaloy-4 sample analysed yielded the following values: Sn=1.30+-0.03%, Fe=0.18+-0.01%, Cr=0.088+-0.004% and Ni=14+-3 ppm. The reproducibility, precision, as well as the theoretical limit of detection of the method are discussed. The determination of the elements O, N and H present as occluded gas in the zircaloy is nearing completion. These analyses are being carried out by a Mass Spectrometric technique where an aliquot of the released gas is analysed. (Author) [pt

  12. The effect of plastic strain on the evolution of crystallographic texture in Zircaloy-2

    International Nuclear Information System (INIS)

    Ballinger, R.G.; Lucas, G.E.; Pelloux, R.M.

    1984-01-01

    The evolution of crystallographic texture during plastic deformation was investigated in Zircaloy-2 using X-ray and metallographic techniques. Inverse pole figures, the resolved fraction of basal poles, and the volume fraction of twinned material, were determined as a function of plastic strain for several strain paths and initial textures at 298 K and 623 K. Incremental transverse platic strain ratios (R) were measured as a function of plastic strain. Texture rotation occurs early in the deformation process, after as little as 1.5% plastic strain. For compressive plastic strains, the resolved fraction of basal poles increases in the direction parallel to the strain axis. For tensile plastic strains, the resolved fraction of basal poles decreases in the direction parallel to the strain axis. The rate of change of the resolved fraction of basal poles with plastic strain is a function of the initial resolved fraction of basal poles. The texture rotation can be explained by considering the operating of the principal tensile twinning systems, [10anti 12], . (orig.)

  13. High-temperature oxidation of Zircaloy-2 and Zircaloy-4 in steam

    International Nuclear Information System (INIS)

    Urbanic, V.F.; Heidrick, T.R.

    1978-01-01

    At temperatures above the (α + β)/β transformation temperature for zirconium alloys, steam reacts with β-Zr to form a superficial layer of zirconium oxide (ZrO 2 ) and an intermediate layer of oxygen-stabilized α-Zr. Reaction kinetics and the rate of growth of the combined (ZrO 2 + α-Zr) layer for Zircaloy-2 and Zircaloy-4 oxidation in steam were measured over the temperature range 1050-1850 o C. The reaction rates for both alloys were similar, obeyed parabolic kinetics and were not limited by gas phase diffusion. The parabolic rate constants were consistently less than those given by the Baker and Just correlation for zirconium oxidation in steam. A discontinuity was found in the temperature dependence of both the reaction rate and the rate of growth of the combined (ZrO 2 + α-Zr) layer. The discontinuity is attributed to a change in the oxide microstructure at the discontinuity temperature, an observation which is consistent with the zirconium-oxygen phase diagram. (author)

  14. Effect of plastic strain on the evolution of crystallographic texture in Zircaloy-2

    Energy Technology Data Exchange (ETDEWEB)

    Ballinger, R.G. (Massachusetts Inst. of Tech., Cambridge (USA)); Lucas, G.E. (California Univ., Santa Barbara (USA)); Pelloux, R.M. (Massachusetts Inst. of Tech., Cambridge (USA). Dept. of Materials Science and Engineering)

    1984-09-01

    The evolution of crystallographic texture during plastic deformation was investigated in Zircaloy-2 using X-ray and metallographic techniques. Inverse pole figures, the resolved fraction of basal poles, and the volume fraction of twinned material, were determined as a function of plastic strain for several strain paths and initial textures at 298 K and 623 K. Incremental transverse platic strain ratios (R) were measured as a function of plastic strain. Texture rotation occurs early in the deformation process, after as little as 1.5% plastic strain. For compressive plastic strains, the resolved fraction of basal poles increases in the direction parallel to the strain axis. For tensile plastic strains, the resolved fraction of basal poles decreases in the direction parallel to the strain axis. The rate of change of the resolved fraction of basal poles with plastic strain is a function of the initial resolved fraction of basal poles. The texture rotation can be explained by considering the operating of the principal tensile twinning systems, (10anti 12), .

  15. The effect of plastic strain on the evolution of crystallographic texture in Zircaloy-2

    Science.gov (United States)

    Ballinger, R. G.; Lucas, G. E.; Pelloux, R. M.

    1984-09-01

    The evolution of crystallographic texture during plastic deformation was investigated in Zircaloy-2 using X-ray and metallographic techniques. Inverse pole figures, the resolved fraction of basal poles, and the volume fraction of twinned material, were determined as a function of plastic strain for several strain paths and initial textures at 298 K and 623 K. Incremental transverse platic strain ratios ( R) were mesured as a function of plastic strain. Texture rotation occurs early in the deformation process, after as little as 1.5% plastic strain. For compressive plastic strains, the resolved fraction of basal poles increases in the direction parallel to the strain axis. For tensile plastic strains, the resolved fraction of basal poles decreases in the direction parallel to the strain axis. The rate of change of the resolved fraction of basal poles with plastic strain is a function of the initial resolved fraction of basal poles. The texture rotation can be explained by considering the operation of the principal tensile twinning systems, {101¯2}.

  16. Inverse strain rate effect on cyclic stress response in annealed Zircaloy-2

    International Nuclear Information System (INIS)

    Sudhakar Rao, G.; Verma, Preeti; Chakravartty, J.K.; Nudurupati, Saibaba; Mahobia, G.S.; Santhi Srinivas, N.C.; Singh, Vakil

    2015-01-01

    Low cycle fatigue behavior of annealed Zircaloy-2 was investigated at 300 and 400 °C at different strain amplitudes and strain rates of 10 −2 , 10 −3 , and 10 −4 s −1 . Cyclic stress response showed initial hardening with decreasing rate of hardening, followed by linear cyclic hardening and finally secondary hardening with increasing rate of hardening for low strain amplitudes at both the temperatures. The rate as well the degree of linear hardening and secondary hardening decreased with decrease in strain rate at 300 °C, however, there was inverse effect of strain rate on cyclic stress response at 400 °C and cyclic stress was increased with decrease in strain rate. The fatigue life decreased with decrease in strain rate at both the temperatures. The occurrence of linear cyclic hardening, inverse effect of strain rate on cyclic stress response and deterioration in fatigue life with decrease in strain rate may be attributed to dynamic strain aging phenomena resulting from enhanced interaction of dislocations with solutes. Fracture surfaces revealed distinct striations, secondary cracking, and oxidation with decrease in strain rate. Deformation substructure showed parallel dislocation lines and dislocation band structure at 300 °C. Persistent slip band wall structure and development of fine Corduroy structure was observed at 400 °C

  17. Assessment of hydrogen levels in Zircaloy-2 by non-destructive testing

    International Nuclear Information System (INIS)

    De, P.K.; John, J.T.; Banerjee, S.; Jayakumar, T.; Thavasimuthu, M.; Raj, B.

    1998-01-01

    A non-destructive assessment of Zircaloy-2 samples charged with hydrogen in the range of 50 to 1150 mg/kg has been made using ultrasonic and eddy current testing. It has been found that the ratio of the longitudinal to the shear wave velocity is a parameter which can be directly correlated with the hydrogen content up to a level of 100 to 200 mg/kg. This parameter together with the values of longitudinal and shear wave velocities can be utilized in a multi-parametric correlation approach for estimation of higher levels of the hydrogen content (up to 1150 mg/kg). The sensitivity at different ranges has been found to be acceptable. Ultrasonic attenuation measurements at higher frequencies and eddy current test parameter are also effective for estimation of hydrogen levels above 250 mg/kg in zirconium alloys. Microstructural characterization including TEM studies have been carried out for studying the influence of the type and the morphology of hydride precipitates on ultrasonic parameters. (orig.)

  18. Static strain aging of Zircaloy-2: the effect of dislocation dynamics on yielding behaviour

    International Nuclear Information System (INIS)

    Thorpe, W.R.; Smith, I.O.

    1981-01-01

    The static strain-aging response of Zircaloy-2 was determined in the temperature range 293-723 K. A modified Hahn yielding model was found to provide a satisfactory description of the magnitude and shape of the yield points after aging, thereby providing information about the mobile dislocation density and the dislocation generation rate. For example, the characteristic double peak in the temperature dependence of strain aging was simplified to a single broad minimum in the mobile dislocation density over the temperature interval 500-700 K. The shape of the yield point was also found to be temperature dependent; the yield drop became less sharp at test temperatures above 648 K. This was ascribed to the inhibition of dislocation multiplication by dynamic strain aging. A kinetic law was developed by applying Snoek ordering kinetics to the process of dislocation locking and the resultant change in mobile dislocation density was then used to predict the strain-aging response as a function of aging time. The stress dependence of strain aging at 573 K was investigated at aging stresses of between 0.07 and 0.975 of the flow stress sigmasub(f). The strain-aging response increased for aging at stresses between 0.07sigmassub(f) and 0.8sigmasub(f), whereafter it declined steeply to the limit of zero at the flow stress. (Auth.)

  19. Evolution of processing of GE fuel clad tubing for corrosion resistance in boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Williams, C.D. [GE Nuclear Energy, Wilmington, NC (United States); Adamson, R.B. [GE Nuclear Energy, Wilmington, NC (United States); Marlowe, M.O. [GE Nuclear Energy, Wilmington, NC (United States); Plaza-Meyer, E. [GE Nuclear Energy, Wilmington, NC (United States); Proebstle, R.A. [GE Nuclear Energy, Wilmington, NC (United States); White, D.W. [GE Nuclear Energy, Wilmington, NC (United States)

    1996-05-01

    The current modification of the primary GE in-process solution-quench heat treatment, an (alpha+beta) solution-quench carried out at a tube diameter requiring only two subsequent reduction and anneal cycles, is applicable to Zr barrier fuel clad tubing, to non-barrier fuel clad tubing, and to the TRICLAD tubing product. A combination of good in-reactor corrosion performance and degradation resistance is anticipated for these products, based on knowledge of metallurgical characteristics and supported by the demonstrated performance capability of the Zircaloy-2 materials used. (orig.)

  20. Uranium

    International Nuclear Information System (INIS)

    Anon.

    1984-01-01

    The article briefly discusses the Australian government policy and the attitude of political party factions towards the mining and exporting of the uranium resources in Australia. Australia has a third of the Western World's low-cost uranium resources

  1. Uranium

    International Nuclear Information System (INIS)

    Poty, B.; Cuney, M.; Bruneton, P.; Virlogeux, D.; Capus, G.

    2010-01-01

    With the worldwide revival of nuclear energy comes the question of uranium reserves. For more than 20 years, nuclear energy has been neglected and uranium prospecting has been practically abandoned. Therefore, present day production covers only 70% of needs and stocks are decreasing. Production is to double by 2030 which represents a huge industrial challenge. The FBR-type reactors technology, which allows to consume the whole uranium content of the fuel, is developing in several countries and will ensure the long-term development of nuclear fission. However, the implementation of these reactors (the generation 4) will be progressive during the second half of the 21. century. For this reason an active search for uranium ores will be necessary during the whole 21. century to ensure the fueling of light water reactors which are huge uranium consumers. This dossier covers all the aspects of natural uranium production: mineralogy, geochemistry, types of deposits, world distribution of deposits with a particular attention given to French deposits, the exploitation of which is abandoned today. Finally, exploitation, ore processing and the economical aspects are presented. Contents: 1 - the uranium element and its minerals: from uranium discovery to its industrial utilization, the main uranium minerals (minerals with tetravalent uranium, minerals with hexavalent uranium); 2 - uranium in the Earth's crust and its geochemical properties: distribution (in sedimentary rocks, in magmatic rocks, in metamorphic rocks, in soils and vegetation), geochemistry (uranium solubility and valence in magmas, uranium speciation in aqueous solution, solubility of the main uranium minerals in aqueous solution, uranium mobilization and precipitation); 3 - geology of the main types of uranium deposits: economical criteria for a deposit, structural diversity of deposits, classification, world distribution of deposits, distribution of deposits with time, superficial deposits, uranium

  2. Uranium

    International Nuclear Information System (INIS)

    Mackay, G.A.

    1978-01-01

    The author discusses the contribution made by various energy sources in the production of electricity. Estimates are made of the future nuclear contribution, the future demand for uranium and future sales of Australian uranium. Nuclear power growth in the United States, Japan and Western Europe is discussed. The present status of the six major Australian uranium deposits (Ranger, Jabiluka, Nabarlek, Koongarra, Yeelerrie and Beverley) is given. Australian legislation relevant to the uranium mining industry is also outlined

  3. Uranium

    International Nuclear Information System (INIS)

    1982-01-01

    The development, prospecting, research, processing and marketing of South Africa's uranium industry and the national policies surrounding this industry form the headlines of this work. The geology of South Africa's uranium occurences and their positions, the processes used in the extraction of South Africa's uranium and the utilisation of uranium for power production as represented by the Koeberg nuclear power station near Cape Town are included in this publication

  4. Uranium

    International Nuclear Information System (INIS)

    Stewart, E.D.J.

    1974-01-01

    A discussion is given of uranium as an energy source in The Australian economy. Figures and predictions are presented on the world supply-demand position and also figures are given on the added value that can be achieved by the processing of uranium. Conclusions are drawn about Australia's future policy with regard to uranium (R.L.)

  5. Uranium

    International Nuclear Information System (INIS)

    Toens, P.D.

    1981-03-01

    The geological setting of uranium resources in the world can be divided in two basic categories of resources and are defined as reasonably assured resources, estimated additional resources and speculative resources. Tables are given to illustrate these definitions. The increasing world production of uranium despite the cutback in the nuclear industry and the uranium requirements of the future concluded these lecture notes

  6. Modelling of oxidation and hydriding behaviour of Zircaloy-2 pressure tubes in PHWR

    International Nuclear Information System (INIS)

    Sah, D.N.; Sunil Kumar; Khan, K.B.

    2002-01-01

    A computer model named DOCTOR (Deuteriding of Coolant Tubes during Operation of Reactor) has been developed for predicting the axial profile of oxide thickness and hydrogen (Deuterium) concentration in PHWR pressure tubes. This model is applicable to single channel or full core analysis. The main source of hydrogen is considered to be oxidation of pressure tube on the i.d. surface by high temperature coolant water. Three stages of oxidation is considered namely, pre- transition, post transition and accelerated. Oxidation rate is considered to be dependent on channel power, axial power/flux distribution, coolant temperature and pre-existing oxide thickness at the location. The kinetics parameters for oxidation model are derived from the actual measurement of oxide thickness on a number of pressure tubes examined in PIE Division. The input data required for the model are: channel power, channel power factor, axial flux distribution, coolant inlet temperature, critical oxide thickness, hydrogen pick up fraction, initial hydrogen in the material and time of operation (efpy). The model calculates the oxide layer thickness on the inside surface of the pressure tube along the length. The amount of hydrogen picked up by the pressure tube is calculated from the oxide thickness using hydrogen pick up fraction determined from the PIE data. The pressure tube length is divided into a number of axial segments for calculation. The temperature and fast neutron flux assumed to be constant in a given segment. The axial temperature profile calculated from the axial power profile in the channel is used for calculating the oxidation rate at various locations in the pressure tube. The model has been validated with PIE data of hydrogen equivalent measurement on a number of irradiated Zircaloy-2 pressure tubes of various PHWRs. The performance of the model in predicting the axial profile of hydrogen in the pressure tubes has been found to be good. (author)

  7. Combined effects of radiation damage and hydrides on the ductility of Zircaloy-2

    International Nuclear Information System (INIS)

    Wisner, S.B.; Adamson, R.B.

    1998-01-01

    Interest remains high regarding the effects of zirconium hydride precipitates on the ductility of reactor Zircaloy components, particularly in irradiated material. Previous studies have reported that ductility reductions are much greater at room temperature compared to reactor component temperatures. It is often concluded that the effects of irradiation dominate the ductility reduction observed in test specimens, although there is no consensus as to whether hydriding effects are additive. Many of the tests reported in the literature are difficult to interpret due to variations in test specimen geometry and material history. In this paper, we present the results of an experimental program aimed at clearly describing the combined effects of irradiation and hydriding on ductility parameters under conditions of a realistic test specimen design and well characterized hydride content, distribution and orientation. Experiments were conducted at 295 and 605 K, respectively on Zircaloy-2 tubing segments containing 10-800 ppm hydrogen and neutron fluences between 0.9 x 10 25 nm -2 (E>1 MeV). Tests utilized the well proven localized ductility specimen which applies plane strain tension in the hoop direction of the tubing segment. In all cases, hydrides were also oriented in the hoop or circumferential direction and were uniformly distributed across the tubing wall. Results indicate that at 605 K, the ductility of irradiated material was almost independent of hydride content, retaining above 4% uniform elongation and 25% reduction in an area for the highest fluences and hydrogen contents. Even at 295 K, measurable ductility was retained for irradiated material with up to 600 ppm hydrogen. In the paper, results of fractographic analyses and strain rate are also discussed

  8. A regression approach for zircaloy-2 in-reactor creep constitutive equations

    International Nuclear Information System (INIS)

    Yung Liu, Y.; Bement, A.L.

    1977-01-01

    In this paper the methodology of multiple regressions as applied to zircaloy-2 in-reactor creep data analysis and construction of constitutive equation are illustrated. While the resulting constitutive equation can be used in creep analysis of in-reactor zircaloy structural components, the methodology itself is entirely general and can be applied to any creep data analysis. From data analysis and model development point of views, both the assumption of independence and prior committment to specific model forms are unacceptable. One would desire means which can not only estimate the required parameters directly from data but also provide basis for model selections, viz., one model against others. Basic understanding of the physics of deformation is important in choosing the forms of starting physical model equations, but the justifications must rely on their abilities in correlating the overall data. The promising aspects of multiple regression creep data analysis are briefly outlined as follows: (1) when there are more than one variable involved, there is no need to make the assumption that each variable affects the response independently. No separate normalizations are required either and the estimation of parameters is obtained by solving many simultaneous equations. The number of simultaneous equations is equal to the number of data sets, (2) regression statistics such as R 2 - and F-statistics provide measures of the significance of regression creep equation in correlating the overall data. The relative weights of each variable on the response can also be obtained. (3) Special regression techniques such as step-wise, ridge, and robust regressions and residual plots, etc., provide diagnostic tools for model selections

  9. Cladding embrittlement during postulated loss-of-coolant accidents.

    Energy Technology Data Exchange (ETDEWEB)

    Billone, M.; Yan, Y.; Burtseva, T.; Daum, R.; Nuclear Engineering Division

    2008-07-31

    The effect of fuel burnup on the embrittlement of various cladding alloys was examined with laboratory tests conducted under conditions relevant to loss-of-coolant accidents (LOCAs). The cladding materials tested were Zircaloy-4, Zircaloy-2, ZIRLO, M5, and E110. Tests were performed with specimens sectioned from as-fabricated cladding, from prehydrided (surrogate for high-burnup) cladding, and from high-burnup fuel rods which had been irradiated in commercial reactors. The tests were designed to determine for each cladding material the ductile-to-brittle transition as a function of steam oxidation temperature, weight gain due to oxidation, hydrogen content, pre-transient cladding thickness, and pre-transient corrosion-layer thickness. For short, defueled cladding specimens oxidized at 1000-1200 C, ring compression tests were performed to determine post-quench ductility at {le} 135 C. The effect of breakaway oxidation on embrittlement was also examined for short specimens oxidized at 800-1000 C. Among other findings, embrittlement was found to be sensitive to fabrication processes--especially surface finish--but insensitive to alloy constituents for these dilute zirconium alloys used as cladding materials. It was also demonstrated that burnup effects on embrittlement are largely due to hydrogen that is absorbed in the cladding during normal operation. Some tests were also performed with longer, fueled-and-pressurized cladding segments subjected to LOCA-relevant heating and cooling rates. Recommendations are given for types of tests that would identify LOCA conditions under which embrittlement would occur.

  10. Quality Selection of Zircaloy-2 Canning Tubes by Ultrasonic Testing on Small Defects; Controle de la Qualite des Gaines en Zircaloy-2: Detection de Petits Defauts par les Ultrasons; Achestvennyjotb ortrub chatykh obolochek iz tsirkalloya-2 putem vyyavleniya nebol'shikh defektov s pomoshch'yu ul'trazvuka; Control de Calidad de los Revestimientos de Zircaloy-2 por Localizacion Ultrasonica de Pequenos Defectos

    Energy Technology Data Exchange (ETDEWEB)

    Van Der Linde, A. [Reactor Centrum Nederland, Petten (Netherlands); Deraad, J. A. [Roentgen Technische Dienst N.V., Rotterdam (Netherlands)

    1965-09-15

    Zircaloy-2 canning tubes, 10.20 mm I.D. x 0.90 mm wall x 1500 mm length, destined for testing as fuel-rod cladding in a high temperature. 330 Degree-Sign C, in-pile pressurized water loop, were tested ultrasonically for defects to get an impression of the tubes' quality. The tested tubes were delivered by manufacturers in the United States of America, United Kingdom and Scandinavia. Our requirement that all delivered tubes should be free from defects with a length greater than 500- 1000 {mu}m and/or with a depth greater than 50-25 {mu}m was not completely accepted by the manufacturers. They could guarantee that defects longer than 1000 {mu}m and/or with a depth greater than 50 {mu}m should be absent. Because only two of the 93 tested tubes had defects with a depth greater than 50 {mu}m it was decided to apply a more severe test by which defects with a depth in the range 10-50 {mu}m could be detected. To detect and record such small defects, longitudinal as well as transverse, a semi-automatic ultrasonic pulse equipment was used in combination with gating systems and a multiple-channel recorder. The adjustment of the scanning system was such that inner and outer defects of the same size were indicated with equal amplitudes. Calibration of the equipment was made on artificial defects. Longitudinal defects were detected with a separated transmitter-receiver system using a focused beam. Transverse defects were scanned by a single probe acting as transceiver. To obtain the sensitivity required the tests were carried out in immersion at a frequency of 4 MHz whereby the tubes were rotated with 120 rpm. A description is given of the mechanical device, the general set-up and the difficulties encountered. The result was that from the 93 tested tubes 21 had defects in the transverse direction with a depth between 10 and 50 {mu}m. Thus a relative qualification of the tubes was obtained. (author) [French] Des gaines en Zircaloy-2, ayant un diametre interieur de 10, 20 mm, une

  11. Study of necking stability in tension test of zircaloy-2, on range from 170 0 C to 620 0 C

    International Nuclear Information System (INIS)

    Okuda, M.Y.

    1975-01-01

    The objective of this work is to study necking behavior of Zircaloy-2 in a tension test in which the temperature range varies from 170 0 C to 620 0 C by means of a model. This model provides strain rate variations in the beginning of necking and the parameters in the / necking stability. A new parameter Ψ is presented which permits necking / stability description in metals by means of a simple tension test. It is also proceeded a behavioral study of ε versus ε curve after necking formation. (author)

  12. Uranium

    International Nuclear Information System (INIS)

    Whillans, R.T.

    1981-01-01

    Events in the Canadian uranium industry during 1980 are reviewed. Mine and mill expansions and exploration activity are described, as well as changes in governmental policy. Although demand for uranium is weak at the moment, the industry feels optimistic about the future. (LL)

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

    International Nuclear Information System (INIS)

    Delafosse, Jacques.

    1976-01-01

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

  14. Fatigue limit of Zircaloy-2 under variable one-directional tension and temperature 300 deg C; Granica zamora zircaloy-2, pri cisto jednosmerno promenljivom opterecenju (A=1) na zatezanje i temperaturi 300 deg C

    Energy Technology Data Exchange (ETDEWEB)

    Spasic, Z; Simic, G [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1968-11-15

    A vacuum chamber wad designed and constructed. It was suitable for study of materials at higher temperatures in vacuum or controlled atmospheres. Zircaloy-2 fatigue at 300 deg C in argon atmosphere was measured. Character of strain is variable one directional (A=1) tension. Obtained results are presented in tables and in the form of Veler's curve. The obtained fatigue limit was {sigma} - 15 kp/mm{sup 2}. The Locati method was allied as well and fatigue limit value obtained was 15,75 kp/mm{sup 2}. Error calculated in reference to the previous value obtained by classical methods was 5%. Konstruisana je i izvedena vakuum-komora koja se pokazala prikladna za izucavanje osobina materijala na povisenim temperaturama u vakuumu ili kontrolisanim atmosferama. Izvrseno je ispitivanje zamaranja Zircaloy-2 na temperaturi 300 deg C u atmosferi preciscenog argona. Karakter opterecenja je bio cisto jednosmerno promenljivo opterecenje (A=1) na zatezanje. Dobiveni rezultati su dati tabelarno i u obliku Velerove krive. Dobijena je granica zamora {sigma} = 15 kp/mm{sup 2}. Primenjen je i metod Locati-a za priblizno odredjivanje granice zamora i dobijena je vrednost 15,75 kp/mm{sup 2}. Greska u odnosu na prethodnu granicu zamora dobijenu klasicnim metodom iznosi 5% (author)

  15. An assessment of the waterside corrosion and hydrogen pick-up in the zircaloy-2 pressure tubes of PHWR

    International Nuclear Information System (INIS)

    Sah, D.N.

    1992-01-01

    In view of the deleterious effect of hydriding on the operating life of zircaloy-2 pressure tubes in PHWRs there is an urgent need for the assessment of the status of the pressure tubes with respect to corrosion and hydrogen pick-up in the operating PHWRs. A model has been developed for analysing the waterside corrosion and hydrogen pick-up in the zircaloy-2 pressure tubes under reactor operating conditions. This model predicts the axial profiles of oxide layer thickness and hydrogen pick-up in the pressure tubes as a function of the operating time of the reactor. The prediction of hydrogen pick-up by the model in the F-10 pressure tube of RAPS-I have been found to be in good agreement with the measured value of hydrogen content. This report gives a brief description of the model and its predictions on the present status of hydrogen pick-up in the pressure tubes of lead reactor RAPS-II. (author). 6 refs., 5 figs., 2 tabs

  16. Uranium

    Energy Technology Data Exchange (ETDEWEB)

    Williams, R M

    1976-01-01

    Evidence of expanding markets, improved prices and the short supply of uranium became abundantly clear in 1975, providing the much needed impetus for widespread activity in all phases of uranium operations. Exploration activity that had been at low levels in recent years in Canada was evident in most provinces as well as the Northwest Territories. All producers were in the process of expanding their uranium-producing facilities. Canada's Atomic Energy Control Board (AECB) by year-end had authorized the export of over 73,000 tons of U/sub 3/0/sub 8/ all since September 1974, when the federal government announced its new uranium export guidelines. World production, which had been in the order of 25,000 tons of U/sub 3/0/sub 8/ annually, was expected to reach about 28,000 tons in 1975, principally from increased output in the United States.

  17. Uranium

    International Nuclear Information System (INIS)

    Perkin, D.J.

    1982-01-01

    Developments in the Australian uranium industry during 1980 are reviewed. Mine production increased markedly to 1841 t U 3 O 8 because of output from the new concentrator at Nabarlek and 1131 t of U 3 O 8 were exported at a nominal value of $37.19/lb. Several new contracts were signed for the sale of yellowcake from Ranger and Nabarlek Mines. Other developments include the decision by the joint venturers in the Olympic Dam Project to sink an exploration shaft and the release of an environmental impact statement for the Honeymoon deposit. Uranium exploration expenditure increased in 1980 and additions were made to Australia's demonstrated economic uranium resources. A world review is included

  18. Uranium

    International Nuclear Information System (INIS)

    Gabelman, J.W.; Chenoweth, W.L.; Ingerson, E.

    1981-01-01

    The uranium production industry is well into its third recession during the nuclear era (since 1945). Exploration is drastically curtailed, and many staffs are being reduced. Historical market price production trends are discussed. A total of 3.07 million acres of land was acquired for exploration; drastic decrease. Surface drilling footage was reduced sharply; an estimated 250 drill rigs were used by the uranium industry during 1980. Land acquisition costs increased 8%. The domestic reserve changes are detailed by cause: exploration, re-evaluation, or production. Two significant discoveries of deposits were made in Mohave County, Arizona. Uranium production during 1980 was 21,850 short tons U 3 O 8 ; an increase of 17% from 1979. Domestic and foreign exploration highlights were given. Major producing areas for the US are San Juan basin, Wyoming basins, Texas coastal plain, Paradox basin, northeastern Washington, Henry Mountains, Utah, central Colorado, and the McDermitt caldera in Nevada and Oregon. 3 figures, 8 tables

  19. In-cell facility for performing mechanical-property tests on irradiated cladding

    International Nuclear Information System (INIS)

    Yaggee, F.L.; Haglund, R.C.; Mattas, R.F.

    1978-11-01

    A new facility was developed for testing cladding sections of LWR fuel rods. This facility and the accompanying test procedures have improved the level of in-cell mechanical-testing capabilities, making them comparable to existing capabilities for unirradiated cladding. The new facility is currently being used to study the susceptibility of irradiated Zircaloy cladding from LWR fuel rods to iodine stress-corrosion cracking. Preliminary testing results indicate a systematic effect of temperature, stress and irradiation on the susceptibility of annealed and stress-relieved Zircaloy-2. Experimental data obtained to date are being used to develop a stress-corrosion cracking model for LWR fuel rod failure. SEM examination of the undisturbed fracture surface of specimens that failed by pinhole leakage provides useful information on crack propagation and morphology

  20. Effects of δ-hydride precipitation at a crack tip on crack propagation in delayed hydride cracking of Zircaloy-2

    Energy Technology Data Exchange (ETDEWEB)

    Kubo, T., E-mail: kubo@nfd.co.jp [Nippon Nuclear Fuel Development Co., Ltd., 2163 Narita-cho, Oarai-machi, Ibaraki 311-1313 (Japan); Kobayashi, Y. [M.O.X. Co., Ltd., 1828-520 Hirasu-cho, Mito, Ibaraki 311-0853 (Japan)

    2013-08-15

    Highlights: • Steady state crack velocity of delayed hydride cracking in Zircaloy-2 was analyzed. • A large stress peak is induced at an end of hydride by volume expansion of hydride. • Hydrogen diffuses to the stress peak, thereby accelerating steady hydride growth. • Crack velocity was estimated from the calculated hydrogen flux into the stress peak. • There was good agreement between calculation results and experimental data. -- Abstract: Delayed hydride cracking (DHC) of Zircaloy-2 is one possible mechanism for the failure of boiling water reactor fuel rods in ramp tests at high burnup. Analyses were made for hydrogen diffusion around a crack tip to estimate the crack velocity of DHC in zirconium alloys, placing importance on effects of precipitation of δ-hydride. The stress distribution around the crack tip is significantly altered by precipitation of hydride, which was strictly analyzed using a finite element computer code. Then, stress-driven hydrogen diffusion under the altered stress distribution was analyzed by a differential method. Overlapping of external stress and hydride precipitation at a crack tip induces two stress peaks; one at a crack tip and the other at the front end of the hydride precipitate. Since the latter is larger than the former, more hydrogen diffuses to the front end of the hydride precipitate, thereby accelerating hydride growth compared with that in the absence of the hydride. These results indicated that, after hydride was formed in front of the crack tip, it grew almost steadily accompanying the interaction of hydrogen diffusion, hydride growth and the stress alteration by hydride precipitation. Finally, crack velocity was estimated from the calculated hydrogen flux into the crack tip as a function of temperature, stress intensity factor and material strength. There was qualitatively good agreement between calculation results and experimental data.

  1. Uranium

    International Nuclear Information System (INIS)

    Anon.

    1983-01-01

    Recent decisions by the Australian Government will ensure a significant expansion of the uranium industry. Development at Roxby Downs may proceed and Ranger may fulfil two new contracts but the decision specifies that apart from Roxby Downs, no new mines should be approved. The ACTU maintains an anti-uranium policy but reaction to the decision from the trade union movement has been muted. The Australian Science and Technology Council (ASTEC) has been asked by the Government to conduct an inquiry into a number of issues relating to Australia's role in the nuclear fuel cycle. The inquiry will examine in particular Australia's nuclear safeguards arrangements and the adequacy of existing waste management technology. In two additional decisions the Government has dissociated itself from a study into the feasibility of establishing an enrichment operation and has abolished the Uranium Advisory Council. Although Australian reserves account for 20% of the total in the Western World, Australia accounts for a relatively minor proportion of the world's uranium production

  2. Uranium

    International Nuclear Information System (INIS)

    Anon.

    1981-01-01

    The French Government has decided to freeze a substantial part of its nuclear power programme. Work has been halted on 18 reactors. This power programme is discussed, as well as the effect it has on the supply of uranium by South Africa

  3. Critical cladding radius for hybrid cladding modes

    Science.gov (United States)

    Guyard, Romain; Leduc, Dominique; Lupi, Cyril; Lecieux, Yann

    2018-05-01

    In this article we explore some properties of the cladding modes guided by a step-index optical fiber. We show that the hybrid modes can be grouped by pairs and that it exists a critical cladding radius for which the modes of a pair share the same electromagnetic structure. We propose a robust method to determine the critical cladding radius and use it to perform a statistical study on the influence of the characteristics of the fiber on the critical cladding radius. Finally we show the importance of the critical cladding radius with respect to the coupling coefficient between the core mode and the cladding modes inside a long period grating.

  4. Uranium

    International Nuclear Information System (INIS)

    Battey, G.C.; McKay, A.D.

    1988-01-01

    Production for 1986 was 4899 t U 3 O 8 (4154 t U), 30% greater than in 1985, mainly because of a 39% increase in production at Ranger. Exports for 1986 were 4166 t U 3 O 8 at an average f.o.b. unit value of $40.57/lb U 3 O 8 . Private exploration expenditure for uranium in Australia during the 1985-86 fiscal year was $50.2 million. Plans were announced to increase the nominal capacity of the processing plant at Ranger from 3000 t/year U 3 O 8 to 4500 t and later to 6000 t/year. Construction and initial mine development at Olympic Dam began in March. Production is planned for mid 1988 at an annual rate of 2000 t U 3 O 8 , 30 000 t Cu, and 90 000 oz (2800 kg) Au. The first long-term sales agreement was concluded in September 1986. At the Manyingee deposit, testing of the alkaline solution mining method was completed, and the treatment plant was dismantled. Spot market prices (in US$/lb U 3 O 8 ) quoted by Nuexco were generally stable. From January-October the exchange value fluctuated from US$17.00-US$17.25; for November and December it was US$16.75. Australia's Reasonably Assured Resources of uranium recoverable at less than US$80/kg U at December 1986 were estimated as 462 000 t U, 3000 t U less than in 1985. This represents 30% of the total low-cost RAR in the WOCA (World Outside the Centrally Planned Economy Areas) countries. Australia also has 257 000 t U in the low-cost Estimated Additional Resources Category I, 29% of the WOCA countries' total resources in this category

  5. CREEP STRAIN CORRELATION FOR IRRADIATED CLADDING

    International Nuclear Information System (INIS)

    P. Macheret

    2001-01-01

    In an attempt to predict the creep deformation of spent nuclear fuel cladding under the repository conditions, different correlations have been developed. One of them, which will be referred to as Murty's correlation in the following, and whose expression is given in Henningson (1998), was developed on the basis of experimental points related to unirradiated Zircaloy cladding (Henningson 1998, p. 56). The objective of this calculation is to adapt Murty's correlation to experimental points pertaining to irradiated Zircaloy cladding. The scope of the calculation is provided by the range of experimental parameters characterized by Zircaloy cladding temperature between 292 C and 420 C, hoop stress between 50 and 630 MPa, and test time extending to 8000 h. As for the burnup of the experimental samples, it ranges between 0.478 and 64 MWd/kgU (i.e., megawatt day per kilogram of uranium), but this is not a parameter of the adapted correlation

  6. Nucleation and growth of intermetallic precipitates in Zircaloy-2 and zircaloy-4 and correlation to nodular corrosion behavior

    International Nuclear Information System (INIS)

    Maussner, G.; Steinberg, E.; Tenckhoff, E.

    1987-01-01

    One of the fundamental aspects in the history of the development of zirconium alloys for nuclear applications is the corrosion behavior under in-pile conditions. In boiling-water reactors (BWRs) and pressurized-water reactors (PWRs) the zirconium alloys Zircaloy-2 and Zircaloy-4 are the most commonly used materials, permitting attainment of a very high level of integrity and reliability. Nevertheless, efforts are required to optimize these well-established alloys with regard to their resistance to nodular corrosion, where improvements will give long-term advantages in fuel integrity and fuel economy. Phenomenological studies allow correlation of the nodular corrosion behavior with the morphological appearance of precipitated intermetallic phases in the microstructures of Zry-2 and Zry-4. To understand the fundamental processes of precipitation, particle nucleation-and-growth studies were made with Zry-2 and Zry-4 in different fabrication dimensions and with variations in β-quenching rates followed by isothermal and isochronical heat treatments. The microstructural characteristics of the precipitates were investigated by optical and transmission-electron microscopy. The macroscopic behavior was studied by electrical-resistivity measurements and hardness measurements. The nodular-corrosion susceptibility was determined by weight-gain and nodule distribution measurements after a 500 0 C laboratory-autoclave test

  7. TEM/STEM study of Zircaloy-2 with protective FeAl(Cr) layers under simulated BWR environment and high-temperature steam exposure

    Science.gov (United States)

    Park, Donghee; Mouche, Peter A.; Zhong, Weicheng; Mandapaka, Kiran K.; Was, Gary S.; Heuser, Brent J.

    2018-04-01

    FeAl(Cr) thin-film depositions on Zircaloy-2 were studied using transmission electron microscopy (TEM) and scanning transmission electron microscopy (STEM) with respect to oxidation behavior under simulated boiling water reactor (BWR) conditions and high-temperature steam. Columnar grains of FeAl with Cr in solid solution were formed on Zircaloy-2 coupons using magnetron sputtering. NiFe2O4 precipitates on the surface of the FeAl(Cr) coatings were observed after the sample was exposed to the simulated BWR environment. High-temperature steam exposure resulted in grain growth and consumption of the FeAl(Cr) layer, but no delamination at the interface. Outward Al diffusion from the FeAl(Cr) layer occurred during high-temperature steam exposure (700 °C for 3.6 h) to form a 100-nm-thick alumina oxide layer, which was effective in mitigating oxidation of the Zircaloy-2 coupons. Zr intermetallic precipitates formed near the FeAl(Cr) layer due to the inward diffusion of Fe and Al. The counterflow of vacancies in response to the Al and Fe diffusion led to porosity within the FeAl(Cr) layer.

  8. Annex 3 - Testing the microstructure of the fuel element with metal uranium with aluminium cladding; Prilog 3 - Ispitivanje mikrostrukture gorivnog elementa na bazi metalnog urana sa aluminijumskom kosuljicom

    Energy Technology Data Exchange (ETDEWEB)

    Milosevic, S; Momcilovic, I [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1961-12-15

    Based on the uranium microstructure in the fuel element, i.e. grain size, presence of other phases, porosity and the position of metal connections, some data could be obtained about the mechanical and thermal treatment of the uranium during fuel element fabrication. The mentioned characteristics could indicate the behaviour of the fuel under reactor operating conditions taking into account its stable dimensions and roughness of its surfaces. Na osnovu mikrostrukture urana u gorivnom elementu, odnosno velicine i oblika njegovih kristalnih zrna, prisustva drugih faza, pozoziteta kao i rasporeda metalnih ukljucaka, mogu se dobiti izvesni podaci o tome koje je postupke mehanickog i termickog tretiranja pretrpeo uran u pripremi za gorivni elemenat. Prethodne karakteristike takodje mogu da daju predstavu o tome kako ce se gorivo ponasati u uslovima rada reaktora, s obzirom na njegovu dimenzionu stabilnost i ogrubljivanje njegove povrsine (author)

  9. High loading uranium plate

    International Nuclear Information System (INIS)

    Wiencek, T.C.; Domagala, R.F.; Thresh, H.R.

    1990-01-01

    Two embodiments of a high uranium fuel plate are disclosed which contain a meat comprising structured uranium compound confined between a pari of diffusion bonded ductile metal cladding plates uniformly covering the meat, the meat hiving a uniform high fuel loading comprising a content of uranium compound greater than about 45 Vol. % at a porosity not greater than about 10 Vol. %. In a first embodiment, the meat is a plurality of parallel wires of uranium compound. In a second embodiment, the meat is a dispersion compact containing uranium compound. The fuel plates are fabricated by a hot isostatic pressing process

  10. Characteristics of hydride precipitation and reorientation in spent-fuel cladding

    International Nuclear Information System (INIS)

    Chung, H. M.; Strain, R. V.; Billone, M. C.

    2000-01-01

    The morphology, number density, orientation, distribution, and crystallographic aspects of Zr hydrides in Zircaloy fuel cladding play important roles in fuel performance during all phases before and after discharge from the reactor, i.e., during normal operation, transient and accident situations in the reactor, temporary storage in a dry cask, and permanent storage in a waste repository. In the past, partly because of experimental difficulties, hydriding behavior in irradiated fuel cladding has been investigated mostly by optical microscopy (OM). In the present study, fundamental metallurgical and crystallographic characteristics of hydride precipitation and reorientation were investigated on the microscopic level by combined techniques of OM and transmission electron and scanning electron microscopy (TEM and SEM) of spent-fuel claddings discharged from several boiling and pressurized water reactors (BWRs and PWRs). Defueled sections of standard and Zr-lined Zircaloy-2 fuel claddings, irradiated to fluences of ∼3.3 x 10 21 n cm -2 and ∼9.2 x 10 21 n cm -2 (E > 1 MeV), respectively, were obtained from spent fuel rods discharged from two BWRs. Sections of standard and low-tin Zircaloy-4 claddings, irradiated to fluences of ∼4.4 x 10 21 n cm -2 , ∼5.9 x 10 21 n cm -2 , and ∼9.6 x 10 21 n cm -2 (E > 1 MeV) in three PWRs, were also obtained. Microstructural characteristics of hydrides were analyzed in as-irradiated condition and after gas-pressurization-burst or expanding-mandrel tests at 292-325 C in Ar for some of the spent-fuel claddings. Analyses were also conducted of hydride habit plane, morphology, and reorientation characteristics on unirradiated Zircaloy-4 cladding that contained dense radial hydrides. Reoriented hydrides in the slowly cooled unirradiated cladding were produced by expanding-mandrel loading

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

    International Nuclear Information System (INIS)

    Boyle, R.F.

    1987-01-01

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

  12. Performance of refractory alloy-clad fuel pins

    International Nuclear Information System (INIS)

    Dutt, D.S.; Cox, C.M.; Millhollen, M.K.

    1984-12-01

    This paper discusses objectives and basic design of two fuel-cladding tests being conducted in support of SP-100 technology development. Two of the current space nuclear power concepts use conventional pin type designs, where a coolant removes the heat from the core and transports it to an out-of-core energy conversion system. An extensive irradiation testing program was conducted in the 1950's and 1960's to develop fuel pins for space nuclear reactors. The program emphasized refractory metal clad uranium nitride (UN), uranium carbide (UC), uranium oxide (UO 2 ), and metal matrix fuels (UCZr and BeO-UO 2 ). Based on this earlier work, studies presented here show that UN and UO 2 fuels in conjunction with several refractory metal cladding materials demonstrated high potential for meeting space reactor requirements and that UC could serve as an alternative but higher risk fuel

  13. Synchrotron X-ray diffraction investigations on strains in the oxide layer of an irradiated Zircaloy fuel cladding

    International Nuclear Information System (INIS)

    Chollet, Mélanie; Valance, Stéphane; Abolhassani, Sousan; Stein, Gene; Grolimund, Daniel; Martin, Matthias; Bertsch, Johannes

    2017-01-01

    For the first time the microstructure of the oxide layer of a Zircaloy-2 cladding after 9 cycles of irradiation in a boiling water reactor has been analyzed with synchrotron micro-X-ray diffraction. Crystallographic strains of the monoclinic and to some extent of the tetragonal ZrO 2 are depicted through the thick oxide layer. Thin layers of sub-oxide at the oxide-metal interface as found for autoclave-tested samples and described in the literature, have not been observed in this material maybe resulting from irradiation damage. Shifts of selected diffraction peaks of the monoclinic oxide show that the uniform strain produced during oxidation is orientated in the lattice and displays variations along the oxide layer. Diffraction peaks and their shifts from families of diffracting planes could be translated into a virtual tensor. This virtual tensor exhibits changes through the oxide layer passing by tensile or compressive components. - Highlights: •A Zircaloy-2 cladding irradiated 9 cycles was investigated thanks to synchrotron X-ray diffraction. •Microstructure and uniform strain through the oxide layer is revealed. •The m-ZrO 2 uniform strain is oriented presenting compression along the (−111) plane. •Virtual tensor is built based on reflecting planes of families of grains. •Tensor components vary from tensile to compressive along the oxide layer.

  14. Synchrotron X-ray diffraction investigations on strains in the oxide layer of an irradiated Zircaloy fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Chollet, Mélanie, E-mail: melanie.chollet@psi.ch [Paul Scherrer Institute, NES, 5232 Villigen (Switzerland); Valance, Stéphane; Abolhassani, Sousan; Stein, Gene [Paul Scherrer Institute, NES, 5232 Villigen (Switzerland); Grolimund, Daniel [Paul Scherrer Institute, SLS, 5232 Villigen (Switzerland); Martin, Matthias; Bertsch, Johannes [Paul Scherrer Institute, NES, 5232 Villigen (Switzerland)

    2017-05-15

    For the first time the microstructure of the oxide layer of a Zircaloy-2 cladding after 9 cycles of irradiation in a boiling water reactor has been analyzed with synchrotron micro-X-ray diffraction. Crystallographic strains of the monoclinic and to some extent of the tetragonal ZrO{sub 2} are depicted through the thick oxide layer. Thin layers of sub-oxide at the oxide-metal interface as found for autoclave-tested samples and described in the literature, have not been observed in this material maybe resulting from irradiation damage. Shifts of selected diffraction peaks of the monoclinic oxide show that the uniform strain produced during oxidation is orientated in the lattice and displays variations along the oxide layer. Diffraction peaks and their shifts from families of diffracting planes could be translated into a virtual tensor. This virtual tensor exhibits changes through the oxide layer passing by tensile or compressive components. - Highlights: •A Zircaloy-2 cladding irradiated 9 cycles was investigated thanks to synchrotron X-ray diffraction. •Microstructure and uniform strain through the oxide layer is revealed. •The m-ZrO{sub 2} uniform strain is oriented presenting compression along the (−111) plane. •Virtual tensor is built based on reflecting planes of families of grains. •Tensor components vary from tensile to compressive along the oxide layer.

  15. In situ measurement of the effect of LiOH on the stability of zircaloy-2 surface film in PWR water

    International Nuclear Information System (INIS)

    Saario, T.; Taehtinen, S.

    1997-01-01

    Surface films on the metals play a major role in corrosion assisted cracking. A new method called Contact Electric Resistance (CER) method has been recently developed for in situ measurement of the electric resistance of surface films in high temperature and high pressure environments. The technique has been used to determine in situ the electric resistance of films on metals when in contact with water and dissolved anions, during formation and destruction of oxides and hydrides and during electroplating of metals. Electric resistance data can be measured with a frequency of the order of one hertz, which makes it possible to investigate in situ the kinetics of surface film related processes which are dependent on the environment, temperature, pH and electrochemical potential. This paper presents the results of the CER investigation on the effects of LiOH on the stability of Zircaloy-2 surface film in water with 2000 ppm H 3 BO 3 . At 300 deg. C the LiOH concentrations higher than 10 -2 M (roughly 70 ppm of Li + ) were found to markedly reduce the electric resistance of the Zircaloy-2 surface film during a test period of less than two hours. The decrease of the film resistance is very abrupt, possibly indicating a phase transformation. Moreover, the advantages of the CER technique over the other competing techniques which rely on the measurement of current are discussed. (author)

  16. In situ measurement of the effect of LiOH on the stability of zircaloy-2 surface film in PWR water

    Energy Technology Data Exchange (ETDEWEB)

    Saario, T; Taehtinen, S [Technical Research Centre of Finland, Espoo (Finland)

    1997-02-01

    Surface films on the metals play a major role in corrosion assisted cracking. A new method called Contact Electric Resistance (CER) method has been recently developed for in situ measurement of the electric resistance of surface films in high temperature and high pressure environments. The technique has been used to determine in situ the electric resistance of films on metals when in contact with water and dissolved anions, during formation and destruction of oxides and hydrides and during electroplating of metals. Electric resistance data can be measured with a frequency of the order of one hertz, which makes it possible to investigate in situ the kinetics of surface film related processes which are dependent on the environment, temperature, pH and electrochemical potential. This paper presents the results of the CER investigation on the effects of LiOH on the stability of Zircaloy-2 surface film in water with 2000 ppm H{sub 3}BO{sub 3}. At 300 deg. C the LiOH concentrations higher than 10{sup -2} M (roughly 70 ppm of Li{sup +}) were found to markedly reduce the electric resistance of the Zircaloy-2 surface film during a test period of less than two hours. The decrease of the film resistance is very abrupt, possibly indicating a phase transformation. Moreover, the advantages of the CER technique over the other competing techniques which rely on the measurement of current are discussed. (author).

  17. Oxide particle size distribution from shearing irradiated and unirradiated LWR fuels in Zircaloy and stainless steel cladding: significance for risk assessment

    International Nuclear Information System (INIS)

    Davis, W. Jr.; West, G.A.; Stacy, R.G.

    1979-01-01

    Sieve fractionation was performed with oxide particles dislodged during shearing of unirradiated or irradiated fuel bundles or single rods of UO 2 or 96 to 97% ThO 2 --3 to 4% UO 2 . Analyses of these data by nonlinear least-squares techniques demonstrated that the particle size distribution is lognormal. Variables involved in the numerical analyses include lognormal median size, lognormal standard deviation, and shear cut length. Sieve-fractionation data are presented for unirradiated bundles of stainless-steel-clad or Zircaloy-2-clad UO 2 or ThO 2 --UO 2 sheared into lengths from 0.5 to 2.0 in. Data are also presented for irradiated single rods (sheared into lengths of 0.25 to 2.0 in.) of Zircaloy-2-clad UO 2 from BWRs and of Zircaloy-4-clad UO 2 from PWRs. Median particle sizes of UO 2 from shearing irradiated stainless-steel-clad fuel ranged from 103 to 182 μm; particle sizes of ThO 2 --UO 2 , under these same conditions, ranged from 137 to 202 μm. Similarly, median particle sizes of UO 2 from shearing unirradiated Zircaloy-2-clad fuel ranged from 230 to 957 μm. Irradiation levels of fuels from reactors ranged from 9,000 to 28,000 MWd/MTU. In general, particle sizes from shearing these irradiated fuels are larger than those from the unirradiated fuels. In addition, variations in particle size parameters pertaining to samples of a single vendor varied as much as those between different vendors. The fraction of fuel dislodged from the cladding is nearly proportional to the reciprocal of the shear cut length, until the cut length attains some minimum value below which all fuel is dislodged. Particles of fuel are generally elongated with a long-to-short axis ratio usually less than 3. Using parameters of the lognormal distribution deduced from experimental data, realistic estimates can be made of fractions of dislodged fuel having dimensions less than specified values

  18. Oxide particle size distribution from shearing irradiated and unirradiated LWR fuels in Zircaloy and stainless steel cladding: significance for risk assessment

    Energy Technology Data Exchange (ETDEWEB)

    Davis, W. Jr.; West, G.A.; Stacy, R.G.

    1979-03-22

    Sieve fractionation was performed with oxide particles dislodged during shearing of unirradiated or irradiated fuel bundles or single rods of UO/sub 2/ or 96 to 97% ThO/sub 2/--3 to 4% UO/sub 2/. Analyses of these data by nonlinear least-squares techniques demonstrated that the particle size distribution is lognormal. Variables involved in the numerical analyses include lognormal median size, lognormal standard deviation, and shear cut length. Sieve-fractionation data are presented for unirradiated bundles of stainless-steel-clad or Zircaloy-2-clad UO/sub 2/ or ThO/sub 2/--UO/sub 2/ sheared into lengths from 0.5 to 2.0 in. Data are also presented for irradiated single rods (sheared into lengths of 0.25 to 2.0 in.) of Zircaloy-2-clad UO/sub 2/ from BWRs and of Zircaloy-4-clad UO/sub 2/ from PWRs. Median particle sizes of UO/sub 2/ from shearing irradiated stainless-steel-clad fuel ranged from 103 to 182 ..mu..m; particle sizes of ThO/sub 2/--UO/sub 2/, under these same conditions, ranged from 137 to 202 ..mu..m. Similarly, median particle sizes of UO/sub 2/ from shearing unirradiated Zircaloy-2-clad fuel ranged from 230 to 957 ..mu..m. Irradiation levels of fuels from reactors ranged from 9,000 to 28,000 MWd/MTU. In general, particle sizes from shearing these irradiated fuels are larger than those from the unirradiated fuels; however, unirradiated fuel from vendors was not available for performing comparative shearing experiments. In addition, variations in particle size parameters pertaining to samples of a single vendor varied as much as those between different vendors. The fraction of fuel dislodged from the cladding is nearly proportional to the reciprocal of the shear cut length, until the cut length attains some minimum value below which all fuel is dislodged. Particles of fuel are generally elongated with a long-to-short axis ratio usually less than 3. Using parameters of the lognormal distribution estimates can be made of fractions of dislodged fuel having

  19. Evaluation of Corrosion of Aluminum Based Reactor Fuel Cladding Materials During Dry Storage

    International Nuclear Information System (INIS)

    Peacock, H.B. Jr.

    1999-01-01

    This report provides an evaluation of the corrosion behavior of aluminum cladding alloys and aluminum-uranium alloys at conditions relevant to dry storage. The details of the corrosion program are described and the results to date are discussed

  20. Application of non-destructive liner thickness measurement technique for manufacturing and inspection process of zirconium lined cladding tube

    International Nuclear Information System (INIS)

    Nakazawa, Norio; Fukuda, Akihiro; Fujii, Noritsugu; Inoue, Koichi

    1986-01-01

    Recently, in order to meet the difference of electric power demand owing to electric power situation, large scale load following operation has become necessary. Therefore, the development of the cladding tubes which withstand power variation has been carried out, as the result, zirconium-lined zircaloy 2 cladding tubes have been developed. In order to reduce the sensitivity to stress corrosion cracking, these zirconium-lined cladding tubes require uniform liner thickness over the whole surface and whole length. Kobe Steel Ltd. developed the nondestructive liner thickness measuring technique based on ultrasonic flaw detection technique and eddy current flaw detection technique. These equipments were applied to the manufacturing and inspection processes of the zirconium-lined cladding tubes, and have demonstrated superiority in the control and assurance of the liner thickness of products. Zirconium-lined cladding tubes, the development of the measuring technique for guaranteeing the uniform liner thickness and the liner thickness control in the manufacturing and inspection processes are described. (Kako, I.)

  1. Laser cladding with powder

    NARCIS (Netherlands)

    Schneider, M.F.; Schneider, Marcel Fredrik

    1998-01-01

    This thesis is directed to laser cladding with powder and a CO2 laser as heat source. The laser beam intensity profile turned out to be an important pa6 Summary rameter in laser cladding. A numerical model was developed that allows the prediction of the surface temperature distribution that is

  2. Compatibility studies on Mo-coating systems for nuclear fuel cladding applications

    Science.gov (United States)

    Koh, Huan Chin; Hosemann, Peter; Glaeser, Andreas M.; Cionea, Cristian

    2017-12-01

    To improve the safety factor of nuclear power plants in accident scenarios, molybdenum (Mo), with its high-temperature strength, is proposed as a potential fuel-cladding candidate. However, Mo undergoes rapid oxidation and sublimation at elevated temperatures in oxygen-rich environments. Thus, it is necessary to coat Mo with a protective layer. The diffusional interactions in two systems, namely, Zircaloy-2 (Zr2) on a Mo tube, and iron-chromium-aluminum (FeCrAl) on a Mo rod, were studied by aging coated Mo substrates in high vacuum at temperatures ranging from 650 °C to 1000° for 1000 h. The specimens were characterized using scanning electron microscopy (SEM), energy-dispersive spectrometry (EDS) and nanoindentation. In both systems, pores in the coating increased in size and number with increasing temperature over time, and cracks were also observed; intermetallic phases formed between the Mo and its coatings.

  3. Young's modulus of crystal bar zirconium and zirconium alloys (zircaloy-2, zircaloy-4, zirconium-2.5wt% niobium) to 1000 K

    International Nuclear Information System (INIS)

    Rosinger, H.E.; Ritchie, I.G.; Shillinglaw, A.J.

    1975-09-01

    This report contains experimentally determined data on the dynamic elastic moduli of zircaloy-2, zircaloy-4, zirconium-2.5wt% niobium and Marz grade crystal bar zirconium. Data on both the dynamic Young's moduli and shear moduli of the alloys have been measured at room temperature and Young's modulus as a function of temperature has been determined over the temperature range 300 K to 1000 K. In every case, Young's modulus decreases linearly with increasing temperature and is expressed by an empirical equation fitted to the data. Differences in Young's modulus values determined from specimens with longitudinal axes parallel and perpendicular to the rolling direction are small, as are the differences between Young's moduli determined from strip, bar stock and fuel sheathing. (author)

  4. A study on the delayed hydride cracking mechanism in cold worked Zr-2.5Nb, heat treated Zr-2.5Nb and zircaloy-2 pressure tubes

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Kwang Sik

    1992-02-15

    Cold worked Zr-2.5Nb, heat treated Zr-2.5Nb and Zircaloy-2 pressure tubes were hydrided to the hydrogen concentration of 68 ppm, 49 ppm and 242-411 ppm, respectively, and compact tension specimens were machined from the hydrided materials. The crack growth rate by delayed hydride cracking was measured by potential drop method at various temperatures on the above mentioned three types of specimens. The activation energy obtained were 43 KJ/mol for cold worked Zr-2.5Nb and 37 KJ/mol for heat treated Zr-2.5Nb, which were in good agreements with that of Coleman (1977), while they were lower than the activation energy of 65.5 KJ/mol obtained by Simpson-puls (1979) and 71.5 KJ/mol by Ambler (1984). The DHC growth rate in Zircaloy-2 were about one fifth of that of Zr-2.5Nb, which is due to the texture and material strength effects. Striations which indicate stepwise DHC growth were observed at fracture surface by scanning electron microscope and unsymmetric crack tunnellings were also observed, which seems to be due to the difference in hydrogen diffusion rate caused by the difference in stress fields between inner and outer surface. The comparison of test results with the DHC growth rate calculated by Simpson-puls model showed good agreement at high temperatures, whereas at the lower temperatures the crack growth rates were 2.5 times higher than the calculated values.

  5. A study on the delayed hydride cracking mechanism in cold worked Zr-2.5Nb, heat treated Zr-2.5Nb and zircaloy-2 pressure tubes

    International Nuclear Information System (INIS)

    Choi, Kwang Sik

    1992-02-01

    Cold worked Zr-2.5Nb, heat treated Zr-2.5Nb and Zircaloy-2 pressure tubes were hydrided to the hydrogen concentration of 68 ppm, 49 ppm and 242-411 ppm, respectively, and compact tension specimens were machined from the hydrided materials. The crack growth rate by delayed hydride cracking was measured by potential drop method at various temperatures on the above mentioned three types of specimens. The activation energy obtained were 43 KJ/mol for cold worked Zr-2.5Nb and 37 KJ/mol for heat treated Zr-2.5Nb, which were in good agreements with that of Coleman (1977), while they were lower than the activation energy of 65.5 KJ/mol obtained by Simpson-puls (1979) and 71.5 KJ/mol by Ambler (1984). The DHC growth rate in Zircaloy-2 were about one fifth of that of Zr-2.5Nb, which is due to the texture and material strength effects. Striations which indicate stepwise DHC growth were observed at fracture surface by scanning electron microscope and unsymmetric crack tunnellings were also observed, which seems to be due to the difference in hydrogen diffusion rate caused by the difference in stress fields between inner and outer surface. The comparison of test results with the DHC growth rate calculated by Simpson-puls model showed good agreement at high temperatures, whereas at the lower temperatures the crack growth rates were 2.5 times higher than the calculated values

  6. Stone cladding engineering

    National Research Council Canada - National Science Library

    Camposinhos, Rui de Sousa

    2014-01-01

    .... Straightforward formulae are provided for computing action on cladding, with special emphasis on the effect of seismic forces, including an extensive general methodology applied to non-structural elements...

  7. Cladding creepdown under compression

    International Nuclear Information System (INIS)

    Hobson, D.O.

    1977-01-01

    Light-water power reactors use Zircaloy tubing as cladding to contain the UO 2 fuel pellets. In-service operating conditions impose an external hydrostatic force on the cladding, causing it to creep down into eventual contact with the fuel. Knowledge of the rate of such creepdown is of great importance to modelers of fuel element performance. An experimental system was devised for studying creepdown that meets several severe requirements by providing (1) correct stress state, (2) multiple positions for measuring radial displacement of the cladding surface, (3) high-precision data, and (4) an experimental configuration compact enough to fit in-reactor. A microcomputer-controlled, eddy-current monitoring system was developed for this study and has proven highly successful in measuring cladding deformation with time at temperatures of 371 0 C (700 0 F) and higher, and at pressures as high as 21 MPa

  8. Influence of uranium hydride oxidation on uranium metal behaviour

    International Nuclear Information System (INIS)

    Patel, N.; Hambley, D.; Clarke, S.A.; Simpson, K.

    2013-01-01

    This work addresses concerns that the rapid, exothermic oxidation of active uranium hydride in air could stimulate an exothermic reaction (burning) involving any adjacent uranium metal, so as to increase the potential hazard arising from a hydride reaction. The effect of the thermal reaction of active uranium hydride, especially in contact with uranium metal, does not increase in proportion with hydride mass, particularly when considering large quantities of hydride. Whether uranium metal continues to burn in the long term is a function of the uranium metal and its surroundings. The source of the initial heat input to the uranium, if sufficient to cause ignition, is not important. Sustained burning of uranium requires the rate of heat generation to be sufficient to offset the total rate of heat loss so as to maintain an elevated temperature. For dense uranium, this is very difficult to achieve in naturally occurring circumstances. Areas of the uranium surface can lose heat but not generate heat. Heat can be lost by conduction, through contact with other materials, and by convection and radiation, e.g. from areas where the uranium surface is covered with a layer of oxidised material, such as burned-out hydride or from fuel cladding. These rates of heat loss are highly significant in relation to the rate of heat generation by sustained oxidation of uranium in air. Finite volume modelling has been used to examine the behaviour of a magnesium-clad uranium metal fuel element within a bottle surrounded by other un-bottled fuel elements. In the event that the bottle is breached, suddenly, in air, it can be concluded that the bulk uranium metal oxidation reaction will not reach a self-sustaining level and the mass of uranium oxidised will likely to be small in relation to mass of uranium hydride oxidised. (authors)

  9. Influence of uranium hydride oxidation on uranium metal behaviour

    Energy Technology Data Exchange (ETDEWEB)

    Patel, N.; Hambley, D. [National Nuclear Laboratory (United Kingdom); Clarke, S.A. [Sellafield Ltd (United Kingdom); Simpson, K.

    2013-07-01

    This work addresses concerns that the rapid, exothermic oxidation of active uranium hydride in air could stimulate an exothermic reaction (burning) involving any adjacent uranium metal, so as to increase the potential hazard arising from a hydride reaction. The effect of the thermal reaction of active uranium hydride, especially in contact with uranium metal, does not increase in proportion with hydride mass, particularly when considering large quantities of hydride. Whether uranium metal continues to burn in the long term is a function of the uranium metal and its surroundings. The source of the initial heat input to the uranium, if sufficient to cause ignition, is not important. Sustained burning of uranium requires the rate of heat generation to be sufficient to offset the total rate of heat loss so as to maintain an elevated temperature. For dense uranium, this is very difficult to achieve in naturally occurring circumstances. Areas of the uranium surface can lose heat but not generate heat. Heat can be lost by conduction, through contact with other materials, and by convection and radiation, e.g. from areas where the uranium surface is covered with a layer of oxidised material, such as burned-out hydride or from fuel cladding. These rates of heat loss are highly significant in relation to the rate of heat generation by sustained oxidation of uranium in air. Finite volume modelling has been used to examine the behaviour of a magnesium-clad uranium metal fuel element within a bottle surrounded by other un-bottled fuel elements. In the event that the bottle is breached, suddenly, in air, it can be concluded that the bulk uranium metal oxidation reaction will not reach a self-sustaining level and the mass of uranium oxidised will likely to be small in relation to mass of uranium hydride oxidised. (authors)

  10. Corrosion testing of uranium silicide fuel specimens

    International Nuclear Information System (INIS)

    Bourns, W.T.

    1968-09-01

    U 3 Si is the most promising high density natural uranium fuel for water-cooled power reactors. Power reactors fuelled with this material are expected to produce cheaper electricity than those fuelled with uranium dioxide. Corrosion tests in 300 o C water preceded extensive in-reactor performance tests of fuel elements and bundles. Proper heat-treatment of U-3.9 wt% Si gives a U 3 5i specimen which corrodes at less than 2 mg/cm 2 h in 300 o C water. This is an order of magnitude lower than the maximum corrosion rate tolerable in a water-cooled reactor. U 3 Si in a defected unbonded Zircaloy-2 sheath showed only a slow uniform sheath expansion in 300 o C water. All tests were done under isothermal conditions in an out-reactor loop. (author)

  11. Corrosion testing of uranium silicide fuel specimens

    Energy Technology Data Exchange (ETDEWEB)

    Bourns, W T

    1968-09-15

    U{sub 3}Si is the most promising high density natural uranium fuel for water-cooled power reactors. Power reactors fuelled with this material are expected to produce cheaper electricity than those fuelled with uranium dioxide. Corrosion tests in 300{sup o}C water preceded extensive in-reactor performance tests of fuel elements and bundles. Proper heat-treatment of U-3.9 wt% Si gives a U{sub 3}5i specimen which corrodes at less than 2 mg/cm{sup 2} h in 300{sup o}C water. This is an order of magnitude lower than the maximum corrosion rate tolerable in a water-cooled reactor. U{sub 3}Si in a defected unbonded Zircaloy-2 sheath showed only a slow uniform sheath expansion in 300{sup o}C water. All tests were done under isothermal conditions in an out-reactor loop. (author)

  12. Laser cladding of turbine blades

    International Nuclear Information System (INIS)

    Shepeleva, L.; Medres, B.; Kaplan, W.D.; Bamberger, M.

    2000-01-01

    A comparative study of two different techniques for the application of wear-resistant coatings for contact surfaces of shroud shelves of gas turbine engine blades (GTE) has been conducted. Wear-resistant coatings were applied on In713 by laser cladding with direct injection of the cladding powder into the melt pool. Laser cladding was conducted with a TRUMPF-2500, CW-CO 2 laser. The laser cladding was compared with commercially available plasma cladding with wire. Both plasma and laser cladded zones were characterized by optical and scanning electron microscopy. It was found that the laser cladded zone has a higher microhardness value (650-820 HV) compared with that of the plasma treated material (420-440 HV). This is a result of the significant reduction in grain size in the case of laser cladding. Unlike the plasma cladded zones, the laser treated material is free of micropores and microcracks. (orig.)

  13. Initial Cladding Condition

    International Nuclear Information System (INIS)

    Siegmann, E.

    2000-01-01

    The purpose of this analysis is to describe the condition of commercial Zircaloy clad fuel as it is received at the Yucca Mountain Project (YMP) site. Most commercial nuclear fuel is encased in Zircaloy cladding. This analysis is developed to describe cladding degradation from the expected failure modes. This includes reactor operation impacts including incipient failures, potential degradation after reactor operation during spent fuel storage in pool and dry storage and impacts due to transportation. Degradation modes include cladding creep, and delayed hydride cracking during dry storage and transportation. Mechanical stresses from fuel handling and transportation vibrations are also included. This Analysis and Model Report (AMR) does not address any potential damage to assemblies that might occur at the YMP surface facilities. Ranges and uncertainties have been defined. This analysis will be the initial boundary condition for the analysis of cladding degradation inside the repository. In accordance with AP-2.13Q, ''Technical Product Development Planning'', a work plan (CRWMS M andO 2000c) was developed, issued, and utilized in the preparation of this document. There are constraints, caveats and limitations to this analysis. This cladding degradation analysis is based on commercial Pressurized Water Reactor (PWR) fuel with Zircaloy cladding but is applicable to Boiling Water Reactor (BWR) fuel. Reactor operating experience for both PWRs and BWRs is used to establish fuel reliability from reactor operation. It is limited to fuel exposed to normal operation and anticipated operational occurrences (i.e. events which are anticipated to occur within a reactor lifetime), and not to fuel that has been exposed to severe accidents. Fuel burnup projections have been limited to the current commercial reactor licensing environment with restrictions on fuel enrichment, oxide coating thickness and rod plenum pressures. The information provided in this analysis will be used in

  14. Cladding failure margins for metallic fuel in the integral fast reactor

    International Nuclear Information System (INIS)

    Bauer, T.H.; Fenske, G.R.; Kramer, J.M.

    1987-01-01

    The reference fuel for Integral Fast Reactor (IFR) is a ternary U-Pu-Zr alloy with a low swelling austenitic or ferritic stainless steel cladding. It is known that low melting point eutectics may form in such metallic fuel-cladding systems which could contribute to cladding failure under accident conditions. This paper will present recent measurements of cladding eutectic penetration rates for the ternary IFR alloy and will compare these results with earlier eutectic penetration data for other fuel and cladding materials. A method for calculating failure of metallic fuel pins is developed by combining cladding deformation equations with a large strain analysis where the hoop stress is calculated using the instantaneous wall thickness as determined from correlations of the eutectic penetration-rate data. This method is applied to analyze the results of in-reactor and out-of-reactor fuel pin failure tests on uranium-fissium alloy EBR-II Mark-II driver fuel

  15. Electra-Clad

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2006-05-04

    The study relates to the use of building-integrated photovoltaics. The Electra-Clad project sought to use steel-based cladding as a substrate for direct fabrication of a fully integrated solar panel of a design similar to the ICP standard glass-based panel. The five interrelated phases of the project are described. The study successfully demonstrated that the principles of the panel design are achievable and sound. But, despite intensive trials, a commercially realistic solar performance has not been achieved: the main failing was the poor solar conversion efficiency as the active area of the panel was increased in size. The problem lies with the coating used on the steel cladding substrates and it was concluded that a new type of coating will be required. ICP Solar Technologies UK carried out the work under contract to the DTI.

  16. Nuclear fuel cladding material

    International Nuclear Information System (INIS)

    Nakahigashi, Shigeo.

    1982-01-01

    Purpose: To largely improve the durability and the safety of fuel cladding material. Constitution: Diffusion preventive layers, e.g., aluminum or the like are covered on both sides of a zirconium alloy base layer of thin material, and corrosion resistant layers, e.g., copper or the like are covered thereon. This thin plate material is intimately wound in a circularly tubular shape in a plurality of layers to form a fuel cladding tube. With such construction, corrosion of the tube due to fuel and impurity can be prevented by the corrosion resistant layers, and the diffusion of the corrosion resistant material to the zirconium alloy can be prevented by the diffusion preventive layers. Since a plurality of layers are cladded, even if the corrosion resistant layers are damaged or cracked due to stress corrosion, only one layer is damaged or cracked, but the other layers are not affected. (Sekiya, K.)

  17. Fuel clad chemical interactions in fast reactor MOX fuels

    Energy Technology Data Exchange (ETDEWEB)

    Viswanathan, R., E-mail: rvis@igcar.gov.in

    2014-01-15

    Clad corrosion being one of the factors limiting the life of a mixed-oxide fast reactor fuel element pin at high burn-up, some aspects known about the key elements (oxygen, cesium, tellurium, iodine) in the clad-attack are discussed and many Fuel–Clad-Chemical-Interaction (FCCI) models available in the literature are also discussed. Based on its relatively superior predictive ability, the HEDL (Hanford Engineering Development Laboratory) relation is recommended: d/μm = ({0.507 ⋅ [B/(at.% fission)] ⋅ (T/K-705) ⋅ [(O/M)_i-1.935]} + 20.5) for (O/M){sub i} ⩽ 1.98. A new model is proposed for (O/M){sub i} ⩾ 1.98: d/μm = [B/(at.% fission)] ⋅ (T/K-800){sup 0.5} ⋅ [(O/M){sub i}-1.94] ⋅ [P/(W cm{sup −1})]{sup 0.5}. Here, d is the maximum depth of clad attack, B is the burn-up, T is the clad inner surface temperature, (O/M){sub i} is the initial oxygen-to-(uranium + plutonium) ratio, and P is the linear power rating. For fuels with [n(Pu)/n(M = U + Pu)] > 0.25, multiplication factors f are recommended to consider the potential increase in the depth of clad-attack.

  18. Contribution to the study of corrosion of zirconium and zircaloy-2 in superheated steam at 400 deg C (105 kg /cm{sup 2}); Contribution a l'etude de la corrosion du zirconium et du zircaloy-2 dans la vapeur d'eau surchauffee a 400 deg C (105 kg /cm{sup 2})

    Energy Technology Data Exchange (ETDEWEB)

    Coriou, H; Gauduchau, J; Grall, L; Hure, J; Pelras, M [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

    1959-07-01

    The corrosion kinetics of zircaloy-2 in water and steam at temperatures between 300 deg. C and 400 deg. C are represented by a curve sharply divided into two stages separated by a so-called transition point. After a first period of decreasing corrosion rate there follows a second period with much faster kinetics in which the speed is constant. After carrying out a methodical study of the corrosion of 'zircaloy-2 in the form of sheets and tubes. We have demonstrated, at 400 deg. C in steam, a systematic anomaly which appears at the transition point. The curve presents three quite distinct points; after the first period a fast corrosion is observed, followed by a third period at a slower speed. This leads us to believe that there may be not a single point but a transition zone, separating two types of kinetic behaviour and corresponding to modifications in the properties of the oxide layer. After this readjustment period a new corrosion law is established, lasting a considerable time, the corrosion speed being slower than that indicated so far. A study of the morphology of the oxide films which develop under these conditions has demonstrated the special part played by mechanical, physical and metallurgical factors in the case of zirconium. Deep penetration of oxide can thus show up on the inner wall of hammer-hardened tubes. Simultaneously a very considerable hydride formation occurs in the metal. (author) [French] La cinetique de corrosion du zircaloy-2 dans l'eau et la vapeur a des temperatures comprises entre 300 et 400 deg. C est representee par une courbe a deux periodes separees par un point singulier appele point de transition. A une premiere periode a vitesse de corrosion decroissante, succede une deuxieme periode a cinetique beaucoup plus rapide dont la vitesse est constante. Apres une etude systematique de la corrosion du zircaloy-2 sous forme de toles et de tubes, nous avons mis en evidence a 400 deg. C, dans la vapeur, une anomalie systematique qui se

  19. Oxidation behavior of fuel cladding tube in spent fuel pool accident condition

    International Nuclear Information System (INIS)

    Nemoto, Yoshiyuki; Kaji, Yoshiyuki; Ogawa, Chihiro; Nakashima, Kazuo; Tojo, Masayuki

    2017-01-01

    In spent fuel pool (SFP) under loss-of-cooling or loss-of-coolant severe accident condition, the spent fuels will be exposed to air and heated by their own residual decay heat. Integrity of fuel cladding is crucial for SFP safety therefore study on cladding oxidation in air at high temperature is important. Zircaloy-2 (Zry2) and zircaloy-4 (Zry4) were applied for thermogravimetric analyses (TGA) in different temperatures in air at different flow rates to evaluate oxidation behavior. Oxidation rate increased with testing temperature. In a range of flow rate of air which is predictable in spent fuel lack during a hypothetical SFP accident, influence of flow rate was not clearly observed below 950degC for the Zry2, or below 1050degC for Zry4. In higher temperature, oxidation rate was higher in high rate condition, and this trend was seen clearer when temperature increased. Oxide layers were carefully examined after the TGA analyses and compared with mass gain data to investigate detail of oxidation process in air. It was revealed that the mass gain data in pre-breakaway regime reflects growth of dense oxide film on specimen surface, meanwhile in post-breakaway regime, it reflects growth of porous oxide layer beneath fracture of the dense oxide film. (author)

  20. Development of uranium metal targets for 99Mo production

    International Nuclear Information System (INIS)

    Wiencek, T.C.; Hofman, G.L.

    1993-10-01

    A substantial amount of high enriched uranium (HEU) is used for the production of medical-grade 99 Mo. Promising methods of producing irradiation targets are being developed and may lead to the reduction or elimination of this HEU use. To substitute low enriched uranium (LEU) for HEU in the production of 99 Mo, the target material may be changed to uranium metal foil. Methods of fabrication are being developed to simplify assembly and disassembly of the targets. Removal of the uranium foil after irradiation without dissolution of the cladding is a primary goal in order to reduce the amount of liquid radioactive waste material produced in the process. Proof-of-concept targets have been fabricated. Destructive testing indicates that acceptable contact between the uranium foil and the cladding can be achieved. Thermal annealing tests, which simulate the cladding/uranium diffusion conditions during irradiation, are underway. Plans are being made to irradiate test targets

  1. Scientific basis for storage criteria for interim dry storage of aluminum-clad fuels

    International Nuclear Information System (INIS)

    Sindelar, R.L.; Peacock, H.B. Jr.; Lam, P.S.; Iyer, N.C.; Louthan, M.R. Jr.; Murphy, J.R.

    1996-01-01

    An engineered system for dry storage of aluminum-clad foreign and domestic research reactor spent fuel owned by the US Department of Energy is being considered to store the fuel up to a nominal period of 40 years prior to ultimate disposition. Scientifically-based criteria for environmental limits to drying and storing the fuels for this system are being developed to avoid excessive degradation in sealed and non-sealed (open to air) dry storage systems. These limits are based on consideration of degradation modes that can cause loss of net section of the cladding, embrittlement of the cladding, distortion of the fuel, or release of fuel and fission products from the fuel/clad system. Potential degradation mechanisms include corrosion mechanisms from exposure to air and/or sources of humidity, hydrogen blistering of the aluminum cladding, distortion of the fuel due to creep, and interdiffusion of the fuel and fission products with the cladding. The aluminum-clad research reactor fuels are predominantly highly-enriched aluminum uranium alloy fuel which is clad with aluminum alloys similar to 1100, 5052, and 6061 aluminum. In the absence of corrodant species, degradation due to creep and diffusion mechanisms limit the maximum fuel storage temperature to 200 C. The results of laboratory scale corrosion tests indicate that this fuel could be stored under air up to 200 C at low relative humidity levels (< 20%) to limit corrosion of the cladding and fuel (exposed to the storage environment through assumed pre-existing pits in the cladding). Excessive degradation of fuels with uranium metal up to 200 C can be avoided if the fuel is sufficiently dried and contained in a sealed system; open storage can be achieved if the temperature is controlled to avoid excessive corrosion even in dry air

  2. Performance of iron–chromium–aluminum alloy surface coatings on Zircaloy 2 under high-temperature steam and normal BWR operating conditions

    Energy Technology Data Exchange (ETDEWEB)

    Zhong, Weicheng; Mouche, Peter A.; Han, Xiaochun [University of Illinois, Department of Nuclear, Radiological, and Plasma Engineering, Urbana, IL 61801 (United States); Heuser, Brent J., E-mail: bheuser@illinois.edu [University of Illinois, Department of Nuclear, Radiological, and Plasma Engineering, Urbana, IL 61801 (United States); Mandapaka, Kiran K.; Was, Gary S. [University of Michigan, Department of Nuclear Engineering and Radiological Sciences, Ann Arbor, MI 48109 (United States)

    2016-03-15

    Iron-chromium-aluminum (FeCrAl) coatings deposited on Zircaloy 2 (Zy2) and yttria-stabilized zirconia (YSZ) by magnetron sputtering have been tested with respect to oxidation weight gain in high-temperature steam. In addition, autoclave testing of FeCrAl-coated Zy2 coupons under pressure-temperature-dissolved oxygen coolant conditions representative of a boiling water reactor (BWR) environment has been performed. Four different FeCrAl compositions have been tested in 700 °C steam; compositions that promote alumina formation inhibited oxidation of the underlying Zy2. Parabolic growth kinetics of alumina on FeCrAl-coated Zy2 is quantified via elemental depth profiling. Autoclave testing under normal BWR operating conditions (288 °C, 9.5 MPa with normal water chemistry) up to 20 days demonstrates observable weight gain over uncoated Zy2 simultaneously exposed to the same environment. However, no FeCrAl film degradation was observed. The 900 °C eutectic in binary Fe–Zr is addressed with the FeCrAl-YSZ system. - Graphical abstract: Weight gain normalized to total sample surface area versus time during 700 °C steam exposure for FeCrAl samples with different composition (A) and Fe/Cr/Al:62/4/34 (B). In both cases, the responses of uncoated Zry2 (Zry2-13A and Zry2-19A) are shown for comparison. This uncoated Zry2 response shows the expected pre-transition quasi-cubic kinetic behavior and eventual breakaway (linear) kinetics. Highlights: • FeCrAl coatings deposited on Zy2 have been tested with respect to oxidation in high-temperature steam. • FeCrAl compositions promoting alumina formation inhibited oxidation of Zy2 and delay weight gain. • Autoclave testing to 20 days of coated Zy2 in a simulated BWR environment demonstrates minimal weight gain and no film degradation. • The 900 °C eutectic in binary Fe-Zr is addressed with the FeCrAl-YSZ system.

  3. Effect of chemical composition on corrosion resistance of Zircaloy fuel cladding tube for BWR

    International Nuclear Information System (INIS)

    Inagaki, Masahisa; Akahori, Kimihiko; Kuniya, Jirou; Masaoka, Isao; Suwa, Masateru; Maru, Akira; Yasuda, Teturou; Maki, Hideo.

    1990-01-01

    Effects of Fe and Ni contents on nodular corrosion susceptibility and hydrogen pick-up of Zircaloy were investigated. Total number of 31 Zr alloys having different chemical compositions; five Zr-Sn-Fe-Cr alloys, eight Zr-Sn-Fe-Ni alloys and eighteen Zr-Sn-Fe-Ni-Cr alloys, were melted and processed to thin plates for the corrosion tests in the environments of a high temperature (510degC) steam and a high temperature (288degC) water. In addition, four 450 kg ingots of Zr-Sn-Fe-Ni-Cr alloys were industrially melted and BWR fuel cladding tubes were manufactured through a current material processing sequence to study their producibility, tensile properties and corrosion resistance. Nodular corrosion susceptibility decreased with increasing Fe and Ni contents of Zircaloys. It was seen that the improved Zircaloys having Fe and Ni contents in the range of 0.30 [Ni]+0.15[Fe]≥0.045 (w%) showed no susceptibility to nodular corrosion. An increase of Fe content resulted in a decrease of hydrogen pick-up fraction in both steam and water environments. An increase of Fe and Ni content of Zircaloys in the range of Fe≤0.25 w% and Ni≤0.1 w% did not cause the changes in tensile properties and fabricabilities of fuel cladding tube. The fuel cladding tube of improved Zircaloy, containing more amount of Fe and Ni than the upper limit of Zircaloy-2 specification showed no susceptibility to nodular corrosion even in the 530degC steam test. (author)

  4. Cladding tube manufacturing technology

    International Nuclear Information System (INIS)

    Hahn, R.; Jeong, Y. H.; Baek, B. J.; Kim, K. H.; Kim, S. J.; Choi, B. K.; Kim, J. M.

    1999-04-01

    This report gives an overview of the manufacturing routine of PWR cladding tubes. The routine essentially consists of a series of deformation and annealing processes which are necessary to transform the ingot geometry to tube dimensions. By changing shape, microstructure and structure-related properties are altered simultaneously. First, a short overview of the basics of that part of deformation geometry is given which is related to tube reducing operations. Then those processes of the manufacturing routine which change the microstructure are depicted, and the influence of certain process parameters on microstructure and material properties are shown. The influence of the resulting microstructure on material properties is not discussed in detail, since it is described in my previous report A lloy Development for High Burnup Cladding . Because of their paramount importance still up to now, and because manufacturing data and their influence on properties for other alloys are not so well established or published, the descriptions are mostly related to Zry4 tube manufacturing, and are only in short for other alloys. (author). 9 refs., 46 figs

  5. Cladding tube manufacturing technology

    Energy Technology Data Exchange (ETDEWEB)

    Hahn, R. [Kraftwerk Union AG, Mulheim (Germany); Jeong, Y.H.; Baek, B.J.; Kim, K.H.; Kim, S.J.; Choi, B.K.; Kim, J.M. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1999-04-01

    This report gives an overview of the manufacturing routine of PWR cladding tubes. The routine essentially consists of a series of deformation and annealing processes which are necessary to transform the ingot geometry to tube dimensions. By changing shape, microstructure and structure-related properties are altered simultaneously. First, a short overview of the basics of that part of deformation geometry is given which is related to tube reducing operations. Then those processes of the manufacturing routine which change the microstructure are depicted, and the influence of certain process parameters on microstructure and material properties are shown. The influence of the resulting microstructure on material properties is not discussed in detail, since it is described in my previous report 'Alloy Development for High Burnup Cladding.' Because of their paramount importance still up to now, and because manufacturing data and their influence on properties for other alloys are not so well established or published, the descriptions are mostly related to Zry4 tube manufacturing, and are only in short for other alloys. (author). 9 refs., 46 figs.

  6. Studies of Corrosion of Cladding Materials in Simulated BWR-environment Using Impedance Measurements. Part I: Measurements in the Pre-transition Region

    International Nuclear Information System (INIS)

    Forsberg, Stefan; Ahlberg, Elisabet; Andersson, Ulf

    2004-09-01

    The corrosion of three Zircaloy 2 cladding materials, LK2, LK2+ and LK3, have been studied in-situ in an autoclave using electrochemical impedance spectroscopy. Measurements were performed in simulated BWR water at temperatures up to 288 deg C. The impedance spectra were successfully modelled using equivalent circuits. When the oxide grew thicker during the experiments, a change-over from one to two time constants was seen, showing that a layered structure was formed. Oxide thickness, oxide conductivity and effective donor density were evaluated from the impedance data. The calculated oxide thickness at the end of the experiments was consistent with the value obtained from SEM. It was shown that the difference in oxide growth rate between the investigated materials is small in the pre-transition region. The effective donor density, which is a measure of electronic conductivity, was found to be lower for the LK3 material compared to the other two materials

  7. CASTI handbook of cladding technology. 2. ed.

    International Nuclear Information System (INIS)

    Smith, L.; Celant, M.

    2000-01-01

    This updated (2000) CASTI handbook covers all aspects of clad products - the different means of manufacture, properties and applications in various industries. Topics include: an introduction to cladding technology, clad plate, clad pipes, bends, clad fittings, specification requirements of clad products, welding clad products, clad product application and case histories from around the world. Unique to this book is the documentation of case histories of major cladding projects from around the world and how the technology of that day has withstood the demands of time. Filled with over 100 photos and graphics illustrating the various cladding technology examples and products, this book truly documents the most recent technologies in the field of cladding technology used worldwide

  8. Compatibility Behavior of the Ferritic-Martensitic Steel Cladding under the Liquid Sodium Environment

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jun Hwan; Baek, Jong Hyuk; Kim, Sung Ho; Lee, Chan Bock [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-05-15

    Fuel cladding is a component which confines uranium fuel to transport energy into the coolant as well as protect radioactive species from releasing outside. Sodium-cooled Fast Reactor (SFR) has been considered as one of the most probable next generation reactors in Korea because it can maximize uranium resource as well as reduce the amount of PWR spent fuel in conjunction with pyroprocessing. Sodium has been selected as the coolant of the SFR because of its superior fast neutron efficiency as well as thermal conductivity, which enables high power core design. However, it is reported that the fuel cladding materials like austenitic and ferritic stainless steel react sodium coolant so that the loss of the thickness, intergranular attack, and carburization or decarburization process may happen to induce the change of the mechanical property of the cladding. This study aimed to evaluate material property of the cladding material under the liquid sodium environment. Ferritic-martensitic steel (FMS) coupon and cladding tube were exposed at the flowing sodium then the microstructural and mechanical property were evaluated. mechanical property of the cladding was evaluated using the ring tension test

  9. Stone cladding engineering

    CERN Document Server

    Sousa Camposinhos, Rui de

    2014-01-01

    This volume presents new methodologies for the design of dimension stone based on the concepts of structural design while preserving the excellence of stonemasonry practice in façade engineering. Straightforward formulae are provided for computing action on cladding, with special emphasis on the effect of seismic forces, including an extensive general methodology applied to non-structural elements. Based on the Load and Resistance Factor Design Format (LRDF), minimum slab thickness formulae are presented that take into consideration stress concentrations analysis based on the Finite Element Method (FEM) for the most commonly used modern anchorage systems. Calculation examples allow designers to solve several anchorage engineering problems in a detailed and objective manner, underlining the key parameters. The design of the anchorage metal parts, either in stainless steel or aluminum, is also presented.

  10. Study of the uranium-zirconium diffusion; Etude de la diffusion uranium-zirconium

    Energy Technology Data Exchange (ETDEWEB)

    Adda, Y; Mairy, C; Bouchet, P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1957-07-01

    The intermetallic diffusion of uranium fuel and zirconium used as cladding is studied. Intermetallic diffusion can occur during the cladding of uranium rods and uranium can penetrate the zirconium cladding. Different parameters are involved in this mechanism as structure and mechanical properties of the diffusion area as well as presence of impurities in the metal. The uses of different analysis techniques (micrography, Castaing electronic microprobe, microhardness and autoradiography) have permitted to determine with great accuracy the diffusion coefficient in gamma phase (body centered cubic system) and the results have given important information on the intermetallic diffusion mechanisms. The existence of the Kirkendall effect in the U-Zr diffusion is also an argument in favor of the generality of the diffusion mechanism by vacancies in body centered cubic system. (M.P.)

  11. Performance testing of refractory alloy-clad fuel elements for space reactors

    International Nuclear Information System (INIS)

    Dutt, D.S.; Cox, C.M.; Karnesky, R.A.; Millhollen, M.K.

    1985-01-01

    Two fast reactor irradiation tests, SP-1 and SP-2, provide a unique and self-consistent data set with which to evaluate the technical feasibility of potential fuel systems for the SP-100 space reactor. Fuel pins fabricated with leading cladding candidates (Nb-1Zr, PWC-11, and Mo-13Re) and fuel forms (UN and UO 2 ) are operated at temperatures typical of those expected in the SP-100 design. The first US fast reactor irradiated, refractory alloy clad fuel pins, from the SP-1 test, reached 1 at. % burnup in EBR-II in March 1985. At that time selected pins were discharged for interim examination. These examinations confirmed the excellent performance of the Nb-1Zr clad uranium oxide and uranium nitride fuel elements, which are the baseline fuel systems for two SP-100 reactor concepts

  12. UK experience on fuel and cladding interaction in oxide fuels

    Energy Technology Data Exchange (ETDEWEB)

    Batey, W [Dounreay Experimental Reactor Establishment, Thurso, Caithness (United Kingdom); Findlay, J R [AERE, Harwell, Didcot, Oxon (United Kingdom)

    1977-04-01

    The occurrence of fuel cladding interactions in fast reactor fuels has been observed in UK irradiations over a period of years. Chemical incompatibility between fuel and clad represents a potential source of failure and has, on this account, been studied using a variety of techniques. The principal fuel of interest to the UK for fast reactor application is mixed uranium plutonium oxide clad in stainless steel and it is in this field that the majority of work has been concentrated. Some consideration has been given to carbide fuels, because of their application as an advanced fuel. This experience is described in the accompanying paper. Several complementary initiatives have been followed to investigate the interactions in oxide fuel. The principal source of experimental information is from the experimental fuel irradiation programme in the Dounreay Fast Reactor (DFR). Supporting information has been obtained from irradiation programmes in Materials Testing Reactors (MTR). Conditions approaching those in a fast reactor are obtained and the effects of specific variables have been examined in specifically designed experiments. Out-of-reactor experiments have been used to determine the limits of fuel and cladding compatibility and also to give indications of corrosion The observations from all experiments have been examined in the light of thermo-dynamic predictions of fuel behaviour to assess the relative significance of various observations and operating conditions. An experimental programme to control and limit the interactions in oxide fuel is being followed.

  13. UK experience on fuel and cladding interaction in oxide fuels

    International Nuclear Information System (INIS)

    Batey, W.; Findlay, J.R.

    1977-01-01

    The occurrence of fuel cladding interactions in fast reactor fuels has been observed in UK irradiations over a period of years. Chemical incompatibility between fuel and clad represents a potential source of failure and has, on this account, been studied using a variety of techniques. The principal fuel of interest to the UK for fast reactor application is mixed uranium plutonium oxide clad in stainless steel and it is in this field that the majority of work has been concentrated. Some consideration has been given to carbide fuels, because of their application as an advanced fuel. This experience is described in the accompanying paper. Several complementary initiatives have been followed to investigate the interactions in oxide fuel. The principal source of experimental information is from the experimental fuel irradiation programme in the Dounreay Fast Reactor (DFR). Supporting information has been obtained from irradiation programmes in Materials Testing Reactors (MTR). Conditions approaching those in a fast reactor are obtained and the effects of specific variables have been examined in specifically designed experiments. Out-of-reactor experiments have been used to determine the limits of fuel and cladding compatibility and also to give indications of corrosion The observations from all experiments have been examined in the light of thermo-dynamic predictions of fuel behaviour to assess the relative significance of various observations and operating conditions. An experimental programme to control and limit the interactions in oxide fuel is being followed

  14. Flat cladding and pellets in the design of an irradiation target

    International Nuclear Information System (INIS)

    Yorio, Daniel; Denis, Alicia C.; Soba, Alejandro; Beuter, Oscar; Marajofsky, Adolfo

    2003-01-01

    The design of an enriched uranium irradiation target made of flat pellets and cladding is proposed in order to improve the fission Mo 99 production. The variation range of each one of the parameters is studied and the basic design of the target is given

  15. The influence of cladding on fission gas release from irradiated U-Mo monolithic fuel

    Energy Technology Data Exchange (ETDEWEB)

    Burkes, Douglas E., E-mail: Douglas.Burkes@pnnl.gov; Casella, Amanda J.; Casella, Andrew M.

    2017-04-01

    The monolithic uranium-molybdenum (U-Mo) alloy has been proposed as a fuel design capable of converting the world's highest power research reactors from use of high enriched uranium to low enriched uranium. However, a zirconium (Zr) diffusion barrier must be used to eliminate interactions that form between the U-Mo monolith and aluminum alloy 6061 (AA6061) cladding during fabrication and are enhanced during irradiation. One aspect of fuel development and qualification is to demonstrate an appropriate understanding of the extent of fission product release from the fuel under anticipated service environments. An exothermic reaction has previously been observed between the AA6061 cladding and Zr diffusion layer. In this paper, two fuel segments with different irradiation history were subjected to specified thermal profiles under a controlled atmosphere using a thermogravimetric/differential thermal analyzer coupled with a mass spectrometer inside a hot cell. Samples from each segment were tested with cladding and without cladding to investigate the effect, if any, that the exothermic reaction has on fission gas release mechanisms. Measurements revealed there is an instantaneous effect of the cladding/Zr exothermic reaction, but not necessarily a cumulative effect above approximately 973 K (700 °C). The mechanisms responsible for fission gas release events are discussed. - Highlights: •Complementary fission gas release events are reported for U-Mo fuel with and without cladding. •Exothermic reaction between Zr diffusion layer and cladding influences fission gas release. •Mechanisms responsible for fission gas release are similar, but with varying timing and magnitude. •Behavior of samples is similar after 800 °C signaling the onset of superlattice destabilization.

  16. Application of the beta particles backscattering technique for determining the thickness of the cladding in nuclear fuels plate

    International Nuclear Information System (INIS)

    Koshimizu, S.; Ferreira, P.I.; Lima, L.F.C.P. de; Vieira, J.M.; Perez, H.E.B.

    1984-01-01

    A prototype of an instalation to measure thickness of cladding and core of nuclear fuels plate using the beta particles backscattering technique is constructed. The method and calibration system is described. The thickness measurements of the cladding and core were done in a natural uranium fuel plate developed at IPEN. The reliability of the method is confirmed by the metalographic measures analysis. (E.G.) [pt

  17. The elastic properties of zirconium alloy fuel cladding and pressure tubing materials

    International Nuclear Information System (INIS)

    Rosinger, H.E.; Northwood, D.O.

    1979-01-01

    A knowledge of the elastic properties of zirconium alloys is required in the mathematical modelling of cladding and pressure tubing performance. Until recently, little of this type of data was available, particularly at elevated temperatures. The dynamic elastic moduli of zircaloy-2, zircaloy-4, the alloys Zr-1.0 wt%Nb, Zr-2.5 wt%Nb and Marz grade zirconium have therefore been determined over the temperature range 275 to 1000 K. Young's modulus and shear modulus for all the zirconium alloys decrease with temperature and are expressed by empirical relations fitted to the data. The elastic properties are texture dependent and a detailed study has been conducted on the effect of texture on the elastic properties of Zr-1.0 wt% Nb over the temperature range 275 to 775 K. The results are compared with polycrystalline elastic constants computed from single crystal elastic constants, and the effect of texture on the dynamic elastic moduli is discussed in detail. (Auth.)

  18. Zirconium-barrier cladding attributes

    International Nuclear Information System (INIS)

    Rosenbaum, H.S.; Rand, R.A.; Tucker, R.P.; Cheng, B.; Adamson, R.B.; Davies, J.H.; Armijo, J.S.; Wisner, S.B.

    1987-01-01

    This metallurgical study of Zr-barrier fuel cladding evaluates the importance of three salient attributes: (1) metallurgical bond between the zirconium liner and the Zircaloy substrate, (2) liner thickness (roughly 10% of the total cladding wall), and (3) softness (purity). The effect that each of these attributes has on the pellet-cladding interaction (PCI) resistance of the Zr-barrier fuel was studied by a combination of analytical model calculations and laboratory experiments using an expanding mandrel technique. Each of the attributes is shown to contribute to PCI resistance. The effect of the zirconium liner on fuel behavior during off-normal events in which steam comes in contact with the zirconium surface was studied experimentally. Simulations of loss-of-coolant accident (LOCA) showed that the behavior of Zr-barrier cladding is virtually indistinguishable from that of conventional Zircaloy cladding. If steam contacts the zirconium liner surface through a cladding perforation and the fuel rod is operated under normal power conditions, the zirconium liner is oxidized more rapidly than is Zircaloy, but the oxidation rate returns to the rate of Zircaloy oxidation when the oxide phase reaches the zirconium-Zircaloy metallurgical bond

  19. Fuel-pin cladding transient failure strain criterion

    International Nuclear Information System (INIS)

    Bard, F.E.; Duncan, D.R.; Hunter, C.W.

    1983-01-01

    A criterion for cladding failure based on accumulated strain was developed for mixed uranium-plutonium oxide fuel pins and used to interpret the calculated strain results from failed transient fuel pin experiments conducted in the Transient Reactor Test (TREAT) facility. The new STRAIN criterion replaced a stress-based criterion that depends on the DORN parameter and that incorrectly predicted fuel pin failure for transient tested fuel pins. This paper describes the STRAIN criterion and compares its prediction with those of the stress-based criterion

  20. Carbon 14 distribution in irradiated BWR fuel cladding and released carbon 14 after aqueous immersion of 6.5 years

    Energy Technology Data Exchange (ETDEWEB)

    Sakuragi, T. [Radioactive Waste Management Funding and Research Center, Tsukishima 1-15-7, Chuo City, Tokyo, 104-0052 (Japan); Yamashita, Y.; Akagi, M.; Takahashi, R. [TOSHIBA Corporation, Ukishima Cho 4-1, Kawasaki Ward, Kawasaki, 210-0862 (Japan)

    2016-07-01

    Spent fuel cladding which is highly activated and strongly contaminated is expected to be disposed of in an underground repository. A typical activation product in the activated metal waste is carbon 14 ({sup 14}C), which is mainly generated by the {sup 14}N(n,p){sup 14}C reaction and produces a significant exposure dose due to the large inventory, long half-life (5730 years), rapid release rate, and the speciation and consequent migration parameters. In the preliminary Japanese safety case, the release of radionuclides from the metal matrix is regarded as the corrosion-related congruent release, and the cladding oxide layer is regarded as a source of instant release fraction (IRF). In the present work, specific activity of {sup 14}C was measured using an irradiated BWR fuel cladding (Zircaloy-2, average rod burnup of 41.6 GWd/tU) which has an external oxide film having a thickness of 25.3 μm. The {sup 14}C specific activity of the base metal was 1.49*10{sup 4} Bq/g, which in the corresponding burnup is comparable to values in the existing literature, which were obtained from various irradiated claddings. Although the specific activity in oxide was 2.8 times the base metal activity due to the additive generation by the {sup 17}O(n,α){sup 14}C reaction, the {sup 14}C abundance in oxide was less than 10% of total inventory. A static leaching test using the cladding tube was carried out in an air-tight vessel filled with a deoxygenated dilute NaOH solution (pH of 12.5) at room temperature. After 6.5 years, {sup 14}C was found in each leachate fraction of gas phase and dissolved organics and inorganics, the total of which was less than 0.01% of the {sup 14}C inventory of the immersed cladding tube. A simple calculation based on the congruent release with Zircaloy corrosion has suggested that the 96.7% of released {sup 14}C was from the external oxide layer and 3.3% was from the base Zircaloy metal. However, both the {sup 14}C abundance and the low leaching rate

  1. The fuel-cladding interfacial friction coefficient in water-cooled reactor fuel rods

    International Nuclear Information System (INIS)

    Smith, E.

    1979-01-01

    A central problem in the development of cladding failure criteria and of effective operational, design or material remedies is to know whether the cladding stress is enhanced significantly near cladding ridges, pellet chips or fuel pellet cracks; the latter may also be coincident with cladding ridges at pellet-pellet interfaces. As regards the fuel pellet crack source of cladding stress concentration, the magnitude of the uranium dioxide-Zircaloy interfacial friction coefficient μ governs the magnitude and distribution of the enhanced cladding stress. Considerable discussion, particularly at a Post-Conference Seminar associated with the SMIRT 4 Conference, has focussed on the value of μ, the author taking the view that it is unlikely to be large (< 0.5). The reasoning behind this view is as follows. A fuel pellet should fracture during a power ramp when the tensile hoop stress within the pellet exceeds the fuel's fracture stress. Since the preferred position for a fuel pellet crack to form is at the fuel-cladding interface midway between existing fuel cracks, where the interfacial shear stress changes sign, the pellet segment size after a power ramp provides a limit to the magnitude of the interfacial shear stresses and consequently to the value of μ. With this argument as a basis, the author's early work used the Gittus fuel rod model, in which there is a symmetric distribution of fuel pellet cracks and symmetric interfacial slippage, to show that μ < 0.5 if it is assumed that the average hoop stress within the cladding attains yield levels. It was therefore suggested that a high interfacial friction coefficient is unlikely to be operative during a power ramp; this result was used to support the view that interfacial friction effects do not play a dominant role in stress corrosion crack formation within the cladding. (orig.)

  2. Development of Diffusion barrier coatings and Deposition Technologies for Mitigating Fuel Cladding Chemical Interactions (FCCI)

    Energy Technology Data Exchange (ETDEWEB)

    Sridharan, Kumar; Allen, Todd; Cole, James

    2013-02-27

    The goal of this project is to develop diffusion barrier coatings on the inner cladding surface to mitigate fuel-cladding chemical interaction (FCCI). FCCI occurs due to thermal and radiation enhanced inter-diffusion between the cladding and fuel materials, and can have the detrimental effects of reducing the effective cladding wall thickness and lowering the melting points of the fuel and cladding. The research is aimed at the Advanced Burner Reactor (ABR), a sodium-cooled fast reactor, in which higher burn-ups will exacerbate the FCCI problem. This project will study both diffusion barrier coating materials and deposition technologies. Researchers will investigate pure vanadium, zirconium, and titanium metals, along with their respective oxides, on substrates of HT-9, T91, and oxide dispersion-strengthened (ODS) steels; these materials are leading candidates for ABR fuel cladding. To test the efficacy of the coating materials, the research team will perform high-temperature diffusion couple studies using both a prototypic metallic uranium fuel and a surrogate the rare-earth element lanthanum. Ion irradiation experiments will test the stability of the coating and the coating-cladding interface. A critical technological challenge is the ability to deposit uniform coatings on the inner surface of cladding. The team will develop a promising non-line-of-sight approach that uses nanofluids . Recent research has shown the feasibility of this simple yet novel approach to deposit coatings on test flats and inside small sections of claddings. Two approaches will be investigated: 1) modified electrophoretic deposition (MEPD) and 2) boiling nanofluids. The coatings will be evaluated in the as-deposited condition and after sintering.

  3. Development of a recovery process of scraps resulting from the manufacture of metallic uranium fuels

    International Nuclear Information System (INIS)

    Camilo, Ruth L.; Kuada, Terezinha A.; Forbicini, Christina A.L.G.O.; Cohen, Victor H.; Araujo, Bertha F.; Lobao, Afonso S.T.

    1996-01-01

    The study of the dissolution of natural metallic uranium fuel samples with aluminium cladding is presented, in order to obtain optimized conditions for the system. The aluminium cladding was dissolved in an alkaline solution of Na OH/Na NO 3 and the metallic uranium with HNO 3 . A fumeless dissolution with total recovery of nitrous gases was achieved. The main purpose of this project was the recovery of uranium from scraps resulting from the manufacture of the metallic uranium fuel or other non specified fuels. (author)

  4. Depleted uranium

    International Nuclear Information System (INIS)

    Huffer, E.; Nifenecker, H.

    2001-02-01

    This document deals with the physical, chemical and radiological properties of the depleted uranium. What is the depleted uranium? Why do the military use depleted uranium and what are the risk for the health? (A.L.B.)

  5. Study on characteristics of spent PWR cladding hull for categorizing into Non-TRU waste

    International Nuclear Information System (INIS)

    Jung, In Ha; Kim, Jong Ho; Park, Jang Jin; Shin, Jin Myeong; Lee, Ho Hee; Yang, Myung Seung

    2005-01-01

    AFCI and GEN-IV programs aim for decreasing the high level radioactive wastes to be disposed. They also try to get valuable materials to recycle as resources such as uranium and plutonium. On the other hand, cladding hull expected to be one-thirds in volume of spent fuel assembly has not studied so much in the point view of recycling to reuse. Since traditional process of reprocessing was wet process, cladding hull generating through the reprocessing process was unavoidably contaminated with TRU by acid solvent during the process. Therefore, cladding hull has been classified into TRU wastes or high level wastes. According to the strategy for TRU high level radioactive wastes of USA as well as Korea, it regulates in two respects. One is activity and the other is heat generation. In respect of activity, TRU waste contains more than 100 nCi/kg of alpha emits with longer half life than 20 years and higher than 92 in atomic number. Also, wastes are categorized into TRU waste when it generates higher than 2kW/m3, in the respect of heat generation. Our results as well as literatures, almost all of TRU nuclides in the cladding hull are responsible for remained uranium and plutonium owing to pellet-cladding interaction. In addition, recoiled fission products on the surface of the cladding hull serve as heat generator. Up to now, decontamination of the cladding hull generating from the reprocessing of wet process is regarded as valueless and un-economic works owing to the amount of second waste produced

  6. Clad Degradation - FEPs Screening Arguments

    International Nuclear Information System (INIS)

    E. Siegmann

    2004-01-01

    The purpose of this report is to document the screening of the cladding degradation features, events, and processes (FEPs) for commercial spent nuclear fuel (CSNF). This report also addresses the effect of some FEPs on both the cladding and the CSNF, DSNF, and HLW waste forms where it was considered appropriate to address the effects on both materials together. This report summarizes the work of others to screen clad degradation FEPs in a manner consistent with, and used in, the Total System Performance Assessment-License Application (TSPA-LA). This document was prepared according to ''Technical Work Plan for Waste Form Degradation Modeling, Testing, and Analyses in Support of LA'' (BSC 2004a [DIRS 167796])

  7. Pin clad strains in Phenix

    International Nuclear Information System (INIS)

    Languille, A.

    1979-07-01

    The Phenix reactor has operated for 4 years in a satisfactory manner. The first 2 sub-assembly loadings contained pins clad in solution treated 316. The principal pin strains are: diametral strain (swelling and irradiation creep), ovality and spiral bending of the pin (interaction of wire and pin cluster and wrapper). A pin cluster irradiated to a dose of 80 dpa F reached a pin diameter strain of 5%. This strain is principally due to swelling (low fission gas pressure). The principal parameters governing the swelling are instantaneous dose, time and temperature for a given type of pin cladding. Other types of steel are or will be irradiated in Phenix. In particular, cold-worked titanium stabilised 316 steel should contribute towards a reduction in the pin clad strains and increase the target burn-up in this reactor. (author)

  8. Fuel-cladding chemical interaction

    International Nuclear Information System (INIS)

    Gueneau, C.; Piron, J.P.; Dumas, J.C.; Bouineau, V.; Iglesias, F.C.; Lewis, B.J.

    2015-01-01

    The chemistry of the nuclear fuel is very complex. Its chemical composition changes with time due to the formation of fission products and depends on the temperature level history within the fuel pellet and the clad during operation. Firstly, in thermal reactors, zircaloy oxidation from reaction with UO 2 fuel under high-temperature conditions will be addressed. Then other fuel-cladding interaction phenomena occurring in fast reactors will be described. Large thermal gradients existing between the centre and the periphery of the pellet induce the radial redistribution of the fuel constituents. The fuel pellet can react with the clad by different corrosion processes which can involve actinide and/or fission product transport via gas, liquid or/and solid phases. All these phenomena are briefly described in the case of different kinds of fuels (oxide, carbide, nitride, metallic) to be used in fast reactors. The way these phenomena are taken into account in fuel performance codes is presented. (authors)

  9. Annex 7 - Task 08/32 final report, Diffusion bond between the fuel and cladding of the RA reactor fuel element; Prilog 7 - Zavrsni izvestaj o podzadatku 08/32, Difuziona veza goriva i kosuljice gorivnog elementa reaktora RA

    Energy Technology Data Exchange (ETDEWEB)

    Kustudic, D [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1961-12-15

    Based on testing the diffusin bond between the fuel and the aluminium cladding of the RA reactor fuel element it was found that the diffusion bond exists; uranium is protected by a nickel layer to prevent the damaging diffusion between uranium and aluminium; the nickel layer is about 5 microns thick. Under assumption that the diffusion annealing was done after cladding the annealing temperature was about 500 deg C for about 1 hour.

  10. Interaction between thorium and potential clad materials

    International Nuclear Information System (INIS)

    Kale, G.B.; Gawde, P.S.; Sengupta, Pranesh

    2005-01-01

    Thorium based fuels are being used for nuclear reactors. The structural stability of fuel-clad assemblies in reactor systems depend upon the nature of interdiffusion reaction between fuel-cladding materials. Interdiffusion reaction thorium and various cladding materials is presented in this paper. (author)

  11. ZIRCONIUM-CLADDING OF THORIUM

    Science.gov (United States)

    Beaver, R.J.

    1961-11-21

    A method of cladding thorium with zirconium is described. The quality of the bond achieved between thorium and zirconium by hot-rolling is improved by inserting and melting a thorium-zirconium alloy foil between the two materials prior to rolling. (AEC)

  12. Clad-coolant chemical interaction

    International Nuclear Information System (INIS)

    Iglesias, F.C.; Lewis, B.J.; Desgranges, C.; Toffolon, C.

    2015-01-01

    This paper provides an overview of the kinetics for zircaloy clad oxidation behaviour in steam and air during reactor accident conditions. The generation of chemical heat from metal/water reaction is considered. Low-temperature oxidation of zircaloy due to water-side corrosion is further described. (authors)

  13. Development of high performance cladding

    International Nuclear Information System (INIS)

    Kiuchi, Kiyoshi

    2003-01-01

    The developments of superior next-generation light water reactor are requested on the basis of general view points, such as improvement of safety, economics, reduction of radiation waste and effective utilization of plutonium, until 2030 year in which conventional reactor plants should be renovate. Improvements of stainless steel cladding for conventional high burn-up reactor to more than 100 GWd/t, developments of manufacturing technology for reduced moderation-light water reactor (RMWR) of breeding ratio beyond 1.0 and researches of water-materials interaction on super critical pressure-water cooled reactor are carried out in Japan Atomic Energy Research Institute. Stable austenite stainless steel has been selected for fuel element cladding of advanced boiling water reactor (ABWR). The austenite stain less has the superiority for anti-irradiation properties, corrosion resistance and mechanical strength. A hard spectrum of neutron energy up above 0.1 MeV takes place in core of the reduced moderation-light water reactor, as liquid metal-fast breeding reactor (LMFBR). High performance cladding for the RMWR fuel elements is required to get anti-irradiation properties, corrosion resistance and mechanical strength also. Slow strain rate test (SSRT) of SUS 304 and SUS 316 are carried out for studying stress corrosion cracking (SCC). Irradiation tests in LMFBR are intended to obtain irradiation data for damaged quantity of the cladding materials. (M. Suetake)

  14. Uranium conversion

    International Nuclear Information System (INIS)

    Oliver, Lena; Peterson, Jenny; Wilhelmsen, Katarina

    2006-03-01

    FOI, has performed a study on uranium conversion processes that are of importance in the production of different uranium compounds in the nuclear industry. The same conversion processes are of interest both when production of nuclear fuel and production of fissile material for nuclear weapons are considered. Countries that have nuclear weapons ambitions, with the intention to produce highly enriched uranium for weapons purposes, need some degree of uranium conversion capability depending on the uranium feed material available. This report describes the processes that are needed from uranium mining and milling to the different conversion processes for converting uranium ore concentrate to uranium hexafluoride. Uranium hexafluoride is the uranium compound used in most enrichment facilities. The processes needed to produce uranium dioxide for use in nuclear fuel and the processes needed to convert different uranium compounds to uranium metal - the form of uranium that is used in a nuclear weapon - are also presented. The production of uranium ore concentrate from uranium ore is included since uranium ore concentrate is the feed material required for a uranium conversion facility. Both the chemistry and principles or the different uranium conversion processes and the equipment needed in the processes are described. Since most of the equipment that is used in a uranium conversion facility is similar to that used in conventional chemical industry, it is difficult to determine if certain equipment is considered for uranium conversion or not. However, the chemical conversion processes where UF 6 and UF 4 are present require equipment that is made of corrosion resistant material

  15. Development of high performance cladding materials

    International Nuclear Information System (INIS)

    Park, Jeong Yong; Jeong, Y. H.; Park, S. Y.

    2010-04-01

    The irradiation test for HANA claddings conducted and a series of evaluation for next-HANA claddings as well as their in-pile and out-of pile performances tests were also carried out at Halden research reactor. The 6th irradiation test have been completed successfully in Halden research reactor. As a result, HANA claddings showed high performance, such as corrosion resistance increased by 40% compared to Zircaloy-4. The high performance of HANA claddings in Halden test has enabled lead test rod program as the first step of the commercialization of HANA claddings. DB has been established for thermal and LOCA-related properties. It was confirmed from the thermal shock test that the integrity of HANA claddings was maintained in more expanded region than the criteria regulated by NRC. The manufacturing process of strips was established in order to apply HANA alloys, which were originally developed for the claddings, to the spacer grids. 250 kinds of model alloys for the next-generation claddings were designed and manufactured over 4 times and used to select the preliminary candidate alloys for the next-generation claddings. The selected candidate alloys showed 50% better corrosion resistance and 20% improved high temperature oxidation resistance compared to the foreign advanced claddings. We established the manufacturing condition controlling the performance of the dual-cooled claddings by changing the reduction rate in the cold working steps

  16. Study of laser cladding nuclear valve parts

    International Nuclear Information System (INIS)

    Shi Shihong; Wang Xinlin; Huang Guodong

    1998-12-01

    The mechanism of laser cladding is discussed by using heat transfer model of laser cladding, heat conduction model of laser cladding and convective transfer mass model of laser melt-pool. Subsequently the laser cladding speed limit and the influence of laser cladding parameters on cladding layer structure is analyzed. A 5 kW with CO 2 transverse flow is used in the research for cladding treatment of sealing surface of stop valve parts of nuclear power stations. The laser cladding layer is found to be 3.0 mm thick. The cladding surface is smooth and has no such defects as crack, gas pore, etc. A series of comparisons with plasma spurt welding and arc bead welding has been performed. The results show that there are higher grain grade and hardness, lower dilution and better performances of resistance to abrasion, wear and of anti-erosion in the laser cladding layer. The new technology of laser cladding can obviously improve the quality of nuclear valve parts. Consequently it is possible to lengthen the service life of nuclear valve and to raise the safety and reliability of the production system

  17. Uranium exploration

    International Nuclear Information System (INIS)

    De Voto, R.H.

    1984-01-01

    This paper is a review of the methodology and technology that are currently being used in varying degrees in uranium exploration activities worldwide. Since uranium is ubiquitous and occurs in trace amounts (0.2 to 5 ppm) in virtually all rocks of the crust of the earth, exploration for uranium is essentially the search of geologic environments in which geologic processes have produced unusual concentrations of uranium. Since the level of concentration of uranium of economic interest is dependent on the present and future price of uranium, it is appropriate here to review briefly the economic realities of uranium-fueled power generation. (author)

  18. SIFAIL: a subprogram to calculate cladding deformation and damage for fast reactor fuel pins

    International Nuclear Information System (INIS)

    Wilson, D.R.; Dutt, D.S.

    1979-05-01

    SIFAIL is a series of subroutines used in conjunction with the thermal performance models of SIEX to assist in the evaluation of mechanical performance of mixed uranium plutonium oxide fuel pins. Cladding deformations due to swelling and creep are calculated. These have been compared to post-irradiation data from fuel pin tests in EBR-II. Several fuel pin cladding failure criteria (cumulative damage, total strain, and thermal creep strain) are evaluated to provide the fuel pin designer with a basis to select design parameters. SIFAIL allows the user many property options for cladding material. Code input is limited to geometric and environmental parameters, with a consistent set of material properties provided by the code. The simplified, yet adequate, thin wall stress--strain calculations provide a reliable estimate of fuel pin mechanical performance, while requiring a small amount of core storage and computer running time

  19. Pyrophoricity of uranium in long-term storage environments

    International Nuclear Information System (INIS)

    Solbrig, C.W.; Krsul, J.R.; Olsen, D.N.

    1994-01-01

    A corrosion cycle for uranium is postulated which can be used to assess whether a given storage situation might produce fire hazards and/or continual uranium corrosion. A significant reaction rate of uranium and moisture occurs at room temperature which produces uranium oxide and hydrogen. If the hydrogen cannot escape, it will react slowly with uranium to form uranium hydride. The hydride is pyrophoric at room temperature when exposed to air. Either the hydrogen or the hydride can produce a dangerous situation as demonstrated by two different incidents described here. Long-term corrosion will occur even if the normal precautions are taken as is demonstrated by the long-term storage of stainless steel clad uranium fuel plates. The major initiator of these problems is attributed to any moisture condensed on the metal or any brought in by the cover gas. The postulated corrosion cycle is used to suggest ways to circumvent these problems

  20. Laser cladding of quasicrystalline alloys

    International Nuclear Information System (INIS)

    Audebert, F.; Sirkin, H.; Colaco, R.; Vilar, R.

    1998-01-01

    Quasicrystals are a new class of ordinated structures with metastable characteristics room temperature. Quasicrystalline phases can be obtained by rapid quenching from the melt of some alloys. In general, quasicrystals present properties which make these alloys promising for wear and corrosion resistant coatings applications. During the last years, the development of quasicrystalline coatings by means of thermal spray techniques has been impulsed. However, no references have been found of their application by means of laser techniques. In this work four claddings of quasicrystalline compositions formed over aluminium substrate, produced by a continuous CO 2 laser using simultaneous powders mixture injection are presented. The claddings were characterized by X ray diffraction, scanning electron microscopy and Vickers microhardness. (Author) 18 refs

  1. Assessment of Neutronic Characteristics of Accident-Tolerant Fuel and Claddings for CANDU Reactors

    Directory of Open Access Journals (Sweden)

    Simon Younan

    2018-01-01

    Full Text Available The objective of this study was to evaluate accident-tolerant fuel (ATF concepts being considered for CANDU reactors. Several concepts, including uranium dioxide/silicon carbide (UO2-SiC composite fuel, dense fuels, microencapsulated fuels, and ATF cladding, were modelled in Serpent 2 to obtain reactor physics parameters, including important feedback parameters such as coolant void reactivity and fuel temperature coefficient. In addition, fuel heat transfer was modelled, and a simple accident model was tested on several ATF cases to compare with UO2. Overall, several concepts would require enrichment of uranium to avoid significant burnup penalties, particularly uranium-molybdenum (U-Mo and fully ceramic microencapsulated (FCM fuels. In addition, none of the fuel types have a significant advantage over UO2 in terms of overall accident response or coping time, though U-9Mo fuel melts significantly sooner due to its low melting point. Instead, the different ATF concepts appear to have more modest advantages, such as reduced fission product release upon cladding failure, or reduced hydrogen generation, though a proper risk assessment would be required to determine the magnitude of these advantages to weigh against economic disadvantages. The use of uranium nitride (UN enriched in N15 would increase exit burnup for natural uranium, providing a possible economic advantage depending on fuel manufacturing costs.

  2. Evaluation of health impact of cans and cladding from fuel processing

    International Nuclear Information System (INIS)

    Pages, P.; Lombard, J.

    1989-10-01

    Radiological impact related to exposure of workers during processing of tubes and cladding was based on data related to processing of 600 tons of uranium per year. Two reprocessing options were taken into account: processing by fusion and by cementation. Data are given for different posts. Apart from professional exposure data are given for public exposure. Two possible site were treated: reprocessing plant at the river bank (Rhone) or on the sea coast. (Manche)

  3. CLAD DEGRADATION - FEPS SCREENING ARGUMENTS

    International Nuclear Information System (INIS)

    R. Schreiner

    2004-01-01

    The purpose of this report is to evaluate and document the screening of the clad degradation features, events, and processes (FEPs) with respect to modeling used to support the Total System Performance Assessment-License Application (TSPA-LA). This report also addresses the effect of certain FEPs on both the cladding and the commercial spent nuclear fuel (CSNF), DOE-owned spent nuclear fuel (DSNF), and defense high-level waste (DHLW) waste forms, as appropriate to address the effects on multiple materials and both components (FEPs 2.1.09.09.0A, 2.1.09.11.0A, 2.1.11.05.0A, 2.1.12.02.0A, and 2.1.12.03.0A). These FEPs are expected to affect the repository performance during the postclosure regulatory period of 10,000 years after permanent closure. Table 1-1 provides the list of cladding FEPs, including their screening decisions (include or exclude). The primary purpose of this report is to identify and document the analysis, screening decision, and TSPA-LA disposition (for included FEPs) or screening argument (for excluded FEPs) for these FEPs related to clad degradation. In some cases, where a FEP covers multiple technical areas and is shared with other FEP reports, this report may provide only a partial technical basis for the screening of the FEP. The full technical basis for shared FEPs is addressed collectively by the sharing FEP reports. The screening decisions and associated TSPA-LA dispositions or screening arguments from all of the FEP reports are cataloged in a project-specific FEPs database

  4. Review and evaluation of cladding attack of LMFBR fuel

    International Nuclear Information System (INIS)

    Koizumi, M.; Nagai, S.; Furuya, H.; Muto, T.

    1977-01-01

    The behavior of cladding inner wall corrosion during irradiation was evaluated in terms of fuel density, fuel form, O/M ratio, plutonium concentration, cladding composition, cladding pretreatment, cladding inner diameter, burnup and cladding inner wall temperature. Factors which influence the corrosion are O/M ratio (oxygen to metal ratio), burn up, cladding inner diameter and cladding inner wall temperature. Maximum cladding inner wall corrosion depth was formulated as a function of O/M ratio, burn up and cladding inner wall temperature

  5. Fundamentals and industrial applications of high power laser beam cladding

    International Nuclear Information System (INIS)

    Bruck, G.J.

    1988-01-01

    Laser beam cladding has been refined such that clad characteristics are precisely determined through routine process control. This paper reviews the state of the art of laser cladding optical equipment, as well as the fundamental process/clad relationships that have been developed for high power processing. Major categories of industrial laser cladding are described with examples chose to highlight particular process attributes

  6. Eutectic reaction analysis between TRU-50%Zr alloy fuel and HT-9 cladding, and temperature prediction of eutectic reaction under steady-state

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Woan; Lee, Byoung Oon; Lee, Bong Sang; Park, Won Seok

    2001-02-01

    Blanket fuel assembly for HYPER contains a bundle of pins arrayed in triangular pitch, which has hexagonal bundle structure. The reference blanket fuel pin consists of the fuel slug of TRU-50wt%Zr alloy and the cladding material of ferritic martensite steel, HT-9. Chemical interaction between fuel slug and cladding is one of the major concerns in metallic fuel rod design. The contact of metallic fuel slug and stainless steel cladding in a fuel rod forms a complex multi-component diffusion couple at elevated temperatures. The potential problem of inter-diffusion of fuel and cladding components is essentially two-fold weakening of cladding mechanical strength due to the formation of diffusion zones in the cladding, and the formation of comparatively low melting point phases in the fuel/cladding interface to develop eutectic reaction. The main components of fuel slug are composed of zirconium alloying element in plutonium matrix, including neptunium, americium and uranium additionally. Therefore basic eutectic reaction change of Pu-Fe binary system can be assessed, while it is estimated how much other elements zirconium, uranium, americium and neptunium influence on plutonium phase stability. Afterwards it is needed that eutectic reaction is verified through experimental necessarily.

  7. Eutectic reaction analysis between TRU-50%Zr alloy fuel and HT-9 cladding, and temperature prediction of eutectic reaction under steady-state

    International Nuclear Information System (INIS)

    Hwang, Woan; Lee, Byoung Oon; Lee, Bong Sang; Park, Won Seok

    2001-02-01

    Blanket fuel assembly for HYPER contains a bundle of pins arrayed in triangular pitch, which has hexagonal bundle structure. The reference blanket fuel pin consists of the fuel slug of TRU-50wt%Zr alloy and the cladding material of ferritic martensite steel, HT-9. Chemical interaction between fuel slug and cladding is one of the major concerns in metallic fuel rod design. The contact of metallic fuel slug and stainless steel cladding in a fuel rod forms a complex multi-component diffusion couple at elevated temperatures. The potential problem of inter-diffusion of fuel and cladding components is essentially two-fold weakening of cladding mechanical strength due to the formation of diffusion zones in the cladding, and the formation of comparatively low melting point phases in the fuel/cladding interface to develop eutectic reaction. The main components of fuel slug are composed of zirconium alloying element in plutonium matrix, including neptunium, americium and uranium additionally. Therefore basic eutectic reaction change of Pu-Fe binary system can be assessed, while it is estimated how much other elements zirconium, uranium, americium and neptunium influence on plutonium phase stability. Afterwards it is needed that eutectic reaction is verified through experimental necessarily

  8. Unirradiated cladding rip-propagation tests

    International Nuclear Information System (INIS)

    Hu, W.L.; Hunter, C.W.

    1981-04-01

    The size of cladding rips which develop when a fuel pin fails can affect the subassembly cooling and determine how rapidly fuel escapes from the pin. The object of the Cladding Rip Propagation Test (CRPT) was to quantify the failure development of cladding so that a more realistic fuel pin failure modeling may be performed. The test results for unirradiated 20% CS 316 stainless steel cladding show significantly different rip propagation behavior at different temperatures. At room temperature, the rip growth is stable as the rip extension increases monotonically with the applied deformation. At 500 0 C, the rip propagation becomes unstable after a short period of stable rip propagation. The rapid propagation rate is approximately 200 m/s, and the critical rip length is 9 mm. At test temperatures above 850 0 C, the cladding exhibits very high failure resistances, and failure occurs by multiple cracking at high cladding deformation. 13 figures

  9. High Temperature Fuel Cladding Chemical Interactions Between TRIGA Fuels and 304 Stainless Steel

    Energy Technology Data Exchange (ETDEWEB)

    Perez, Emmanuel [Idaho National Lab. (INL), Idaho Falls, ID (United States); Keiser, Jr., Dennis D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Forsmann, Bryan [Boise State Univ., ID (United States); Janney, Dawn E. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Henley, Jody [Idaho National Lab. (INL), Idaho Falls, ID (United States); Woolstenhulme, Eric C. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-02-01

    High-temperature fuel-cladding chemical interactions (FCCI) between TRIGA (Training, Research, Isotopes, General Atomics) fuel elements and the 304 stainless steel (304SS) are of interest to develop an understanding of the fuel behavior during transient reactor scenarios. TRIGA fuels are composed of uranium (U) particles dispersed in a zirconium-hydride (Zr-H) matrix. In reactor, the fuel is encased in 304-stainless-steel (304SS) or Incoloy 800 clad tubes. At high temperatures, the fuel can readily interact with the cladding, resulting in FCCI. A number of FCCI can take place in this system. Interactions can be expected between the cladding and the Zr-H matrix, and/or between the cladding and the U-particles. Other interactions may be expected between the Zr-H matrix and the U-particles. Furthermore, the fuel contains erbium-oxide (Er-O) additions. Interactions can also be expected between the Er-O, the cladding, the Zr-H and the U-particles. The overall result is that very complex interactions may take place as a result of fuel and cladding exposures to high temperatures. This report discusses the characterization of the baseline fuel microstructure in the as-received state (prior to exposure to high temperature), characterization of the fuel after annealing at 950C for 24 hours and the results from diffusion couple experiments carries out at 1000C for 5 and 24 hours. Characterization was carried out via scanning electron microscopy (SEM) and transmission electron microscopy (TEM) with sample preparation via focused ion beam in situ-liftout-technique.

  10. High Temperature Fuel Cladding Chemical Interactions Between TRIGA Fuels and 304 Stainless Steel

    International Nuclear Information System (INIS)

    Perez, Emmanuel; Keiser Jr, Dennis D.; Forsmann, Bryan; Janney, Dawn E.; Henley, Jody; Woolstenhulme, Eric C.

    2016-01-01

    High-temperature fuel-cladding chemical interactions (FCCI) between TRIGA (Training, Research, Isotopes, General Atomics) fuel elements and the 304 stainless steel (304SS) are of interest to develop an understanding of the fuel behavior during transient reactor scenarios. TRIGA fuels are composed of uranium (U) particles dispersed in a zirconium-hydride (Zr-H) matrix. In reactor, the fuel is encased in 304-stainless-steel (304SS) or Incoloy 800 clad tubes. At high temperatures, the fuel can readily interact with the cladding, resulting in FCCI. A number of FCCI can take place in this system. Interactions can be expected between the cladding and the Zr-H matrix, and/or between the cladding and the U-particles. Other interactions may be expected between the Zr-H matrix and the U-particles. Furthermore, the fuel contains erbium-oxide (Er-O) additions. Interactions can also be expected between the Er-O, the cladding, the Zr-H and the U-particles. The overall result is that very complex interactions may take place as a result of fuel and cladding exposures to high temperatures. This report discusses the characterization of the baseline fuel microstructure in the as-received state (prior to exposure to high temperature), characterization of the fuel after annealing at 950C for 24 hours and the results from diffusion couple experiments carries out at 1000C for 5 and 24 hours. Characterization was carried out via scanning electron microscopy (SEM) and transmission electron microscopy (TEM) with sample preparation via focused ion beam in situ-liftout-technique.

  11. Effect of laser power on clad metal in laser-TIG combined metal cladding

    Science.gov (United States)

    Utsumi, Akihiro; Hino, Takanori; Matsuda, Jun; Tasoda, Takashi; Yoneda, Masafumi; Katsumura, Munehide; Yano, Tetsuo; Araki, Takao

    2003-03-01

    TIG arc welding has been used to date as a method for clad welding of white metal as bearing material. We propose a new clad welding process that combines a CO2 laser and a TIG arc, as a method for cladding at high speed. We hypothesized that this method would permit appropriate control of the melted quantity of base metal by varying the laser power. We carried out cladding while varying the laser power, and investigated the structure near the boundary between the clad layer and the base metal. Using the laser-TIG combined cladding, we found we were able to control appropriately the degree of dilution with the base metal. By applying this result to subsequent cladding, we were able to obtain a clad layer of high quality, which was slightly diluted with the base metal.

  12. Dynamic strain aging of zircaloy-4 PWR fuel cladding in biaxial stress state

    International Nuclear Information System (INIS)

    Park, Ki Seong; Lee, Byong Whi

    1989-01-01

    The expanding copper mandrel test performed at three strain rates (3.2x10E-5/s,2.0x10E-6/s and 1.2x10E-7/s) over 553-873 K temperature range by varying the heating rates (8-10deg C/s,1-2deg C/s and 0.5deg C/s) in air and in vacuum (5x10E-5 torr). The yield stress peak, the strain rate sensitivity minimum and the activation volume peaks could be explained in terms of the dynamic strain aging. The activation energy for dynamic strain aging obtained from the yield stress peak temperature and strain rate was 196 KJ/mol and this value was in good agreement with the activation energy for oxygen diffusion in α-zirconium and Zircaloy-2 (207-220KJ/mol). Therefore, oxygen atoms are responsible for the dynamic strain aging which appeared between 573K and 673K. The yield stress increase due to the oxidation was obtained by comparing the yield stress in air with that in vacuum and represented by the percentage increase of yield stress (σ y a -σ y v /σ y v ). The slower the strain rate, the greater the percentage increase occurs. In order to estimate the yield stress of PWR fuel cladding material under the service environment, the yield stress in water was obtained by comparing the oxidation rate in air that in water assuming the relationship between the oxygen pick-up amount and the yield stress increase. (Author)

  13. LASER SURFACE CLADDING FOR STRUCTURAL REPAIR

    OpenAIRE

    SANTANU PAUL

    2018-01-01

    Laser cladding is a powder deposition technique, which is used to deposit layers of clad material on a substrate to improve its surface properties. It has widespread application in the repair of dies and molds used in the automobile industry. These molds and dies are subjected to cyclic thermo-mechanical loading and therefore undergo localized damage and wear. The final clad quality and integrity is influenced by various physical phenomena, namely, melt pool morphology, microst...

  14. Nuclear-powered pacemaker fuel cladding study

    International Nuclear Information System (INIS)

    Shoup, R.L.

    1976-07-01

    The fabrication of fuel capsules with refractory metal and alloy clads used in nuclear-powered cardiac pacemakers precludes the expedient dissolution of the clad in inorganic acid solutions. An experiment to measure penetration rates of acids on commonly used fuel pellet clads indicated that it is not impossible, but that it would be very difficult to dissolve the multiple cladding. This work was performed because of a suggestion that a 238 PuO 2 -powered pacemaker could be transformed into a terrorism weapon

  15. Laser surface cladding:a literature survey

    OpenAIRE

    Gedda, Hans

    2000-01-01

    This work consists of a literature survey of a laser surface cladding in order to investigate techniques to improve the cladding rate for the process. The high local heat input caused by the high power density of the laser generates stresses and the process is consider as slow when large areas are processed. To avoid these disadvantages the laser cladding process velocity can be increased three or four times by use of preheated wire instead of the powder delivery system. If laser cladding is ...

  16. Modelling cladding response to changing conditions

    Energy Technology Data Exchange (ETDEWEB)

    Tulkki, Ville; Ikonen, Timo [VTT Technical Research Centre of Finland ltd (Finland)

    2016-11-15

    The cladding of the nuclear fuel is subjected to varying conditions during fuel reactor life. Load drops and reversals can be modelled by taking cladding viscoelastic behaviour into account. Viscoelastic contribution to the deformation of metals is usually considered small enough to be ignored, and in many applications it merely contributes to the primary part of the creep curve. With nuclear fuel cladding the high temperature and irradiation as well as the need to analyse the variable load all emphasise the need to also inspect the viscoelasticity of the cladding.

  17. Pulsed Laser Cladding of Ni Based Powder

    Science.gov (United States)

    Pascu, A.; Stanciu, E. M.; Croitoru, C.; Roata, I. C.; Tierean, M. H.

    2017-06-01

    The aim of this paper is to optimize the operational parameters and quality of one step Metco Inconel 718 atomized powder laser cladded tracks, deposited on AISI 316 stainless steel substrate by means of a 1064 nm high power pulsed laser, together with a Precitec cladding head manipulated by a CLOOS 7 axes robot. The optimization of parameters and cladding quality has been assessed through Taguchi interaction matrix and graphical output. The study demonstrates that very good cladded layers with low dilution and increased mechanical proprieties could be fabricated using low laser energy density by involving a pulsed laser.

  18. Cladding creepdown model for FRAPCON-2

    International Nuclear Information System (INIS)

    Shah, V.N.; Tolli, J.E.

    1985-02-01

    This report presents a cladding deformation model developed to analyze cladding creepdown during steady state operation in both a pressurized water reactor (PWR) and a boiling water reactor (BWR). This model accounts for variations in zircaloy cladding heat treatment; cold worked and stress relieved material, typically used in a PWR, and fully recrystallized material, typically used in a BWR. The model calculates cladding creepdown as a function of hoop stress, fast neutron flux, exposure time, and temperature. This report also presents a comparison between cladding creep calculations by this model and corresponding measurements from the KWU/CE program, ORNL HOBBIE experiments, and EPRI/Westinghouse Engineering cooperative project. The comparisons show that the model calculates cladding creep strains well. The analyses of non-fueled rods by FRAPCON-2 show that the cladding creepdown model was correctly incorporated. Also, analysis of a PWR rod test case shows that the FRAPCON-2 code can analyze pellet-cladding mechanical interaction caused by cladding creepdown and fuel swelling

  19. Protective claddings for high strength chromium alloys

    Science.gov (United States)

    Collins, J. F.

    1971-01-01

    The application of a Cr-Y-Hf-Th alloy as a protective cladding for a high strength chromium alloy was investigated for its effectiveness in inhibiting nitrogen embrittlement of a core alloy. Cladding was accomplished by a combination of hot gas pressure bonding and roll cladding techniques. Based on bend DBTT, the cladding alloy was effective in inhibiting nitrogen embrittlement of the chromium core alloy for up to 720 ks (200hours) in air at 1422 K (2100 F). A significant increase in the bend DBTT occurred with longer time exposures at 1422 K or short time exposures at 1589 K (2400 F).

  20. Engineered Zircaloy Cladding Modifications for Improved Accident Tolerance of LWR Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Heuser, Brent [Univ. of Illinois, Urbana-Champaign, IL (United States); Stubbins, James [Univ. of Illinois, Urbana-Champaign, IL (United States); Kozlowski, Tomasz [Univ. of Illinois, Urbana-Champaign, IL (United States); Uddin, Rizwan [Univ. of Illinois, Urbana-Champaign, IL (United States); Trinkle, Dallas [Univ. of Illinois, Urbana-Champaign, IL (United States); Downar, Thoms [Univ. of Michigan, Ann Arbor, MI (United States); Was, Gary [Univ. of Michigan, Ann Arbor, MI (United States); ang, Yong [Univ. of Florida, Gainesville, FL (United States); Phillpot, Simon [Univ. of Florida, Gainesville, FL (United States); Sabharwall, piyush [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2017-07-25

    The DOE NEUP sponsored IRP on accident tolerant fuel (ATF) entitled Engineered Zircaloy Cladding Modifications for Improved Accident Tolerance of LWR Nuclear Fuel involved three academic institutions, Idaho National Laboratory (INL), and ATI Materials (ATI). Detailed descriptions of the work at the University of Illinois (UIUC, prime), the University of Florida (UF), the University of Michigan (UMich), and INL are included in this document as separate sections. This summary provides a synopsis of the work performed across the IRP team. Two ATF solution pathways were initially proposed, coatings on monolithic Zr-based LWR cladding material and selfhealing modifications of Zr-based alloys. The coating pathway was extensively investigated, both experimentally and in computations. Experimental activities related to ATF coatings were centered at UIUC, UF, and UMich and involved coating development and testing, and ion irradiation. Neutronic and thermal hydraulic aspects of ATF coatings were the focus of computational work at UIUC and UMich, while materials science aspects were the focus of computational work at UF and INL. ATI provided monolithic Zircaloy 2 and 4 material and a binary Zr-Y alloy material. The selfhealing pathway was investigated with advanced computations only. Beryllium was identified as a valid self-healing additive early in this work. However, all attempts to fabricate a Zr-Be alloy failed. Several avenues of fabrication were explored. ATI ultimately declined our fabrication request over health concerns associated with Be (we note that Be was not part of the original work scope and the ATI SOW). Likewise, Ames Laboratory declined our fabrication request, citing known litigation dating to the 1980s and 1990s involving the U.S. Federal government and U.S. National Laboratory employees involving the use of Be. Materion (formerly, Brush Wellman) also declined our fabrication request, citing the difficulty in working with a highly reactive Zr and Be

  1. Engineered Zircaloy Cladding Modifications for Improved Accident Tolerance of LWR Nuclear Fuel

    International Nuclear Information System (INIS)

    Heuser, Brent; Stubbins, James; Kozlowski, Tomasz; Uddin, Rizwan; Trinkle, Dallas; Downar, Thoms; Was, Gary; Ang, Yong; Phillpot, Simon; Sabharwall, Piyush

    2017-01-01

    The DOE NEUP sponsored IRP on accident tolerant fuel (ATF) entitled Engineered Zircaloy Cladding Modifications for Improved Accident Tolerance of LWR Nuclear Fuel involved three academic institutions, Idaho National Laboratory (INL), and ATI Materials (ATI). Detailed descriptions of the work at the University of Illinois (UIUC, prime), the University of Florida (UF), the University of Michigan (UMich), and INL are included in this document as separate sections. This summary provides a synopsis of the work performed across the IRP team. Two ATF solution pathways were initially proposed, coatings on monolithic Zr-based LWR cladding material and selfhealing modifications of Zr-based alloys. The coating pathway was extensively investigated, both experimentally and in computations. Experimental activities related to ATF coatings were centered at UIUC, UF, and UMich and involved coating development and testing, and ion irradiation. Neutronic and thermal hydraulic aspects of ATF coatings were the focus of computational work at UIUC and UMich, while materials science aspects were the focus of computational work at UF and INL. ATI provided monolithic Zircaloy 2 and 4 material and a binary Zr-Y alloy material. The selfhealing pathway was investigated with advanced computations only. Beryllium was identified as a valid self-healing additive early in this work. However, all attempts to fabricate a Zr-Be alloy failed. Several avenues of fabrication were explored. ATI ultimately declined our fabrication request over health concerns associated with Be (we note that Be was not part of the original work scope and the ATI SOW). Likewise, Ames Laboratory declined our fabrication request, citing known litigation dating to the 1980s and 1990s involving the U.S. Federal government and U.S. National Laboratory employees involving the use of Be. Materion (formerly, Brush Wellman) also declined our fabrication request, citing the difficulty in working with a highly reactive Zr and Be

  2. BWR fuel clad behaviour following LOCA

    International Nuclear Information System (INIS)

    Chaudhry, S.M.; Vyas, K.N.; Dinesh Babu, R.

    1996-01-01

    Flow and pressure through the fuel coolant channel reduce rapidly following a loss of coolant accident. Due to stored energy and decay heat, fuel and cladding temperatures rise rapidly. Increase in clad temperature causes deterioration of mechanical properties of clad material. This coupled with increase of pressure inside the cladding due to accumulation of fission gases and de-pressurization of coolant causes the cladding to balloon. This phenomenon is important as it can reduce or completely block the flow passages in a fuel assembly causing reduction of emergency coolant flow. Behaviour of a BWR clad is analyzed in a design basis LOCA. Fuel and clad temperatures following a LOCA are calculated. Fission gas release and pressure is estimated using well established models. An elasto-plastic analysis of clad tube is carried out to determine plastic strains and corresponding deformations using finite-element technique. Analysis of neighbouring pins gives an estimate of flow areas available for emergency coolant flow. (author). 7 refs, 6 figs, 3 tabs

  3. Cladding properties under simulated fuel pin transients

    International Nuclear Information System (INIS)

    Hunter, C.W.; Johnson, G.D.

    1975-01-01

    A description is given of the HEDL fuel pin testing program utilizing a recently developed Fuel Cladding Transient Tester (FCTT) to generate the requisite mechanical property information on irradiated and unirradiated fast reactor fuel cladding under temperature ramp conditions. The test procedure is described, and data are presented

  4. Analysis of corrosion behavior of KOFA cladding

    International Nuclear Information System (INIS)

    Lee, Chan Bock; Kim, Ki Hang; Seo, Keum Seok; Chung, Jin Gon

    1994-01-01

    The corrosion behavior of KOFA cladding was analyzed using the oxide measurement data of KOFA fuel irradiated up to the fuel rod burnup of 35,000 MWD/MTU for two cycles in Kori-2. Even though KOFA cladding is a standard Zircaloy-4 manufactured by Westinghouse according to the Siemens/KWU's HCW (Highly Cold Worked) standard Zircaloy-4 specification, it was expected that in-pile corrosion behavior of KOFA cladding would not be equivalent to that of Siemens/KWU's cladding due to the differences in such manufacturing processes as cold work and heat treatment. The analysis of measured KOFA cladding oxidation showed that oxidation of KOFA cladding is at least 19 % lower than the design analysis based upon Siemens/KWU's HCW standard Zircaloy-4 cladding. Lower corrosion of KOFA cladding seems to result from the differences in the manufacturing processes and chemical composition although the burnup and oxide layer thickness of the measured fuel rods is relatively low and the amount of the oxidation data base is small

  5. Development of advanced zirconium fuel cladding

    International Nuclear Information System (INIS)

    Jeong, Young Hwan; Park, S. Y.; Lee, M. H.

    2007-04-01

    This report includes the manufacturing technology developed for HANA TM claddings, a series of their characterization results as well as the results of their in-pile and out-of pile performances tests which were carried out to develop some fuel claddings for a high burn-up (70,000MWd/mtU) which are competitive in the world market. Some of the HANA TM claddings, which had been manufactured based on the results from the 1st and 2nd phases of the project, have been tested in a research reactor in Halden of Norway for an in-pile performance qualification. The results of the in-pile test showed that the performance of the HANA TM claddings for corrosion and creep was better than 50% compared to that of Zircaloy-4 or A cladding. It was also found that the out-of pile performance of the HANA TM claddings for such as LOCA and RIA in some accident conditions corrosion creep, tensile, burst and fatigue was superior or equivalent to that of the Zircaloy-4 or A cladding. The project also produced the other many data which were required to get a license for an in-pile test of HANA TM claddings in a commercial reactor. The data for the qualification or characterization were provided for KNFC to assist their activities to get the license for the in-pile test of HANA TM Lead Test Rods(LTR) in a commercial reactor

  6. Corrosion characteristics of K-claddings

    International Nuclear Information System (INIS)

    Park, J. Y.; Choi, B. K.; Jung, Y. H.; Jung, Y. H.

    2004-01-01

    The Improvement of the corrosion resistance of nuclear fuel claddings is the critical issue for the successful development of the high burn-up fuel. KAERI have developed the K-claddings having a superior corrosion resistance by controlling the alloying element addition and optimizing the manufacturing process. The comparative evaluation of the corrosion resistance for K-claddings and the foreign claddings was performed and the effect of the heat treatment on the corrosion behavior of K-claddings was also examined. Corrosion tests were carried out in the conditions of 360 .deg. C pure water, PWR-simulating loop and 400 .deg. C steam, From the results of the corrosion tests, it was found that the corrosion resistance of K-claddings is superior to those of Zry4 and A claddings and K6 showed a better corrosion resistance than K3. The corrosion behavior of K-cladding was strongly influenced by the final annealing rather than the intermediate annealing, and the corrosion resistance increased with decreasing the final annealing temperature

  7. Cladding Alloys for Fluoride Salt Compatibility

    Energy Technology Data Exchange (ETDEWEB)

    Muralidharan, Govindarajan [ORNL; Wilson, Dane F [ORNL; Walker, Larry R [ORNL; Santella, Michael L [ORNL; Holcomb, David Eugene [ORNL

    2011-06-01

    This report provides an overview of several candidate technologies for cladding nickel-based corrosion protection layers onto high-temperature structural alloys. The report also provides a brief overview of the welding and weld performance issues associated with joining nickel-clad nickel-based alloys. From the available techniques, two cladding technologies were selected for initial evaluation. The first technique is a line-of-sight method that would be useful for cladding large structures such as vessel interiors or large piping. The line-of-sight method is a laser-based surface cladding technique in which a high-purity nickel powder mixed into a polymer binder is first sprayed onto the surface, baked, and then rapidly melted using a high-power laser. The second technique is a vapor phase technique based on the nickel-carbonyl process that is suitable for cladding inaccessible surfaces such as the interior surfaces of heat exchangers. An initial evaluation for performed on the quality of nickel claddings processed using the two selected cladding techniques.

  8. Use of ion beams to simulate reaction of reactor fuels with their cladding

    International Nuclear Information System (INIS)

    Birtcher, R.C.; Baldo, P.

    2006-01-01

    Processes occurring within reactor cores are not amenable to direct experimental observation. Among major concerns are damage, fission gas accumulation and reaction between the fuel and its cladding all of which lead to swelling. These questions can be investigated through simulation with ion beams. As an example, we discuss the irradiation driven interaction of uranium-molybdenum alloys, intended for use as low-enrichment reactor fuels, with aluminum, which is used as fuel cladding. Uranium-molybdenum coated with a 100 nm thin film of aluminum was irradiated with 3 MeV Kr ions to simulate fission fragment damage. Mixing and diffusion of aluminum was followed as a function of irradiation with RBS and nuclear reaction analysis using the 27 Al(p,γ) 28 Si reaction which occurs at a proton energy of 991.9 keV. During irradiation at 150 deg. C, aluminum diffused into the uranium alloy at a irradiation driven diffusion rate of 30 nm 2 /dpa. At a dose of 90 dpa, uranium diffusion into the aluminum layer resulted in formation of an aluminide phase at the initial interface. The thickness of this phase grew until it consumed the aluminum layer. The rapid diffusion of Al into these reactor fuels may offer explanation of the observation that porosity is not observed in the fuel particles but on their periphery

  9. Zr-rich layers electrodeposited onto stainless steel cladding during the electrorefining of EBR-II fuel

    International Nuclear Information System (INIS)

    Keiser, D.D. Jr.; Mariani, R.D.

    1999-01-01

    Argonne National Laboratory is developing an electrometallurgical treatment for spent nuclear fuels. The initial demonstration of this process is being conducted on U-Zr alloy fuel elements irradiated in the experimental breeder reactor II (EBR-II). We report the first metallographic characterization of cladding hull remains for the electrometallurgical treatment of spent metallic fuel. During the electrorefining process, Zr-rich layers, with some U, deposit on all exposed surfaces of irradiated cladding segments (hulls) that originally contained the fuel alloy that was being treated. In some cases, not only was residual Zr (and U) found inside the cladding hulls, but a Zr-rind was also observed near the interior cladding hull surface. The Zr-rind was originally formed during the fuel casting process on the fuel slug. The observation of Zr deposits on all exposed cladding surfaces is explained with thermodynamic principles, when two conditions are met. These conditions are partial oxidation of Zr and the presence of residual uranium in the hulls when the electrorefining experiment is terminated. Comparisons are made between the structure of the initial irradiated fuel before electrorefining and the morphology of the material remaining in the cladding hulls after electrorefining. (orig.)

  10. Flaw behavior in mechanically loaded clad plates

    International Nuclear Information System (INIS)

    Iskander, S.K.; Robinson, G.C.; Oland, C.B.

    1989-01-01

    A small crack near the inner surface of clad nuclear reactor pressure vessels is an important consideration in the safety assessment of the structural integrity of the vessel. Four-point bend tests on large plate specimens, conforming to ASTM specification for pressure vessel plates, alloy steels, quenched and tempered, Mn-Mo and Mn-Mo-Ni (A533) grade B six clad and two unclad with stainless steels 308, 309 and 312 weld wires, were performed to determine the effect of cladding upon the propagation of small surface cracks subjected to stress states. Results indicated that the tough surface layer composed of cladding and/or heat-affected zone has enhanced the load-bearing capacity of plates under conditions where unclad plates have ruptured. The results are interpreted in terms of fracture mechanics. The behavior of flaws in clad reactor pressure vessels is examined in the light of the test results. 11 refs., 8 figs., 2 tabs

  11. Some proposed mechanisms for internal cladding corrosion

    International Nuclear Information System (INIS)

    Bradbury, M.H.; Pickering, S.; Whitlow, W.H.

    1977-01-01

    In spite of extensive research during recent years, a comprehensive model for internal cladding corrosion in fast reactor oxide fuel pins has not yet been established. In this paper, a model is proposed which accounts for many of the features normally associated with this type of corrosion. The model is composed of a number of parts which describe the chronological sequence of events at the fuel/cladding interface. The corrosion reaction is visualised as being primarily chemical in character, involving the cladding steel, the fuel and the more aggressive fission products, notably caesium in the presence of oxygen. The model attempts to explain how corrosion starts, how it depends on the oxygen potential, why it occurs non-uniformly; also covered are phase changes within the cladding steel and morphological features such as the intergranular form of attack and the distribution of corrosion products in the fuel/cladding gap. (author)

  12. Some proposed mechanisms for internal cladding corrosion

    Energy Technology Data Exchange (ETDEWEB)

    Bradbury, M H; Pickering, S; Whitlow, W H [EURATOM (United Kingdom)

    1977-04-01

    In spite of extensive research during recent years, a comprehensive model for internal cladding corrosion in fast reactor oxide fuel pins has not yet been established. In this paper, a model is proposed which accounts for many of the features normally associated with this type of corrosion. The model is composed of a number of parts which describe the chronological sequence of events at the fuel/cladding interface. The corrosion reaction is visualised as being primarily chemical in character, involving the cladding steel, the fuel and the more aggressive fission products, notably caesium in the presence of oxygen. The model attempts to explain how corrosion starts, how it depends on the oxygen potential, why it occurs non-uniformly; also covered are phase changes within the cladding steel and morphological features such as the intergranular form of attack and the distribution of corrosion products in the fuel/cladding gap. (author)

  13. Czechoslovak uranium

    International Nuclear Information System (INIS)

    Pluskal, O.

    1992-01-01

    Data and knowledge related to the prospecting, mining, processing and export of uranium ores in Czechoslovakia are presented. In the years between 1945 and January 1, 1991, 98,461.1 t of uranium were extracted. In the period 1965-1990 the uranium industry was subsidized from the state budget to a total of 38.5 billion CSK. The subsidies were put into extraction, investments and geologic prospecting; the latter was at first, ie. till 1960 financed by the former USSR, later on the two parties shared costs on a 1:1 basis. Since 1981 the prospecting has been entirely financed from the Czechoslovak state budget. On Czechoslovak territory uranium has been extracted from deposits which may be classified as vein-type deposits, deposits in uranium-bearing sandstones and deposits connected with weathering processes. The future of mining, however, is almost exclusively being connected with deposits in uranium-bearing sandstones. A brief description and characteristic is given of all uranium deposits on Czechoslovak territory, and the organization of uranium mining in Czechoslovakia is described as is the approach used in the world to evaluate uranium deposits; uranium prices and actual resources are also given. (Z.S.) 3 figs

  14. Metal-clad waveguide sensors

    DEFF Research Database (Denmark)

    Skivesen, Nina

    This work concerns planar optical waveguide sensors for biosensing applications, with the focus on deep-probe sensing for micron-scale biological objects like bacteria and whole cells. In the last two decades planar metal-clad waveguides have been brieflyintroduced in the literature applied...... for various biosensing applications, however a thorough study of the sensor configurations has not been presented, but is the main subject of this thesis. Optical sensors are generally well suited for bio-sensing asthey show high sensitivity and give an immediate response for minute changes in the refractive...... index of a sample, due to the high sensitivity of optical bio-sensors detection of non-labeled biological objects can be performed. The majority of opticalsensors presented in the literature and commercially available optical sensors are based on evanescent wave sensing, however most of these sensors...

  15. Cladding failure margins for metallic fuel in the integral fast reactor

    International Nuclear Information System (INIS)

    Bauer, T.H.; Fenske, G.R.; Kramer, J.M.

    1987-01-01

    The Integral Fast Reactor (IFR) concept being developed at Argonne National Laboratory has prompted a renewed interest in uranium-based metal alloys as a fuel for sodium-cooled fast reactors. In this paper we will present recent measurements of cladding eutectic penetration rates for the ternary IFR alloy and will compare these results with earlier eutectic penetration data for other fuel and cladding materials. A method for calculating failure of metallic fuel pins is developed by combining cladding deformation equations with a large strain analysis where the hoop stress is calculated using the instantaneous wall thickness as determined from correlations of the eutectic penetration-rate data. This method is applied to analyze the results of in-reactor and out-of-reactor fuel pin failure tests on uranium-fissium alloy EBR-II Mark-II driver fuel. In the final section of this paper we extend the calculations to consider the failure of IFR ternary fuel under reactor accident conditions. (orig./GL)

  16. Uranium ores

    International Nuclear Information System (INIS)

    Poty, B.; Roux, J.

    1998-01-01

    The processing of uranium ores for uranium extraction and concentration is not much different than the processing of other metallic ores. However, thanks to its radioactive property, the prospecting of uranium ores can be performed using geophysical methods. Surface and sub-surface detection methods are a combination of radioactive measurement methods (radium, radon etc..) and classical mining and petroleum prospecting methods. Worldwide uranium prospecting has been more or less active during the last 50 years, but the rise of raw material and energy prices between 1970 and 1980 has incited several countries to develop their nuclear industry in order to diversify their resources and improve their energy independence. The result is a considerable increase of nuclear fuels demand between 1980 and 1990. This paper describes successively: the uranium prospecting methods (direct, indirect and methodology), the uranium deposits (economical definition, uranium ores, and deposits), the exploitation of uranium ores (use of radioactivity, radioprotection, effluents), the worldwide uranium resources (definition of the different categories and present day state of worldwide resources). (J.S.)

  17. Uranium market

    International Nuclear Information System (INIS)

    Rubini, L.A.; Asem, M.A.D.

    1990-01-01

    The historical development of the uranium market is present in two periods: The initial period 1947-1970 and from 1970 onwards, with the establishment of a commercial market. The world uranium requirements are derived from the corresponding forecast of nuclear generating capacity, with, particular emphasis to the brazilian requirements. The forecast of uranium production until the year 2000 is presented considering existing inventories and the already committed demand. The balance between production and requirements is analysed. Finally the types of contracts currently being used and the development of uranium prices in the world market are considered. (author)

  18. Uranium enrichment

    International Nuclear Information System (INIS)

    1990-01-01

    This report looks at the following issues: How much Soviet uranium ore and enriched uranium are imported into the United States and what is the extent to which utilities flag swap to disguise these purchases? What are the U.S.S.R.'s enriched uranium trading practices? To what extent are utilities required to return used fuel to the Soviet Union as part of the enriched uranium sales agreement? Why have U.S. utilities ended their contracts to buy enrichment services from DOE?

  19. Investigation and recovery of unrecovered fuel pellets and cladding tube pieces

    International Nuclear Information System (INIS)

    Kobayashi, Keiji

    1980-01-01

    The total weight of the fuel pellets lost due to break was about 1206 g, and cladding tube pieces were about 217 g. Among these, the pellets of about 527 g and the cladding tube pieces of about 152 g were recovered when broken fuel rods were discovered. It is not desirable to leave these broken pieces as unrecovered in view of safety and the management of nuclear fuel materials. Kansai Electric Power Co., Inc., investigated the position and the amount of these pellets and cladding tube pieces for about a year, and recovered a part of them. The results were written in two reports. The objects of the investigation and recovery, and the method of recovery are explained. The UO 2 and zirconium recovered were 58.52 g and 369.58 g, respectively. The solid pellets were recovered from the reactor, fuel assemblies, a spent fuel pit and canals, and the content in sludge was recovered from other installations. The amounts of unrecovered pellets and cladding tube pieces in primary cooling water, coolant filters, sealing water filters, primary cooling pipes, waste resins and fuel assemblies were estimated. The problems concerning the recovery and estimation are pointed out. The results of estimating the amount of uranium in coolant filters and sealing water filters are useful to know the time of the occurrence of accident. (Kako, I.)

  20. Fuel cladding mechanical interaction during power ramps

    International Nuclear Information System (INIS)

    Guerin, Y.

    1985-01-01

    Mechanical interaction between fuel and cladding may occur as a consequence of two types of phenomenon: i) fuel swelling especially at levels of caesium accumulation, and ii) thermal differential expansion during power changes. Slow overpower ramps which may occur during incidental events are of course one of the circumstances responsible for this second type of fuel cladding mechanical interaction (FCMI). Experiments and analysis of this problem that have been done at C.E.A. allow to determine the main parameters which will fix the level of stress and the risk of damage induced by the fuel in the cladding during overpower transients

  1. Diffused zircaloy 2/stainless steel junctions

    International Nuclear Information System (INIS)

    Jacques, F.

    1964-01-01

    The diffusion permits to realize joints between two different materials, in fact of the formation of a liquid phase at the contact face. The study of the tensile properties allowed the determination of the ideal conditions for the diffusion treatment which are, within 2 and 3 minutes for a temperature within 1020 C and 1030 C. The characteristics of the so obtained joints were, studied: mechanical properties, tightness, resistance to thermal cycling. Analysis of the thermal stress, owing to the differential dilatation of the two materials mode the object of a particular study. The investigation on the diffusion zone, includes specially, an analysis of the constituents distribution formed during the diffusion treatment. (author) [fr

  2. Chemical compatibility between cladding alloys and advanced fuels

    International Nuclear Information System (INIS)

    Fee, D.C.; Johnson, C.E.

    1975-05-01

    The National Advanced Fuels Program requires chemical, mechanical, and thermophysical properties data for cladding alloys. The compatibility behavior of cladding alloys with advanced fuels is critically reviewed. in carbide fuel pins, the principal compatibility problem is cladding carburization, diffusion of carbon into the cladding matrix accompanied by carbide precipitation. Carburization changes the mechanical properties of the cladding alloy. The extent of carburization increases in sodium (versus gas) bonded fuels. The depth of carburization increases with increasing sesquicarbide (M 2 C 3 ) content of the fuel. In nitride fuel pins, the principal compatibility problem is cladding nitriding, diffusion of nitrogen into the cladding matrix accompanied by nitride precipitation. Nitriding changes the mechanical properties of the cladding alloy. In both carbide and nitride fuel pins, fission products do not migrate appreciably to the cladding and do not appear to contribute to cladding attack. 77 references. (U.S.)

  3. Corrosion of research reactor aluminium clad spent fuel in water

    International Nuclear Information System (INIS)

    2003-01-01

    This report describes research performed in ten laboratories within the framework of the IAEA Co-ordinated Research Project on Corrosion of Research Reactor Aluminium Clad Spent Fuel in Water. The project consisted of exposure of standard racks of corrosion coupons in the spent fuel pools of the participating research reactor laboratories and the evaluation of the coupons after predetermined exposure times, along with periodic monitoring of the storage water. A group of experts in the field contributed a state of the art review and provided technical supervision of the project. Localized corrosion mechanisms are notoriously difficult to understand, and it was clear from the outset that obtaining consistency in the results and their interpretation from laboratory to laboratory would depend on the development of an excellent set of experimental protocols. These experimental protocols are described in the report together with guidelines for the maintenance of optimum water chemistry to minimize the corrosion of aluminium clad research reactor fuel in wet storage. A large database on corrosion of aluminium clad materials has been generated from the CRP and the SRS corrosion surveillance programme. An evaluation of these data indicates that the most important factors contributing to the corrosion of the aluminium are: (1) High water conductivity (100-200 μS/cm); (2) Aggressive impurity ion concentrations (Cl - ); (3) Deposition of cathodic particles on aluminium (Fe, etc.); (4) Sludge (containing Fe, Cl - and other ions in concentrations greater than ten times the concentrations in the water); (5) Galvanic couples between dissimilar metals (stainless steel-aluminium, aluminium-uranium, etc); (6) Scratches and imperfections (in protective oxide coating on cladding); (7) Poor water circulation. These factors operating both independently and synergistically may cause corrosion of the aluminium. The single most important key to preventing corrosion is maintaining good

  4. Friction Surface Cladding of AA1050 on AA2024-T351; influence of clad layer thickness and tool rotation rate

    NARCIS (Netherlands)

    Liu, Shaojie; Bor, Teunis Cornelis; Geijselaers, Hubertus J.M.; Akkerman, Remko

    2015-01-01

    Friction Surfacing Cladding (FSC) is a recently developed solid state process to deposit thin metallic clad layers on a substrate. The process employs a rotating tool with a central opening to supply clad material and support the distribution and bonding of the clad material to the substrate. The

  5. Irradiation of TZM: Uranium dioxide fuel pin at 1700 K

    Science.gov (United States)

    Mcdonald, G. E.

    1973-01-01

    A fuel pin clad with TZM and containing solid pellets of uranium dioxide was fission heated in a static helium-cooled capsule at a maximum surface temperature of 1700 K for approximately 1000 hr and to a total burnup of 2.0 percent of the uranium-235. The results of the postirradiation examination indicated: (1) A transverse, intergranular failure of the fuel pin occurred when the fuel pin reached 2.0-percent burnup. This corresponds to 1330 kW-hr/cu cm, where the volume is the sum of the fuel, clad, and void volumes in the fuel region. (2) The maximum swelling of the fuel pin was less than 1.5 percent on the fuel-pin diameter. (3) There was no visible interaction between the TZM clad and the UO2. (4) Irradiation at 1700 K produced a course-grained structure, with an average grain diameter of 0.02 centimeter and with some of the grains extending one-half of the thickness of the clad. (5) Below approximately 1500 K, the irradiation of the clad produced a moderately fine-grained structure, with an average grain diameter of 0.004 centimeter.

  6. Uranium mining

    International Nuclear Information System (INIS)

    Lange, G.

    1975-01-01

    The winning of uranium ore is the first stage of the fuel cycle. The whole complex of questions to be considered when evaluating the profitability of an ore mine is shortly outlined, and the possible mining techniques are described. Some data on uranium mining in the western world are also given. (RB) [de

  7. Pellet-Cladding Mechanical Interaction Failure Threshold for Reactivity Initiated Accidents for Pressurized Water Reactors and Boiling Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Beyer, Carl E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Geelhood, Kenneth J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2013-06-01

    Pacific Northwest National Laboratory (PNNL) has been requested by the U.S. Nuclear Regulatory Commission to evaluate the reactivity initiated accident (RIA) tests that have recently been performed in the Nuclear Safety Research Reactor (NSRR) and CABRI (French research reactor) on uranium dioxide (UO2) and mixed uranium and plutonium dioxide (MOX) fuels, and to propose pellet-cladding mechanical interaction (PCMI) failure thresholds for RIA events. This report discusses how PNNL developed PCMI failure thresholds for RIA based on least squares (LSQ) regression fits to the RIA test data from cold-worked stress relief annealed (CWSRA) and recrystallized annealed (RXA) cladding alloys under pressurized water reactor (PWR) hot zero power (HZP) conditions and boiling water reactor (BWR) cold zero power (CZP) conditions.

  8. MODELLING OF NUCLEAR FUEL CLADDING TUBES CORROSION

    Directory of Open Access Journals (Sweden)

    Miroslav Cech

    2016-12-01

    Full Text Available This paper describes materials made of zirconium-based alloys used for nuclear fuel cladding fabrication. It is focused on corrosion problems their theoretical description and modeling in nuclear engineering.

  9. Method of processing spent fuel cladding tubes

    International Nuclear Information System (INIS)

    Nakatsuka, Masafumi; Ouchi, Atsuhiro; Imahashi, Hiromichi.

    1986-01-01

    Purpose: To decrease the residual activity of spent fuel cladding tubes in a short period of time and enable safety storage with simple storage equipments. Constitution: Spent fuel cladding tubes made of zirconium alloys discharged from a nuclear fuel reprocessing step are exposed to a grain boundary embrittling atmosphere to cause grain boundary destruction. This causes grain boundary fractures to the zirconium crystal grains as the matrix of nuclear fuels and then precipitation products precipitated to the grain boundary fractures are removed. The zirconium constituting the nuclear fuel cladding tube and other ingredient elements contained in the precipitation products are separated in this removing step and they are separately stored respectively. As a result, zirconium constituting most part of the composition of the spent nuclear fuel cladding tubes can be stored safely at a low activity level. (Takahashi, M.)

  10. GSGG edge cladding development: Final technical report

    International Nuclear Information System (INIS)

    Izumitani, T.; Meissner, H.E.; Toratani, H.

    1986-01-01

    The objectives of this project have been: (1) Investigate the possibility of chemical etching of GSGG crystal slabs to obtain increased strength. (2) Design and construct a simplified mold assembly for casting cladding glass to the edges of crystal slabs of different dimensions. (3) Conduct casting experiments to evaluate the redesigned mold assembly and to determine stresses as function of thermal expansion coefficient of cladding glass. (4) Clad larger sizes of GGG slabs as they become available. These tasks have been achieved. Chemical etching of GSGG slabs does not appear possible with any other acid than H 3 PO 4 at temperatures above 300 0 C. A mold assembly has been constructed which allowed casting cladding glass around the edges of the largest GGG slabs available (10 x 20 x 160 mm) without causing breakage through the annealing step

  11. Duplex stainless steel surface bay laser cladding

    International Nuclear Information System (INIS)

    Amigo, V.; Pineda, Y.; Segovia, F.; Vicente, A.

    2004-01-01

    Laser cladding is one of the most promising techniques to restore damaged surfaces and achieve properties similar to those of the base metal. In this work, duplex stainless steels have been cladded by a nickel alloy under different processing conditions. The influence of the beam speed and defocusing variables ha been evaluated in the microstructure both of the cladding and heat affected zone, HAZ. These results have been correlated to mechanical properties by means of microhardness measurements from cladding area to base metal through the interface. This technique has shown to be very appropriate to obtain controlled mechanical properties as they are determined by the solidification microstructure, originated by the transfer of mass and heat in the system. (Author) 21 refs

  12. Corrosion behaviour of cladded nickel base alloys

    International Nuclear Information System (INIS)

    Brandl, W.; Ruczinski, D.; Nolde, M.; Blum, J.

    1995-01-01

    As a consequence of the high cost of nickel base alloys their use as surface layers is convenient. In this paper the properties of SA-as well as RES-cladded NiMo 16Cr16Ti and NiCr21Mo14W being produced in single and multi-layer technique are compared and discussed with respect to their corrosion behaviour. Decisive criteria describing the qualities of the claddings are the mass loss, the susceptibility against intergranular corrosion and the pitting corrosion resistance. The results prove that RES cladding is the most suitable technique to produce corrosion resistant nickel base coatings. The corrosion behaviour of a two-layer RES deposition shows a better resistance against pitting than a three layer SAW cladding. 7 refs

  13. Uranium enrichment

    International Nuclear Information System (INIS)

    1989-01-01

    GAO was asked to address several questions concerning a number of proposed uranium enrichment bills introduced during the 100th Congress. The bill would have restructured the Department of Energy's uranium enrichment program as a government corporation to allow it to compete more effectively in the domestic and international markets. Some of GAO's findings discussed are: uranium market experts believe and existing market models show that the proposed DOE purchase of a $750 million of uranium from domestic producers may not significantly increase production because of large producer-held inventories; excess uranium enrichment production capacity exists throughout the world; therefore, foreign producers are expected to compete heavily in the United States throughout the 1990s as utilities' contracts with DOE expire; and according to a 1988 agreement between DOE's Offices of Nuclear Energy and Defense Programs, enrichment decommissioning costs, estimated to total $3.6 billion for planning purposes, will be shared by the commercial enrichment program and the government

  14. Uranium resources

    International Nuclear Information System (INIS)

    1976-01-01

    This is a press release issued by the OECD on 9th March 1976. It is stated that the steep increases in demand for uranium foreseen in and beyond the 1980's, with doubling times of the order of six to seven years, will inevitably create formidable problems for the industry. Further substantial efforts will be needed in prospecting for new uranium reserves. Information is given in tabular or graphical form on the following: reasonably assured resources, country by country; uranium production capacities, country by country; world nuclear power growth; world annual uranium requirements; world annual separative requirements; world annual light water reactor fuel reprocessing requirements; distribution of reactor types (LWR, SGHWR, AGR, HWR, HJR, GG, FBR); and world fuel cycle capital requirements. The information is based on the latest report on Uranium Resources Production and Demand, jointly issued by the OECD's Nuclear Energy Agency (NEA) and the International Atomic Energy Agency. (U.K.)

  15. Optimization of metal-clad waveguide sensors

    DEFF Research Database (Denmark)

    Skivesen, N.; Horvath, R.; Pedersen, H.C.

    2005-01-01

    The present paper deals with the optimization of metal-clad waveguides for sensor applications to achieve high sensitivity for adlayer and refractive index measurements. By using the Fresnel reflection coefficients both the angular shift and the width of the resonances in the sensorgrams are taken...... into account. Our optimization shows that it is possible for metal-clad waveguides to achieve a sensitivity improvement of 600% compared to surface-plasmon-resonance sensors....

  16. Uranium supply and demand

    Energy Technology Data Exchange (ETDEWEB)

    Spriggs, M J

    1976-01-01

    Papers were presented on the pattern of uranium production in South Africa; Australian uranium--will it ever become available; North American uranium resources, policies, prospects, and pricing; economic and political environment of the uranium mining industry; alternative sources of uranium supply; whither North American demand for uranium; and uranium demand and security of supply--a consumer's point of view. (LK)

  17. Laser cladding technology to small diameter pipes

    International Nuclear Information System (INIS)

    Fujimagari, H.; Hagiwara, M.; Kojima, T.

    2000-01-01

    A laser cladding method which produces a highly corrosion-resistant material coating layers (cladding) on the austenitic stainless steel (type 304 SS) pipe inner surface was developed to prevent SCC (stress corrosion cracking) occurrence. This technology is applicable to a narrow and long distance area from operators, because of the good accessibility of the YAG (yttrium-aluminum-garnet) laser beam that can be transmitted through an optical fiber. In this method a mixed paste metallic powder and heating-resistive organic solvent are firstly placed on the inner surface of a small pipe, and then a YAG laser beam transmitted through an optical fiber irradiates to the pasted area. A mixed paste will be melted and form a cladding layer subsequently. A cladding layer shows as excellent corrosion resistance property. This laser cladding (LC) method had already applied to several domestic nuclear power plants and had obtained a good reputation. This report introduces the outline of laser cladding technology, the developed equipment for practical application in the field, and the circumstance in actual plant application. (orig.)

  18. Clad Degradation- Summary and Abstraction for LA

    International Nuclear Information System (INIS)

    D. Stahl

    2004-01-01

    The purpose of this model report is to develop the summary cladding degradation abstraction that will be used in the Total System Performance Assessment for the License Application (TSPA-LA). Most civilian commercial nuclear fuel is encased in Zircaloy cladding. The model addressed in this report is intended to describe the postulated condition of commercial Zircaloy-clad fuel as a function of postclosure time after it is placed in the repository. Earlier total system performance assessments analyzed the waste form as exposed UO 2 , which was available for degradation at the intrinsic dissolution rate. Water in the waste package quickly became saturated with many of the radionuclides, limiting their release rate. In the total system performance assessments for the Viability Assessment and the Site Recommendation, cladding was analyzed as part of the waste form, limiting the amount of fuel available at any time for degradation. The current model is divided into two stages. The first considers predisposal rod failures (most of which occur during reactor operation and associated activities) and postdisposal mechanical failure (from static loading of rocks) as mechanisms for perforating the cladding. Other fuel failure mechanisms including those caused by handling or transportation have been screened out (excluded) or are treated elsewhere. All stainless-steel-clad fuel, which makes up a small percentage of the overall amount of fuel to be stored, is modeled as failed upon placement in the waste packages. The second stage of the degradation model is the splitting of the cladding from the reaction of water or moist air and UO 2 . The splitting has been observed to be rapid in comparison to the total system performance assessment time steps and is modeled to be instantaneous. After the cladding splits, the rind buildup inside the cladding widens the split, increasing the diffusion area from the fuel rind to the waste package interior. This model report summarizes the

  19. Evaluation of structural integrity of IPNS-I and ZING-P' targets

    International Nuclear Information System (INIS)

    Carpenter, J.; Ahmed, H.; Loomis, B.; Ball, J.; Ewing, T.; Bailey, J.; D'Souza, A.F.

    1982-12-01

    This report discusses the design, production, and evaluation of clad uranium-alloy targets that function as spallation neutron sources in the ZING-P' and IPNS-I facilities with a pulsed (10 to 30 Hz), 500-MeV proton beam. The methodology and results of theoretical nuclear-particle transport, heat transport, and stress analyses that were used in the development of a design for the targets are described. The production of a zirconium-clad uranium-alloy cylinder for ZING-P' and Zircaloy-2-clad uranium-alloy discs for IPNS-I is discussed with particular attention to the procedural details. The theoretical analyses were verified by measuring the thermal and mechanical response of the clad uranium under conditions designed to simulate the operations of the pulsed-neutron sources

  20. Evaluation of structural integrity of IPNS-I and ZING-P' targets

    Energy Technology Data Exchange (ETDEWEB)

    Carpenter, J.; Ahmed, H.; Loomis, B.; Ball, J.; Ewing, T.; Bailey, J.; D' Souza, A.F.

    1982-12-01

    This report discusses the design, production, and evaluation of clad uranium-alloy targets that function as spallation neutron sources in the ZING-P' and IPNS-I facilities with a pulsed (10 to 30 Hz), 500-MeV proton beam. The methodology and results of theoretical nuclear-particle transport, heat transport, and stress analyses that were used in the development of a design for the targets are described. The production of a zirconium-clad uranium-alloy cylinder for ZING-P' and Zircaloy-2-clad uranium-alloy discs for IPNS-I is discussed with particular attention to the procedural details. The theoretical analyses were verified by measuring the thermal and mechanical response of the clad uranium under conditions designed to simulate the operations of the pulsed-neutron sources.

  1. Uranium, depleted uranium, biological effects; Uranium, uranium appauvri, effets biologiques

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-07-01

    Physicists, chemists and biologists at the CEA are developing scientific programs on the properties and uses of ionizing radiation. Since the CEA was created in 1945, a great deal of research has been carried out on the properties of natural, enriched and depleted uranium in cooperation with university laboratories and CNRS. There is a great deal of available data about uranium; thousands of analyses have been published in international reviews over more than 40 years. This presentation on uranium is a very brief summary of all these studies. (author)

  2. Uranium toxicology

    International Nuclear Information System (INIS)

    Ferreyra, Mariana D.; Suarez Mendez, Sebastian

    1997-01-01

    In this paper are presented the methods and procedures optimized by the Nuclear Regulatory Authority (ARN) for the determination of: natural uranium mass, activity of enriched uranium in samples of: urine, mucus, filters, filter heads, rinsing waters and Pu in urine, adopted and in some cases adapted, by the Environmental Monitoring and Internal Dosimetry Laboratory. The analyzed material corresponded to biological and environmental samples belonging to the staff professionally exposed that work in plants of the nuclear fuel cycle. For a better comprehension of the activities of this laboratory, it is included a brief description of the uranium radiochemical toxicity and the limits internationally fixed to preserve the workers health

  3. Simulation of uranium aluminide dissolution in a continuous aluminum dissolver system

    International Nuclear Information System (INIS)

    Evans, D.R.; Farman, R.F.; Christian, J.D.

    1990-01-01

    This paper reports on the Idaho Chemical Processing Plant (ICPP) which recovers highly-enriched uranium (uranium that contains at least 20 atom percent 235 U) from spent nuclear reactor fuel by dissolution of the fuel elements and extraction of the uranium from the aqueous dissolver product. Because the uranium is highly-enriched, consideration must be given to whether a critical mass can form at any stage of the process. In particular, suspended 235 U-containing particles are of special concern, due to their high density (6.8 g/cm 3 ) and due to the fact that they can settle into geometrically unfavorable configurations when not adequately mixed. A portion of the spent fuel is aluminum-alloy-clad uranium aluminide (UAl 3 ) particles, which dissolve more slowly than the cladding. As the aluminum alloy cladding dissolves in mercury-catalyzed nitric acid, UAl 3 is released. Under standard operating conditions, the UAl 3 dissolves rapidly enough to preclude the possibility of forming a critical mass anywhere in the system. However, postulated worst-case abnormal operating conditions retard uranium aluminide dissolution, and thus require evaluation. To establish safety limits for operating parameters, a computerized simulation model of uranium aluminide dissolution in the aluminum fuel dissolver system was developed

  4. Rossing uranium

    International Nuclear Information System (INIS)

    Anon.

    1979-01-01

    In this article the geology of the deposits of the Rossing uranium mine in Namibia is discussed. The planning of the open-pit mining, the blasting, drilling, handling and the equipment used for these processes are described

  5. Accident tolerant fuel cladding development: Promise, status, and challenges

    Science.gov (United States)

    Terrani, Kurt A.

    2018-04-01

    The motivation for transitioning away from zirconium-based fuel cladding in light water reactors to significantly more oxidation-resistant materials, thereby enhancing safety margins during severe accidents, is laid out. A review of the development status for three accident tolerant fuel cladding technologies, namely coated zirconium-based cladding, ferritic alumina-forming alloy cladding, and silicon carbide fiber-reinforced silicon carbide matrix composite cladding, is offered. Technical challenges and data gaps for each of these cladding technologies are highlighted. Full development towards commercial deployment of these technologies is identified as a high priority for the nuclear industry.

  6. Fabrication and Testing of Prototype APM-Clad UO{sub 2} Fuel Elements; Fabrication et essai de prototypes de cartouches de combustible en bioxyde d'uranium gaine d'aluminium (APM); Izgotovlenie i ispytanie prototipa toplivnykh ehlementov na osnove UO{sub 2} s obolochkoj iz alyuminiya metodom poroshkovoj metallurgii; Elaboracion y ensayo de elementos combustibles prototipo de UO{sub 2} con revestimiento de aluminio sinterizado

    Energy Technology Data Exchange (ETDEWEB)

    Ballif, III, J. L.; Friske, W. H.; Gordon, R. B. [Atomics International, Canoga Park, California (United States)

    1963-11-15

    In support of the 50-MW(e) Prototype Organic Power Reactor Programme (POPR), extensive development work has been performed on aluminium powder metallurgy (ARM) products, toward their use as cladding for UO{sub 2} fuel. As part of this development work, eutectic bonding, flash butt welding, and cold-pressure welding were investigated as methods for making end closures in die fuel element cladding. Vibratory packing was studied as a means of filling APM tubes with UO{sub 2}. Out-of-pile tests were conducted to obtain information on APM-UO{sub 2} compatibility. This work revealed that, under present conditions, eutectic bonding was the most suitable method for making end closures; vibratory packing produced fuel densities in the range of 80 to 88% of theoretical density; and no APM-UO{sub 2} reaction took place in the range of POPR operating temperatures (850{sup o}F maximum fuel-cladding interface temperature). As a result o f this development work, five APM-clad UO{sub 2} prototype fuel elements have been fabricated for testing in the Organic Moderated Reactor Experiment (OMRE). Each element consisted of 24 or 25 APM-clad fuel rods, arranged in a 5 x 5 array in a nickel-plated steel or an APM fuel box. To increase surface area, the extruded APM cladding had eight fins which were spiralled to a pitch of 45 or 90e/ ft to further improve heat transfer. The fuel rod end closures were made by eutectic bonding of silver-plated aluminium end plugs to the APM tubing. The elements were instrumented to: (1) Measure cladding surface and coolant temperatures, (2) Detect fuel rod failure, (3) Change coolant velocity (means of achieving peak cladding surface temperature of 850{sup o}F), (4) Measure coolant velocity, and (5) Measure fission gas build-up. These elements have been installed in the OMRE with target fuel burn-ups of 25000 to 30000 MWd/t of uranium. As of 1 April 1963, they had achieved accumulated burn-ups ranging from 7700 to 12 000 MWd/t of uranium. Two of the

  7. Uranium, depleted uranium, biological effects

    International Nuclear Information System (INIS)

    2001-01-01

    Physicists, chemists and biologists at the CEA are developing scientific programs on the properties and uses of ionizing radiation. Since the CEA was created in 1945, a great deal of research has been carried out on the properties of natural, enriched and depleted uranium in cooperation with university laboratories and CNRS. There is a great deal of available data about uranium; thousands of analyses have been published in international reviews over more than 40 years. This presentation on uranium is a very brief summary of all these studies. (author)

  8. Uranium loans

    International Nuclear Information System (INIS)

    Anon.

    1990-01-01

    When NUEXCO was organized in 1968, its founders conceived of a business based on uranium loans. The concept was relatively straightforward; those who found themselves with excess supplies of uranium would deposit those excesses in NUEXCO's open-quotes bank,close quotes and those who found themselves temporarily short of uranium could borrow from the bank. The borrower would pay interest based on the quantity of uranium borrowed and the duration of the loan, and the bank would collect the interest, deduct its service fee for arranging the loan, and pay the balance to those whose deposits were borrowed. In fact, the original plan was to call the firm Nuclear Bank Corporation, until it was discovered that using the word open-quotes Bankclose quotes in the name would subject the firm to various US banking regulations. Thus, Nuclear Bank Corporation became Nuclear Exchange Corporation, which was later shortened to NUEXCO. Neither the nuclear fuel market nor NUEXCO's business developed quite as its founders had anticipated. From almost the very beginning, the brokerage of uranium purchases and sales became a more significant activity for NUEXCO than arranging uranium loans. Nevertheless, loan transactions have played an important role in the international nuclear fuel market, requiring the development of special knowledge and commercial techniques

  9. A pellet-clad interaction failure criterion

    International Nuclear Information System (INIS)

    Howl, D.A.; Coucill, D.N.; Marechal, A.J.C.

    1983-01-01

    A Pellet-Clad Interaction (PCI) failure criterion, enabling the number of fuel rod failures in a reactor core to be determined for a variety of normal and fault conditions, is required for safety analysis. The criterion currently being used for the safety analysis of the Pressurized Water Reactor planned for Sizewell in the UK is defined and justified in this paper. The criterion is based upon a threshold clad stress which diminishes with increasing fast neutron dose. This concept is consistent with the mechanism of clad failure being stress corrosion cracking (SCC); providing excess corrodant is always present, the dominant parameter determining the propagation of SCC defects is stress. In applying the criterion, the SLEUTH-SEER 77 fuel performance computer code is used to calculate the peak clad stress, allowing for concentrations due to pellet hourglassing and the effect of radial cracks in the fuel. The method has been validated by analysis of PCI failures in various in-reactor experiments, particularly in the well-characterised power ramp tests in the Steam Generating Heavy Water Reactor (SGHWR) at Winfrith. It is also in accord with out-of-reactor tests with iodine and irradiated Zircaloy clad, such as those carried out at Kjeller in Norway. (author)

  10. Fuel assembly and fuel cladding tube

    International Nuclear Information System (INIS)

    Tsutsumi, Shinro; Ito, Ken-ichi; Inagaki, Masatoshi; Nakajima, Junjiro.

    1996-01-01

    A fuel cladding tube is a zirconium liner tube formed by lining a pure zirconium layer on the inner side of a zirconium alloy tube. The fuel cladding tube is formed by extrusion molding of a composite billet formed by inserting a pure zirconium billet into a zirconium alloy billet. Accordingly, the pure zirconium layer and the zirconium alloy tube are strongly joined by metal bond. The fuel cladding tube has an external oxide film on the outer surface of the zirconium alloy tube and an internal oxide film on the inner side of the pure zirconium layer. The external oxide film has a thickness preferably of about 1μm. The internal oxide film has a thickness of not more than 10μm, preferably, from 1 to 5μm. With such a constitution, flaws to be formed on both inner and outer surfaces of the cladding tube upon assembling a fuel assembly can be reduced thereby enabling to reduce the amount of hydrogen absorbed to the cladding tube. (I.N.)

  11. Clad buffer rod sensors for liquid metals

    International Nuclear Information System (INIS)

    Jen, C.-K.; Ihara, I.

    1999-01-01

    Clad buffer rods, consisting of a core and a cladding, have been developed for ultrasonic monitoring of liquid metal processing. The cores of these rods are made of low ultrasonic-loss materials and the claddings are fabricated by thermal spray techniques. The clad geometry ensures proper ultrasonic guidance. The lengths of these rods ranges from tens of centimeters to 1m. On-line ultrasonic level measurements in liquid metals such as magnesium at 700 deg C and aluminum at 960 deg C are presented to demonstrate their operation at high temperature and their high ultrasonic performance. A spherical concave lens is machined at the rod end for improving the spatial resolution. High quality ultrasonic images have been obtained in the liquid zinc at 600 deg C. High spatial resolution is needed for the detection of inclusions in liquid metals during processing. We also show that the elastic properties such as density, longitudinal and shear wave velocities of liquid metals can be measured using a transducer which generates and receives both longitudinal and shear waves and is mounted at the end of a clad buffer rod. (author)

  12. Uranium extraction from gold-uranium ores

    Energy Technology Data Exchange (ETDEWEB)

    Laskorin, B.N.; Golynko, Z.Sh.

    1981-01-01

    The process of uranium extraction from gold-uranium ores in the South Africa is considered. Flowsheets of reprocessing gold-uranium conglomerates, pile processing and uranium extraction from the ores are presented. Continuous counter flow ion-exchange process of uranium extraction using strong-active or weak-active resins is noted to be the most perspective and economical one. The ion-exchange uranium separation with the succeeding extraction is also the perspective one.

  13. 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

  14. Capturing reflected cladding modes from a fiber Bragg grating with a double-clad fiber coupler.

    Science.gov (United States)

    Baiad, Mohamad Diaa; Gagné, Mathieu; Lemire-Renaud, Simon; De Montigny, Etienne; Madore, Wendy-Julie; Godbout, Nicolas; Boudoux, Caroline; Kashyap, Raman

    2013-03-25

    We present a novel measurement scheme using a double-clad fiber coupler (DCFC) and a fiber Bragg grating (FBG) to resolve cladding modes. Direct measurement of the optical spectra and power in the cladding modes is obtained through the use of a specially designed DCFC spliced to a highly reflective FBG written into slightly etched standard photosensitive single mode fiber to match the inner cladding diameter of the DCFC. The DCFC is made by tapering and fusing two double-clad fibers (DCF) together. The device is capable of capturing backward propagating low and high order cladding modes simply and efficiently. Also, we demonstrate the capability of such a device to measure the surrounding refractive index (SRI) with an extremely high sensitivity of 69.769 ± 0.035 μW/RIU and a resolution of 1.433 × 10(-5) ± 8 × 10(-9) RIU between 1.37 and 1.45 RIU. The device provides a large SRI operating range from 1.30 to 1.45 RIU with sufficient discrimination for all individual captured cladding modes. The proposed scheme can be adapted to many different types of bend, temperature, refractive index and other evanescent wave based sensors.

  15. Inspection system for Zircaloy clad fuel rods

    International Nuclear Information System (INIS)

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

    1975-10-01

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

  16. Fuel cladding mechanical properties for transient analysis

    International Nuclear Information System (INIS)

    Johnson, G.D.; Hunter, C.W.; Hanson, J.E.

    1976-01-01

    Out-of-pile simulated transient tests have been conducted on irradiated fast-reactor fuel pin cladding specimens at heating rates of 10 0 F/s (5.6 0 K/s) and 200 0 F/s (111 0 K/s) to generate mechanical property information for use in describing cladding behavior during off-normal events. Mechanical property data were then analyzed, applying the Larson-Miller Parameter to the effects of heating rate and neutron fluence. Data from simulated transient tests on TREAT-tested fuel pins demonstrate that Plant Protective System termination of 3$/s transients prevents significant damage to cladding. The breach opening produced during simulated transient testing is shown to decrease in size with increasing neutron fluence

  17. Laser cladding to select new glassy alloys

    International Nuclear Information System (INIS)

    Medrano, L.L.O.; Afonso, C.R.M.; Kiminami, C.S.; Gargarella, P.; Ramasco, B.

    2016-01-01

    A new experimental technique used to analyze the effect of compositional variation and cooling rate in the phase formation in a multicomponent system is the laser cladding. This work have evaluated the use of laser cladding to discover a new bulk metallic glass (BMG) in the Al-Co-Zr system. Coatings with composition variation have made by laser cladding using Al-Co-Zr alloys powders and the samples produced have been characterized by X ray diffraction, microscopy and energy-dispersive X-ray spectroscopy. The results did not show the composition variation as expected, because of incomplete melting during laser process. It was measured a composition variation tendency that allowed the glass forming investigation by the glass formation criterion λ+Δh 1/2 . The results have showed no glass formation in the coating samples, which prove a limited capacity of Zr-Co-Al system to form glass (author)

  18. Potential effects of gallium on cladding materials

    International Nuclear Information System (INIS)

    Wilson, D.F.; Beahm, E.C.; Besmann, T.M.; DeVan, J.H.; DiStefano, J.R.; Gat, U.; Greene, S.R.; Rittenhouse, P.L.; Worley, B.A.

    1997-10-01

    This paper identifies and examines issues concerning the incorporation of gallium in weapons derived plutonium in light water reactor (LWR) MOX fuels. Particular attention is given to the more likely effects of the gallium on the behavior of the cladding material. The chemistry of weapons grade (WG) MOX, including possible consequences of gallium within plutonium agglomerates, was assessed. Based on the calculated oxidation potentials of MOX fuel, the effect that gallium may have on reactions involving fission products and possible impact on cladding performance were postulated. Gallium transport mechanisms are discussed. With an understanding of oxidation potentials and assumptions of mechanisms for gallium transport, possible effects of gallium on corrosion of cladding were evaluated. Potential and unresolved issues and suggested research and development (R and D) required to provide missing information are presented

  19. Uranium mining

    International Nuclear Information System (INIS)

    2008-01-01

    Full text: The economic and environmental sustainability of uranium mining has been analysed by Monash University researcher Dr Gavin Mudd in a paper that challenges the perception that uranium mining is an 'infinite quality source' that provides solutions to the world's demand for energy. Dr Mudd says information on the uranium industry touted by politicians and mining companies is not necessarily inaccurate, but it does not tell the whole story, being often just an average snapshot of the costs of uranium mining today without reflecting the escalating costs associated with the process in years to come. 'From a sustainability perspective, it is critical to evaluate accurately the true lifecycle costs of all forms of electricity production, especially with respect to greenhouse emissions, ' he says. 'For nuclear power, a significant proportion of greenhouse emissions are derived from the fuel supply, including uranium mining, milling, enrichment and fuel manufacture.' Dr Mudd found that financial and environmental costs escalate dramatically as the uranium ore is used. The deeper the mining process required to extract the ore, the higher the cost for mining companies, the greater the impact on the environment and the more resources needed to obtain the product. I t is clear that there is a strong sensitivity of energy and water consumption and greenhouse emissions to ore grade, and that ore grades are likely to continue to decline gradually in the medium to long term. These issues are critical to the current debate over nuclear power and greenhouse emissions, especially with respect to ascribing sustainability to such activities as uranium mining and milling. For example, mining at Roxby Downs is responsible for the emission of over one million tonnes of greenhouse gases per year and this could increase to four million tonnes if the mine is expanded.'

  20. Filtration and Leach Testing for PUREX Cladding Sludge and REDOX Cladding Sludge Actual Waste Sample Composites

    Energy Technology Data Exchange (ETDEWEB)

    Shimskey, Rick W.; Billing, Justin M.; Buck, Edgar C.; Casella, Amanda J.; Crum, Jarrod V.; Daniel, Richard C.; Draper, Kathryn E.; Edwards, Matthew K.; Hallen, Richard T.; Kozelisky, Anne E.; MacFarlan, Paul J.; Peterson, Reid A.; Swoboda, Robert G.

    2009-03-02

    A testing program evaluating actual tank waste was developed in response to Task 4 from the M-12 External Flowsheet Review Team (EFRT) issue response plan (Barnes and Voke 2006). The test program was subdivided into logical increments. The bulk water-insoluble solid wastes that are anticipated to be delivered to the Hanford Waste Treatment and Immobilization Plant (WTP) were identified according to type such that the actual waste testing could be targeted to the relevant categories. Under test plan TP RPP WTP 467 (Fiskum et al. 2007), eight broad waste groupings were defined. Samples available from the 222S archive were identified and obtained for testing. Under this test plan, a waste testing program was implemented that included: • Homogenizing the archive samples by group as defined in the test plan. • Characterizing the homogenized sample groups. • Performing parametric leaching testing on each group for compounds of interest. • Performing bench-top filtration/leaching tests in the hot cell for each group to simulate filtration and leaching activities if they occurred in the UFP2 vessel of the WTP Pretreatment Facility. This report focuses on a filtration/leaching test performed using two of the eight waste composite samples. The sample groups examined in this report were the plutonium-uranium extraction (PUREX) cladding waste sludge (Group 3, or CWP) and reduction-oxidation (REDOX) cladding waste sludge (Group 4, or CWR). Both the Group 3 and 4 waste composites were anticipated to be high in gibbsite, thus requiring caustic leaching. WTP RPT 167 (Snow et al. 2008) describes the homogenization, characterization, and parametric leaching activities before benchtop filtration/leaching testing of these two waste groups. Characterization and initial parametric data in that report were used to plan a single filtration/leaching test using a blend of both wastes. The test focused on filtration testing of the waste and caustic leaching for aluminum, in the form

  1. Comparison of corrosion behavior between fusion cladded and explosive cladded Inconel 625/plain carbon steel bimetal plates

    International Nuclear Information System (INIS)

    Zareie Rajani, H.R.; Akbari Mousavi, S.A.A.; Madani Sani, F.

    2013-01-01

    Highlights: ► Both explosive and fusion cladding aggravate the corrosion resistance of Inconel 625. ► Fusion cladding is more detrimental to nonuniform corrosion resistance. ► Single-layered fusion coat does not show any repassivation ability. ► Adding more layers enhance the corrosion resistance of fusion cladding Inconel 625. ► High impact energy spoils the corrosion resistance of explosive cladding Inconel 625. -- Abstract: One of the main concerns in cladding Inconel 625 superalloy on desired substrates is deterioration of corrosion resistance due to cladding process. The present study aims to compare the effect of fusion cladding and explosive cladding procedures on corrosion behavior of Inconel 625 cladding on plain carbon steel as substrate. Also, an attempt has been made to investigate the role of load ratio and numbers of fusion layers in corrosion behavior of explosive and fusion cladding Inconel 625 respectively. In all cases, the cyclic polarization as an electrochemical method has been applied to assess the corrosion behavior. According to the obtained results, both cladding methods aggravate the corrosion resistance of Inconel 625. However, the fusion cladding process is more detrimental to nonuniform corrosion resistance, where the chemical nonuniformity of fusion cladding superalloy issuing from microsegregation, development of secondary phases and contamination of clad through dilution hinders formation of a stable passive layer. Moreover, it is observed that adding more fusion layers can enhance the nonuniform corrosion resistance of fusion cladding Inconel 625, though this resistance still remains weaker than explosive cladding superalloy. Also, the results indicate that raising the impact energy in explosive cladding procedure drops the corrosion resistance of Inconel 625.

  2. Modelling of pellet-clad interaction during power ramps

    International Nuclear Information System (INIS)

    Zhou, G.; Lindback, J.E.; Schutte, H.C.; Jernkvist, L.O.; Massih, A.R.; Massih, A.R.

    2005-01-01

    A computational method to describe the pellet-clad interaction phenomenon is presented. The method accounts for the mechanical contact between fragmented pellets and the zircaloy clad, as well as for chemical reaction of fission products with zircaloy during power ramps. Possible pellet-clad contact states, soft, hard and friction, are taken into account in the computational algorithm. The clad is treated as an elastic-plastic-viscoplastic material with irradiation hardening. Iodine-induced stress corrosion cracking is described by using a fracture mechanics-based model for crack propagation. This integrated approach is used to evaluate two power ramp experiments made on boiling water reactor fuel rods in test reactors. The influence of the pellet-clad coefficient of friction on clad deformation is evaluated and discussed. Also, clad deformations, pellet-clad gap size and fission product gas release for one of the ramped rods are calculated and compared with measured data. (authors)

  3. Multilayer cladding with hyperbolic dispersion for plasmonic waveguides

    DEFF Research Database (Denmark)

    Babicheva, Viktoriia; Shalaginov, Mikhail Y.; Ishii, Satoshi

    2015-01-01

    We study the properties of plasmonic waveguides with a dielectric core and multilayer metal-dielectric claddings that possess hyperbolic dispersion. The waveguides hyperbolic multilayer claddings show better performance in comparison to conventional plasmonic waveguides. © OSA 2015....

  4. Cladding using a 15 kW CO2 laser

    International Nuclear Information System (INIS)

    Vesely, E.J.; Verma, S.K.

    1989-01-01

    Laser alloying or cladding differs little in principle from the traditional forms of weld overlays, but lasers as a heat source offer some distinct advantages. With the selective heating attainable using high power lasers, good metallurgical bond of the clad layer, minimal dilution and typically, a very fine homogeneous microstructure can be obtained in the clad layer. This is a review of work in laser cladding using the 15 kW CO 2 laser. The authors discuss the ability of the laser clad surface to increase the high temperature oxidation resistance of a low-alloy carbon steel (4140). Examples of clads subjected to high- temperature thermal cycling of nickel-20% aluminum and TaC + 4140 clad low-alloy steel and straight high-temperature oxidation of Stellite 6-304L cladding on a 4140 substrate are given

  5. Evaluation of fast experimental reactor claddings, (2)

    International Nuclear Information System (INIS)

    Miura, Makoto; Nagaki, Hiroshi; Koyama, Masahiro; Tanaka, Yasumasa

    1974-01-01

    Thin-walled fine tubes of Type 316 austenitic stainless steel are used for fuel cladding in Joyo (experimental FBR). The material exhibits the change of the mechanical properties in long-time annealing at high temperature, resulting from the precipitation of carbide in structure. In this connection, the experiment and the results on the changes of the microstructure and mechanical properties (proof stress and hardness) are described. The test specimens are the fuel cladding tubes produced for trial for Joyo core and those for FFTF core made in the U.S.A. They were heated between 400 0 and 850 0 C for 1000 hr in vacuum. (Mori, K.)

  6. Management of cladding hulls and fuel hardware

    International Nuclear Information System (INIS)

    1985-01-01

    The reprocessing of spent fuel from power reactors based on chop-leach technology produces a solid waste product of cladding hulls and other metallic residues. This report describes the current situation in the management of fuel cladding hulls and hardware. Information is presented on the material composition of such waste together with the heating effects due to neutron-induced activation products and fuel contamination. As no country has established a final disposal route and the corresponding repository, this report also discusses possible disposal routes and various disposal options under consideration at present

  7. Inpile (in PWR) testing of cladding materials

    International Nuclear Information System (INIS)

    Hahn, R.; Jeong, Y. H.; Baek, B. J.; Kim, K. H.; Kim, S. J.; Choi, B. K.; Kim, J. M.

    1999-04-01

    As an introduction, the reasons to perform inpile tests are depicted. An overview over general inpile test procedure is given, and test details which are necessary for the development of new alloys for high burnup claddings, like sample geometries and measuring techniques for inpile corrosion testing, are described in detail. Tests for the creep and length change behavior of cladding tubes are described briefly. Finally, conclusions are drawn and literature citations for further test details are given. (author). 9 refs., 2 tabs., 17 figs

  8. Microstructure of laser cladded martensitic stainless steel

    CSIR Research Space (South Africa)

    Van Rooyen, C

    2006-08-01

    Full Text Available and martensite with 10% ferrite for Material B. Table 7 - Proposed martensitic stainless steel alloys for laser cladding Material C* Cr Ni Mn Si Mo Co Ms (ºC)* Cr eq Ni eq Material A 0.4 13 - 1 0.5 2.5 5.5 120 16.5 12.5 Material B 0.2 15 2 1 0.7 2.5 5.5 117... dilution, low heat input, less distortion, increased mechanical and corrosion properties excellent repeatability and control of process parameters. Solidification of laser cladded martensitic stainless steel is primarily austenitic. Microstructures...

  9. Phosphate-core silica-clad Er/Yb-doped optical fiber and cladding pumped laser.

    Science.gov (United States)

    Egorova, O N; Semjonov, S L; Velmiskin, V V; Yatsenko, Yu P; Sverchkov, S E; Galagan, B I; Denker, B I; Dianov, E M

    2014-04-07

    We present a composite optical fiber with a Er/Yb co-doped phosphate-glass core in a silica glass cladding as well as cladding pumped laser. The fabrication process, optical properties, and lasing parameters are described. The slope efficiency under 980 nm cladding pumping reached 39% with respect to the absorbed pump power and 28% with respect to the coupled pump power. Due to high doping level of the phosphate core optimal length was several times shorter than that of silica core fibers.

  10. Uranium enrichment

    International Nuclear Information System (INIS)

    Rae, H.K.; Melvin, J.G.

    1988-06-01

    Canada is the world's largest producer and exporter of uranium, most of which is enriched elsewhere for use as fuel in LWRs. The feasibility of a Canadian uranium-enrichment enterprise is therefore a perennial question. Recent developments in uranium-enrichment technology, and their likely impacts on separative work supply and demand, suggest an opportunity window for Canadian entry into this international market. The Canadian opportunity results from three particular impacts of the new technologies: 1) the bulk of the world's uranium-enrichment capacity is in gaseous diffusion plants which, because of their large requirements for electricity (more than 2000 kW·h per SWU), are vulnerable to competition from the new processes; 2) the decline in enrichment costs increases the economic incentive for the use of slightly-enriched uranium (SEU) fuel in CANDU reactors, thus creating a potential Canadian market; and 3) the new processes allow economic operation on a much smaller scale, which drastically reduces the investment required for market entry and is comparable with the potential Canadian SEU requirement. The opportunity is not open-ended. By the end of the century the enrichment supply industry will have adapted to the new processes and long-term customer/supplier relationships will have been established. In order to seize the opportunity, Canada must become a credible supplier during this century

  11. Mechanical modelling of transient- to- failure SFR fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Feria, F.; Herranz, L. E.

    2014-07-01

    The response of Sodium Fast Reactor (SFR) fuel rods to transient accident conditions is an important safety concern. During transients the cladding strain caused by the stress due to pellet cladding mechanical interaction (PCMI) can lead to failure. Due to the fact that SFR fuel rods are commonly clad with strengthened material made of stainless steel (SS), cladding is usually treated as an elastic-perfectly-plastic material. However, viscoplastic behaviour can contribute to mechanical strain at high temperature (> 1000 K). (Author)

  12. Analysis of coaxial laser micro cladding processing conditions

    OpenAIRE

    Tarasova, Tatiana Vasilievna; Gvozdeva, Galina Olegovna; Nowotny, Steffen; Ableyeva, Riana R.; Dolzhikova, Evgenia Yu

    2018-01-01

    The laser build-up cladding is a well-known technique for repair, coatings and additive manufacturing tasks. Modern equipment for the laser cladding enables material to be deposited with the lateral resolution of about 100 μm and to manufacture miniature precise parts. However, the micro cladding regimes are unknown. Determination of these regimes is an expensive task as a well-known relation between laser cladding parameters and melt pool dimensions are changing by technology micro-miniaturi...

  13. Electron beam cladding of titanium on stainless steel plate

    International Nuclear Information System (INIS)

    Tomie, Michio; Abe, Nobuyuki; Yamada, Masanori; Noguchi, Shuichi.

    1990-01-01

    Fundamental characteristics of electron beam cladding was investigated. Titanium foil of 0.2mm thickness was cladded on stainless steel plate of 3mm thickness by scanning electron beam. Surface roughness and cladded layer were analyzed by surface roughness tester, microscope, scanning electron microscope and electron probe micro analyzer. Electron beam conditions were discussed for these fundamental characteristics. It is found that the energy density of the electron beam is one of the most important factor for cladding. (author)

  14. Polarization effects in silicon-clad optical waveguides

    Science.gov (United States)

    Carson, R. F.; Batchman, T. E.

    1984-01-01

    By changing the thickness of a semiconductor cladding layer deposited on a planar dielectric waveguide, the TE or TM propagating modes may be selectively attenuated. This polarization effect is due to the periodic coupling between the lossless propagating modes of the dielectric slab waveguide and the lossy modes of the cladding layer. Experimental tests involving silicon claddings show high selectivity for either polarization.

  15. The SLOWPOKE-2 reactor with low enrichment uranium oxide fuel

    International Nuclear Information System (INIS)

    Townes, B.M.; Hilborn, J.W.

    1985-06-01

    A SLOWPOKE-2 reactor core contains less than 1 kg of highly enriched uranium (HEU) and the proliferation risk is very low. However, to overcome proliferation concerns a new low enrichment uranium (LEU) fuelled reactor core has been designed. This core contains approximately 180 fuel elements based on the Zircaloy-4 clad UOsub(2) CANDU fuel element, but with a smaller outside diameter. The physics characteristics of this new reactor core ensure the inherent safety of the reactor under all conceivable conditions and thus the basic SLOWPOKE safety philosophy which permits unattended operation is not affected

  16. Uranium update

    International Nuclear Information System (INIS)

    Steane, R.

    1997-01-01

    This paper is about the current uranium mining situation, especially that in Saskatchewan. Canada has a unique advantage with the Saskatchewan uranium deposits. Making the most of this opportunity is important to Canada. The following is reviewed: project development and the time and capital it takes to bring a new project into production; the supply and demand situation to show where the future production fits into the world market; and our foreign competition and how we have to be careful not to lose our opportunity. (author)

  17. Experimental assessment of fuel-cladding interactions

    Energy Technology Data Exchange (ETDEWEB)

    Wood, Elizabeth Sooby [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-06-29

    A range of fuel concepts designed to better tolerate accident scenarios and reactor transients are currently undergoing fundamental development at national laboratories as well as university and industrial partners. Pellet-clad mechanical and chemical interaction can be expected to affect fuel failure rates experienced during steady state operation, as well as dramatically impact the response of the fuel form under loss of coolant and other accident scenarios. The importance of this aspect of fuel design prompted research initiated by AFC in FY14 to begin exploratory efforts to characterize this phenomenon for candidate fuelcladding systems of immediate interest. Continued efforts in FY15 and FY17 aimed to better understand and simulate initial pellet-clad interaction with little-to-no pressure on the pellet-clad interface. Reported here are the results from 1000 h heat treatments at 400, 500, and 600°C of diffusion couples pairing UN with a FeCrAl alloy, SiC, and Zr-based cladding candidate sealed in evacuated quartz ampoules. No gross reactions were observed, though trace elemental contaminants were identified.

  18. The measurement of residual stresses in claddings

    International Nuclear Information System (INIS)

    Hofer, G.; Bender, N.

    1978-01-01

    The ring core method, a variation of the hole drilling method for the measurement of biaxial residual stresses, has been extended to measure stresses from depths of about 5 to 25mm. It is now possible to measure the stress profiles of clad material. Examples of measured stress profiles are shown and compared with those obtained with a sectioning technique. (author)

  19. Cladding For Transversely-Pumped Laser Rod

    Science.gov (United States)

    Byer, Robert L.; Fan, Tso Yee

    1989-01-01

    Combination of suitable dimensioning and cladding of neodymium:yttrium aluminum garnet of similar solid-state laser provides for more efficient utilization of transversely-incident pump light from diode lasers. New design overcomes some of limitations of longitudinal- and older transverse-pumping concepts and promotes operation at higher output powers in TEM00 mode.

  20. Prevention of nuclear fuel cladding materials corrosion

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  1. Mechanism for iodine cracking of zirconium claddings

    International Nuclear Information System (INIS)

    Novikov, V.V.

    1991-01-01

    The mechanism of iodine cracking of zirconium cladding is analyzed taking into account the effect of stresses on diffusion. A decisive effect of the stress gradiemt on crack propagation in an agressive medium is shown. The experimental data are compared with the proposed model

  2. Thermodynamics of pellet-cladding interaction

    International Nuclear Information System (INIS)

    Kyoh, Bunkei; Fuji, Kensho

    1987-01-01

    Equilibrium thermodynamic calculations are performed on the U-Zr-Cs-I-O system that is assumed to exist in the fuel-cladding gap of light water reactor (LWR) fuel under pellet-cladding interaction (PCI) failure condition. For this purpose a computer program called SOLGASMIX-PV for the calculation of complex multi-component equilibria is used, and the results of postirradiation examination are interpreted. The analysis of the thermodynamics of the system U-Zr-Cs-I-O indicates that cesium and iodine are assumed to be released from fuel pellet into the fuel-cladding gap as CsI, therefore, the Cs/I ratio in fuel-cladding bonding zone is one. The important condensed phases in this region are UO 2 , U 3 O 8 , Cs 2 U 2 O 7 , Cs 2 U 15 O 46 , ZrO 2 and CsI, and the major gaseous species are CsI, I 2 and I. Under this situation where Cs/I ratio is one, cesium-zirconate is not present. If, however, cesium rich phase is partially present then cesium will be associated with zirconium, possibly as Cs 2 ZrO 3 . (author)

  3. Study and Behaviour of Prefabricated Composite Cladding

    Science.gov (United States)

    Sai Avinash, P.; Thiagarajan, N.; Santhi, A. S.

    2017-07-01

    The incessant population rise entailed for an expeditious construction at competitive prices that steered the customary path to the light weight structural components. This lead to construction of structural components using ferrocement. The load bearing structural cladding, sizing 3200x900x100 mm, is chosen for the study, which, is analyzed using the software ABAQUS 6.14 in accordance with the IS:875-87 Part1, IS:875-87 Part2, ACI 549R-97, ACI 318R-08 and NZS:3101-06 Part1 standards. The Ferrocement claddings (FCs) are fabricated to a scaled dimension of 400x115x38 mm. The light weight-high strength phenomena are corroborated by incorporating Glass Fibre Reinforced Polymer Laminates (GFRPL) of thickness 6mm, engineered with the aid of hand layup (wet layup) technique wielding epoxy resin, followed by curing under room temperature. The epoxy resin is employed for fastening ferrocement cladding with the Glass fiber reinforced polymer laminate, with the contemporary methodology. The compressive load carrying capacity of the amalgamated assembly, both in presence and absence of Glass Fibre Reinforced polymer laminates (GFRPL) on either side of Ferrocement cladding, has been experimented.

  4. Method for decontaminating stainless cladding tubes

    International Nuclear Information System (INIS)

    Komatsu, Fumiaki.

    1986-01-01

    Purpose: To form an oxide film over the surface of stainless cladding tubes and to efficiently remove radioactive materials from the steel surface together with the oxide layer by the use of an acid water solution. Method: After the removal of water from cladding tubes that have passed through the re-processing process, an oxide film is formed on the surface of the cladding tubes by heating over 400 deg C in an oxidizing atmosphere and thereafter washed again in an acid water solution. When the cladding tubes are thus oxidized once, the stainless base metal itself is oxidized, an oxide layer of several 10 μm or more being formed thereon. In consequence, since the oxide layer is far inferior in corrosion resistance to stainless metals, a pickling liquid easily penetrates into the stainless metal through the oxide layer, thereby remarkably promoting the peeling of the layer from the base metal surface and also improving the residual radioactive material removing efficiency together. (Takahashi, M.)

  5. Advanced ceramic cladding for water reactor fuel

    International Nuclear Information System (INIS)

    Feinroth, H.

    2000-01-01

    Under the US Department of Energy's Nuclear Energy Research Initiatives (NERI) program, continuous fiber ceramic composites (CFCCs) are being developed as cladding for water reactor fuel elements. The purpose is to substantially increase the passive safety of water reactors. A development effort was initiated in 1991 to fabricate CFCC-clad tubes using commercially available fibers and a sol-gel process developed by McDermott Technologies. Two small-diameter CFCC tubes were fabricated using pure alumina and alumina-zirconia fibers in an alumina matrix. Densities of approximately 60% of theoretical were achieved. Higher densities are required to guarantee fission gas containment. This NERI work has just begun, and only preliminary results are presented herein. Should the work prove successful, further development is required to evaluate CFCC cladding and performance, including in-pile tests containing fuel and exploring a marriage of CFCC cladding materials with suitable advanced fuel and core designs. The possibility of much higher temperature core designs, possibly cooled with supercritical water, and achievement of plant efficiencies ge50% would be examined

  6. Cladding Effects on Structural Integrity of Nuclear Components

    International Nuclear Information System (INIS)

    Sattari-Far, Iradi; Andersson, Magnus

    2006-06-01

    Based on this study, the following conclusions and recommendations can be made: Due to significant differences in the thermal and mechanical properties between the austenitic cladding and the ferritic base metal, residual stresses are induced in the cladding and the underlying base metal. These stresses are left in clad components even after Post-Weld Heat Treatment (PWHT). The different restraint conditions of the clad component have a minor influence on the magnitude of the cladding residual stresses in the cladding layer. The thickness of the clad object is the main impacting geometrical dimension in developing cladding residual stresses. A clad object having a base material thickness exceeding 10 times the cladding thickness would be practically sufficient to introduce cladding residual stresses of a thick reactor pressure vessel. For a clad component that received PWHT, the peak tensile stress is in the cladding layer, and the residual stresses in the underlying base material are negligible. However, for clad components not receiving PWHT, for instance the repair welding of the cladding, the cladding residual stresses of tensile type exist even in the base material. This implies a higher risk for underclad cracking for clad repairs that received no PWHT. For certain clad geometries, like nozzles, the profile of the cladding residual stresses depends on the clad thickness and position, and significant tensile stresses can also exist in the base material. Based on different measurements reported in the literature, a value of 150 GPa can be used as Young's Modulus of the austenitic cladding material at room temperature. The control measurements of small samples from the irradiated reactor pressure vessel head did not reveal a significant difference of Young's Modulus between the irradiated and the unirradiated cladding material condition. No significant differences between the axial and tangential cladding residual stresses are reported in the measurement of

  7. Cladding Effects on Structural Integrity of Nuclear Components

    Energy Technology Data Exchange (ETDEWEB)

    Sattari-Far, Iradi; Andersson, Magnus [lnspecta Technology AB, Stockholm (Sweden)

    2006-06-15

    Based on this study, the following conclusions and recommendations can be made: Due to significant differences in the thermal and mechanical properties between the austenitic cladding and the ferritic base metal, residual stresses are induced in the cladding and the underlying base metal. These stresses are left in clad components even after Post-Weld Heat Treatment (PWHT). The different restraint conditions of the clad component have a minor influence on the magnitude of the cladding residual stresses in the cladding layer. The thickness of the clad object is the main impacting geometrical dimension in developing cladding residual stresses. A clad object having a base material thickness exceeding 10 times the cladding thickness would be practically sufficient to introduce cladding residual stresses of a thick reactor pressure vessel. For a clad component that received PWHT, the peak tensile stress is in the cladding layer, and the residual stresses in the underlying base material are negligible. However, for clad components not receiving PWHT, for instance the repair welding of the cladding, the cladding residual stresses of tensile type exist even in the base material. This implies a higher risk for underclad cracking for clad repairs that received no PWHT. For certain clad geometries, like nozzles, the profile of the cladding residual stresses depends on the clad thickness and position, and significant tensile stresses can also exist in the base material. Based on different measurements reported in the literature, a value of 150 GPa can be used as Young's Modulus of the austenitic cladding material at room temperature. The control measurements of small samples from the irradiated reactor pressure vessel head did not reveal a significant difference of Young's Modulus between the irradiated and the unirradiated cladding material condition. No significant differences between the axial and tangential cladding residual stresses are reported in the

  8. Machining of uranium and uranium alloys

    International Nuclear Information System (INIS)

    Morris, T.O.

    1981-01-01

    Uranium and uranium alloys can be readily machined by conventional methods in the standard machine shop when proper safety and operating techniques are used. Material properties that affect machining processes and recommended machining parameters are discussed. Safety procedures and precautions necessary in machining uranium and uranium alloys are also covered. 30 figures

  9. The improvement of technology for high-uranium-density Al-base dispersion fuel plates

    International Nuclear Information System (INIS)

    Shouhui, Dai; Rongxian, Sun; Hejian, Mao; Baosheng, Zhao; Changgen, Yin

    1987-01-01

    An improved rolling process was developed for manufacturing Al-base dispersion fuel plates. When the fuel content in the meat increased up to 50 vol%, the non-uniformity of uranium is not more than ± 7.2%, and the minimum cladding thickness is not less than 0.32 mm. (Author)

  10. Kr ion irradiation study of the depleted-uranium alloys

    Science.gov (United States)

    Gan, J.; Keiser, D. D.; Miller, B. D.; Kirk, M. A.; Rest, J.; Allen, T. R.; Wachs, D. M.

    2010-12-01

    Fuel development for the reduced enrichment research and test reactor (RERTR) program is tasked with the development of new low enrichment uranium nuclear fuels that can be employed to replace existing high enrichment uranium fuels currently used in some research reactors throughout the world. For dispersion type fuels, radiation stability of the fuel-cladding interaction product has a strong impact on fuel performance. Three depleted-uranium alloys are cast for the radiation stability studies of the fuel-cladding interaction product using Kr ion irradiation to investigate radiation damage from fission products. SEM analysis indicates the presence of the phases of interest: U(Al, Si) 3, (U, Mo)(Al, Si) 3, UMo 2Al 20, U 6Mo 4Al 43 and UAl 4. Irradiations of TEM disc samples were conducted with 500 keV Kr ions at 200 °C to ion doses up to 2.5 × 10 19 ions/m 2 (˜10 dpa) with an Kr ion flux of 10 16 ions/m 2/s (˜4.0 × 10 -3 dpa/s). Microstructural evolution of the phases relevant to fuel-cladding interaction products was investigated using transmission electron microscopy.

  11. Kr ion irradiation study of the depleted-uranium alloys

    Energy Technology Data Exchange (ETDEWEB)

    Gan, J., E-mail: Jian.Gan@inl.go [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Keiser, D.D. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Miller, B.D. [University of Wisconsin, 1500 Engineering Drive, Madison, WI 53706 (United States); Kirk, M.A.; Rest, J. [Argonne National Laboratory, 9700 South Cass Ave., Argonne, IL 60439 (United States); Allen, T.R. [University of Wisconsin, 1500 Engineering Drive, Madison, WI 53706 (United States); Wachs, D.M. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States)

    2010-12-01

    Fuel development for the reduced enrichment research and test reactor (RERTR) program is tasked with the development of new low enrichment uranium nuclear fuels that can be employed to replace existing high enrichment uranium fuels currently used in some research reactors throughout the world. For dispersion type fuels, radiation stability of the fuel-cladding interaction product has a strong impact on fuel performance. Three depleted-uranium alloys are cast for the radiation stability studies of the fuel-cladding interaction product using Kr ion irradiation to investigate radiation damage from fission products. SEM analysis indicates the presence of the phases of interest: U(Al, Si){sub 3}, (U, Mo)(Al, Si){sub 3}, UMo{sub 2}Al{sub 20}, U{sub 6}Mo{sub 4}Al{sub 43} and UAl{sub 4}. Irradiations of TEM disc samples were conducted with 500 keV Kr ions at 200 {sup o}C to ion doses up to 2.5 x 10{sup 19} ions/m{sup 2} ({approx}10 dpa) with an Kr ion flux of 10{sup 16} ions/m{sup 2}/s ({approx}4.0 x 10{sup -3} dpa/s). Microstructural evolution of the phases relevant to fuel-cladding interaction products was investigated using transmission electron microscopy.

  12. Evolution of transmission spectra of double cladding fiber during etching

    Science.gov (United States)

    Ivanov, Oleg V.; Tian, Fei; Du, Henry

    2017-11-01

    We investigate the evolution of optical transmission through a double cladding fiber-optic structure during etching. The structure is formed by a section of SM630 fiber with inner depressed cladding between standard SMF-28 fibers. Its transmission spectrum exhibits two resonance dips at wavelengths where two cladding modes have almost equal propagation constants. We measure transmission spectra with decreasing thickness of the cladding and show that the resonance dips shift to shorter wavelengths, while new dips of lower order modes appear from long wavelength side. We calculate propagation constants of cladding modes and resonance wavelengths, which we compare with the experiment.

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

    International Nuclear Information System (INIS)

    Bezold, H.

    1979-01-01

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

  14. Diffused zircaloy 2/stainless steel junctions; Jonctions diffusees zircaloy 2 - acier inoxydable

    Energy Technology Data Exchange (ETDEWEB)

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

    1964-07-01

    The diffusion permits to realize joints between two different materials, in fact of the formation of a liquid phase at the contact face. The study of the tensile properties allowed the determination of the ideal conditions for the diffusion treatment which are, within 2 and 3 minutes for a temperature within 1020 C and 1030 C. The characteristics of the so obtained joints were, studied: mechanical properties, tightness, resistance to thermal cycling. Analysis of the thermal stress, owing to the differential dilatation of the two materials mode the object of a particular study. The investigation on the diffusion zone, includes specially, an analysis of the constituents distribution formed during the diffusion treatment. (author) [French] La diffusion permet de realiser des joints entre deux materiaux differents, du fait de la formation d'une phase liquide a l'interface de contact. L'etude de la resistance a la traction a permis de determiner les conditions optimum du traitement de diffusion: une duree de 2 a 3 minutes pour une temperature comprise entre 1020 C et 1030 C. Les caracteristiques des jonctions ainsi obtenues ont ete etudiees: proprietes mecaniques, etancheite, resistance au cyclage thermique. L'analyse des contraintes thermiques dues a la difference de dilatation des deux materiaux, a fait l'objet d'une etude particuliere. L'etude metallurgique de la zone diffusee comporte en particulier une analyse de la repartition des constituants formes lors du traitement de diffusion. (auteur)

  15. High-temperature irradiation of niobium-1 w/o zirconium-clad UO/sub 2/. [Compatibility with lithium

    Energy Technology Data Exchange (ETDEWEB)

    Kangilaski, M.; Fromm, E.O.; Lozier, D.H.; Storhok, V.W.; Gates, J.E.

    1965-06-28

    Twenty-four 0.225-in.-diameter and six 0.290-in.-diameter UO/sub 2/ specimens clad with 80 mils of niobium-1 w/o zirconium were irradiated to burnups of 1.4 to 6.0 at. % of uranium at surface temperatures of 900 to 1400/sup 0/C. UO/sub 2/ and lithium were found to be incompatible at these temperatures, and the thick cladding was used primarily to minimize the chances of contact of UO/sub 2/ and the lithium coolant. The thickly clad specimens did not undergo any dimensional changes as a result of irradiation, although it was found that movement of UO/sub 2/ took place in the axial direction by a vaporization-redeposition mechanism. It was found that 32 to 87% of the fission gases was released from the fuel, depending on the temperature of the specimen. Metallographic examination of longitudinal and transverse sections of the specimens indicated the usual UO/sub 2/ microstructure with columnar grains. Grain-boundary thickening was observed in the UO/sub 2/ at higher burnups. The oxygen/uranium ratio of UO/sub 2/ increased with increasing burnup.

  16. Research on laser cladding control system based on fuzzy PID

    Science.gov (United States)

    Zhang, Chuanwei; Yu, Zhengyang

    2017-12-01

    Laser cladding technology has a high demand for control system, and the domestic laser cladding control system mostly uses the traditional PID control algorithm. Therefore, the laser cladding control system has a lot of room for improvement. This feature is suitable for laser cladding technology, Based on fuzzy PID three closed-loop control system, and compared with the conventional PID; At the same time, the laser cladding experiment and friction and wear experiment were carried out under the premise of ensuring the reasonable control system. Experiments show that compared with the conventional PID algorithm in fuzzy the PID algorithm under the surface of the cladding layer is more smooth, the surface roughness increases, and the wear resistance of the cladding layer is also enhanced.

  17. Uranium mining

    International Nuclear Information System (INIS)

    Cheeseman, E.W.

    1980-01-01

    The international uranium market appears to be currently over-supplied with a resultant softening in prices. Buyers on the international market are unhappy about some of the restrictions placed on sales by the government, and Canadian sales may suffer as a result. About 64 percent of Canada's shipments come from five operating Ontario mines, with the balance from Saskatchewan. Several other properties will be producing within the next few years. In spite of the adverse effects of the Three Mile Island incident and the default by the T.V.A. of their contract, some 3 600 tonnes of new uranium sales were completed during the year. The price for uranium had stabilized at US $42 - $44 by mid 1979, but by early 1980 had softened somewhat. The year 1979 saw the completion of major environmental hearings in Ontario and Newfoundland and the start of the B.C. inquiry. Two more hearings are scheduled for Saskatchewan in 1980. The Elliot Lake uranium mining expansion hearings are reviewed, as are other recent hearings. In the production of uranium for nuclear fuel cycle, environmental matters are of major concern to the industry, the public and to governments. Research is being conducted to determine the most effective method for removing radium from tailings area effluents. Very stringent criteria are being drawn up by the regulatory agencies that must be met by the industry in order to obtain an operating licence from the AECB. These criteria cover seepages from the tailings basin and through the tailings retention dam, seismic stability, and both short and long term management of the tailings waste management area. (auth)

  18. Uranium industry annual 1996

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-04-01

    The Uranium Industry Annual 1996 (UIA 1996) provides current statistical data on the US uranium industry`s activities relating to uranium raw materials and uranium marketing. The UIA 1996 is prepared for use by the Congress, Federal and State agencies, the uranium and nuclear electric utility industries, and the public. Data on uranium raw materials activities for 1987 through 1996 including exploration activities and expenditures, EIA-estimated reserves, mine production of uranium, production of uranium concentrate, and industry employment are presented in Chapter 1. Data on uranium marketing activities for 1994 through 2006, including purchases of uranium and enrichment services, enrichment feed deliveries, uranium fuel assemblies, filled and unfilled market requirements, uranium imports and exports, and uranium inventories are shown in Chapter 2. A feature article, The Role of Thorium in Nuclear Energy, is included. 24 figs., 56 tabs.

  19. Uranium industry annual 1996

    International Nuclear Information System (INIS)

    1997-04-01

    The Uranium Industry Annual 1996 (UIA 1996) provides current statistical data on the US uranium industry's activities relating to uranium raw materials and uranium marketing. The UIA 1996 is prepared for use by the Congress, Federal and State agencies, the uranium and nuclear electric utility industries, and the public. Data on uranium raw materials activities for 1987 through 1996 including exploration activities and expenditures, EIA-estimated reserves, mine production of uranium, production of uranium concentrate, and industry employment are presented in Chapter 1. Data on uranium marketing activities for 1994 through 2006, including purchases of uranium and enrichment services, enrichment feed deliveries, uranium fuel assemblies, filled and unfilled market requirements, uranium imports and exports, and uranium inventories are shown in Chapter 2. A feature article, The Role of Thorium in Nuclear Energy, is included. 24 figs., 56 tabs

  20. Uranium industry annual, 1991

    International Nuclear Information System (INIS)

    1992-10-01

    In the Uranium Industry Annual 1991, data on uranium raw materials activities including exploration activities and expenditures, resources and reserves, mine production of uranium, production of uranium concentrate, and industry employment are presented in Chapter 1. Data on uranium marketing activities including domestic uranium purchases, commitments by utilities, procurement arrangements, uranium imports under purchase contracts and exports, deliveries to enrichment suppliers, inventories, secondary market activities, utility market requirements, and uranium for sale by domestic suppliers are presented in Chapter 2. A feature article entitled ''The Uranium Industry of the Commonwealth of Independent States'' is included in this report

  1. Theoretical studies of the influence of filler material gas gap and cladding material on rewetting rate of nuclear reactor fuel pins

    International Nuclear Information System (INIS)

    Blackburn, D.; Pearson, K.G.; Shires, G.L.

    1977-03-01

    Theoretical studies of the effect of fuel and gas gap on the rewetting rate of overheated fuel pins quenched by a falling film of water are presented. Two approaches have been made: a finite difference technique and an approximate analytical solution. The results obtained by the two methods for the case of a uranium-dioxide-filled Zircaloy clad fuel pin are in close agreement. The paper shows that under high pressure conditions the delaying effect of the stored heat within the fuel on the wetting rate is relatively small, particularly if a gas gap is present between the clad and the fuel. At low pressure conditions, however, the effect of the fuel may be very important. Simplification of the analytical solution shows that at low wetting rates a constant fractional reduction in wetting speed may be anticipated the magnitude of which depends only on the relative thermal diffusivities and heat capacities of the fuel and cladding. (author)

  2. Uranium - what role

    International Nuclear Information System (INIS)

    Grey, T.; Gaul, J.; Crooks, P.; Robotham, R.

    1980-01-01

    Opposing viewpoints on the future role of uranium are presented. Topics covered include the Australian Government's uranium policy, the status of nuclear power around the world, Australia's role as a uranium exporter and problems facing the nuclear industry

  3. Brazilian uranium exploration program

    International Nuclear Information System (INIS)

    Marques, J.P.M.

    1981-01-01

    General information on Brazilian Uranium Exploration Program, are presented. The mineralization processes of uranium depoits are described and the economic power of Brazil uranium reserves is evaluated. (M.C.K.) [pt

  4. Uranium enrichment

    International Nuclear Information System (INIS)

    1991-11-01

    This paper analyzes under four different scenarios the adequacy of a $500 million annual deposit into a fund to pay for the cost of cleaning up the Department of Energy's (DOE) three aging uranium enrichment plants. These plants are located in Oak Ridge, Tennessee; Paducah, Kentucky; and Portsmouth, Ohio. In summary the following was found: A fixed annual $500 million deposit made into a cleanup fund would not be adequate to cover total expected cleanup costs, nor would it be adequate to cover expected decontamination and decommissioning (D and D) costs. A $500 million annual deposit indexed to an inflation rate would likely be adequate to pay for all expected cleanup costs, including D and D costs, remedial action, and depleted uranium costs

  5. Uranium production

    International Nuclear Information System (INIS)

    Spriggs, M.

    1980-01-01

    The balance between uranium supply and demand is examined. Should new resources become necessary, some unconventional sources which could be considered include low-grade extensions to conventional deposits, certain types of intrusive rock, tuffs, and lake and sea-bed sediments. In addition there are large but very low grade deposits in carbonaceous shales, granites, and seawater. The possibility of recovery is discussed. Programmes of research into the feasibility of extraction of uranium from seawater, as a by-product from phosphoric acid production, and from copper leach solutions, are briefly discussed. Other possible sources are coal, old mine dumps and tailings, the latter being successfully exploited commercially in South Africa. The greatest constraints on increased development of U from lower grade sources are economics and environmental impact. It is concluded that apart from U as a by-product from phosphate, other sources are unlikely to contribute much to world requirements in the foreseeable future. (U.K.)

  6. Spent fuel cladding containment credit test

    International Nuclear Information System (INIS)

    Wilson, C.N.

    1983-01-01

    As an initial step in addressing the effectiveness of breached cladding as a barrier to radionuclide release from the repository during the post-containment period, preliminary scoping tests have been initiated which compare radionuclide releases from spent fuel specimens with artificially induced cladding defects of various severities. The artificially induced defects are all more severe than the typical in-reactor type breaches which are expected to be the principal type of breach entering the repository for terminal storage. These preliminary scoping tests being conducted by Westinghouse Hanford Company for the Lawrence Livermore National Laboratory Waste Package Development Program in support of the Tuff repository project at the Nevada Test Site are described. Also included in this presentation are selected initial results from these tests. 22 figures

  7. Automatic welding and cladding in heavy fabrication

    International Nuclear Information System (INIS)

    Altamer, A. de

    1980-01-01

    A description is given of the automatic welding processes used by an Italian fabricator of pressure vessels for petrochemical and nuclear plant. The automatic submerged arc welding, submerged arc strip cladding, pulsed TIG, hot wire TIG and MIG welding processes have proved satisfactory in terms of process reliability, metal deposition rate, and cost effectiveness for low alloy and carbon steels. An example shows sequences required during automatic butt welding, including heat treatments. Factors which govern satisfactory automatic welding include automatic anti-drift rotator device, electrode guidance and bead programming system, the capability of single and dual head operation, flux recovery and slag removal systems, operator environment and controls, maintaining continuity of welding and automatic reverse side grinding. Automatic welding is used for: joining vessel sections; joining tubes to tubeplate; cladding of vessel rings and tubes, dished ends and extruded nozzles; nozzle to shell and butt welds, including narrow gap welding. (author)

  8. Creep collapse of TAPS fuel cladding

    International Nuclear Information System (INIS)

    Chaudhry, S.M.; Anand, A.K.

    1975-01-01

    Densification of UO 2 can cause axial gaps between fuel pelets and cladding in unsupported (internally) at these regions. An analysis is carried out regarding the possibility of creep collapse in these regions. The analysis is based on Timoshenko's theory of collapse. At various times during the residence of fuel in reactor following parameters are calculated : (1) inelastic collapse of perfectly circular tubes (2) plastic instability in oval tubes (3) effect of creep on ovality. Creep is considered to be a non-linear combination of the following : (a) thermal creep (b) intresenic creep (c) stress aided radiation enhanced (d) stress free growth (4) Critical pressure ratio. The results obtained are compared with G.E. predictions. The results do not predict collapse of TAPS fuel cladding for five year residence time. (author)

  9. Development of Cr Electroplated Cladding Tube for preventing Fuel-Cladding Chemical Interaction (FCCI)

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jun Hwan; Woo, Je Woong; Kim, Sung Ho; Cheon, Jin Sik; Lee, Byung Oon; Lee, Chan Bock [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    Metal fuel has been selected as a candidate fuel in the SFR because of its superior thermal conductivity as well as enhanced proliferation resistance in connection with the pyroprocessing. However, metal fuel suffers eutectic reaction (Fuel Cladding Chemical Interaction, FCCI) with the fuel cladding made of stainless steel at reactor operating temperature so that cladding thickness gradually reduces to endanger reactor safety. In order to mitigate FCCI, barrier concept has been proposed between the fuel and the cladding in designing fuel rod. Regarding this, KAERI has initiated barrier cladding development to prevent interdiffusion process as well as enhance the SFR fuel performance. Previous study revealed that Cr electroplating has been selected as one of the most promising options because of its technical and economic viability. This paper describes the development status of the Cr electroplating technology for the usage of fuel rod in SFR. This paper summarizes the status of Cr electroplating technology to prevent FCCI in metal fuel rod. It has been selected for the ease of practical application at the tube inner surface. Technical scoping, performance evaluation and optimization have been carried out. Application to the tube inner surface and in-pile test were conducted which revealed as effective.

  10. Uranium enrichment

    International Nuclear Information System (INIS)

    1991-08-01

    This paper reports that in 1990 the Department of Energy began a two-year project to illustrate the technical and economic feasibility of a new uranium enrichment technology-the atomic vapor laser isotope separation (AVLIS) process. GAO believes that completing the AVLIS demonstration project will provide valuable information about the technical viability and cost of building an AVLIS plant and will keep future plant construction options open. However, Congress should be aware that DOE still needs to adequately demonstrate AVLIS with full-scale equipment and develop convincing cost projects. Program activities, such as the plant-licensing process, that must be completed before a plant is built, could take many years. Further, an updated and expanded uranium enrichment analysis will be needed before any decision is made about building an AVLIS plant. GAO, which has long supported legislation that would restructure DOE's uranium enrichment program as a government corporation, encourages DOE's goal of transferring AVLIS to the corporation. This could reduce the government's financial risk and help ensure that the decision to build an AVLIS plant is based on commercial concerns. DOE, however, has no alternative plans should the government corporation not be formed. Further, by curtailing a planned public access program, which would have given private firms an opportunity to learn about the technology during the demonstration project, DOE may limit its ability to transfer AVLIS to the private sector

  11. LASER CLADDING ON ALUMINIUM BASE ALLOYS

    OpenAIRE

    Pilloz , M.; Pelletier , J.; Vannes , A.; Bignonnet , A.

    1991-01-01

    laser cladding is often performed on iron or titanium base alloys. In the present work, this method is employed on aluminum alloys ; nickel or silicon are added by powder injection. Addition of silicon leads to sound surface layers, but with moderated properties, while the presence of nickel induces the formation of hard intermetallic compounds and then to an attractive hardening phenomena ; however a recovery treatment has to be carried out, in order to eliminate porosity in the near surface...

  12. Nuclear-powered pacemaker fuel cladding study

    International Nuclear Information System (INIS)

    Shoup, R.L.

    1976-01-01

    The composite of metals and alloys used in the fabrication of 238 Pu cardiac pacemaker fuel capsules resists the effects of high temperatures, high mechanical forces, and chemical corrosives and provides more than adequate protection to the fuel pellet even from deliberate attempts to dissolve the cladding in inorganic acids. This does not imply that opening a pacemaker fuel capsule by inorganic acids is impossible but that it would not be a wise choice

  13. Corrosion behaviour of laser clad stainless steels

    International Nuclear Information System (INIS)

    Damborenea, J.J. de; Weerasinghe, V.M.; West, D.R.F.

    1993-01-01

    The present paper is focussed in the study of the properties of a clad layer of stainless steel on a mild steel. By blowing powder of the alloy into a melt pool generated by a laser of 2 KW, an homogeneous layer of 316 stainless steel can be obtained. Structure, composition and corrosion behaviour are similar to those of a stainless steel in as-received condition. (Author)

  14. Plasma spheroidization and cladding of powders

    Energy Technology Data Exchange (ETDEWEB)

    Petrunichev, V.A.; Averin, V.V.; Sorokin, L.M.; Koroleva, E.B.

    1987-02-01

    With reference to experimental results for nickel and chromium alloys, it is shown that complex alloy powders can be spheroidized in plasma discharges using an argon plasma with hydrogen. The spheroidizing process is accompanied by the reduction of surface oxides, with uniform element distribution within the particles; the granulometric composition of the particles is preserved. It is also shown that plasma technology can be used for producing metal-clad oxide and carbide powders, which improve the performance of cermets and coatings.

  15. Derived enriched uranium market

    International Nuclear Information System (INIS)

    Rutkowski, E.

    1996-01-01

    The potential impact on the uranium market of highly enriched uranium from nuclear weapons dismantling in the Russian Federation and the USA is analyzed. Uranium supply, conversion, and enrichment factors are outlined for each country; inventories are also listed. The enrichment component and conversion components are expected to cause little disruption to uranium markets. The uranium component of Russian derived enriched uranium hexafluoride is unresolved; US legislation places constraints on its introduction into the US market

  16. Alloy development for high burnup cladding (PWR)

    Energy Technology Data Exchange (ETDEWEB)

    Hahn, R. [Kraftwerk Union AG, Mulheim (Germany); Jeong, Y.H.; Baek, K.H.; Kim, S.J.; Choi, B.K.; Kim, J.M. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1999-04-01

    An overview on current alloy development for high burnup PWR fuel cladding is given. It is mainly based on literature data. First, the reasons for an increase of the current mean discharge burnup from 35 MWd / kg(U) to 70 MWd / kg(U) are outlined. From the material data, it is shown that a batch average burnup of 60-70 MWd / kg(U), as aimed by many fuel vendors, can not be achieved with stand (=ASTM-) Zry-4 cladding tubes without violating accepted design criteria. Specifically criteria which limit maximum oxide scale thickness and maximum hydrogen content, and to a less degree, maximum creep and growth rate, can not be achieved. The development potential of standard Zry-4 is shown. Even when taking advantage of this potential, it is shown that an 'improved' Zry-4 is reaching its limits when it achieves the target burnup. The behavior of some Zr alloys outside the ASTM range is shown, and the advantages and disadvantages of the 3 alloy groups (ZrSn+transition metals, ZrNb, ZrSnNb+transition metals) which are currently considered to have the development potential for high burnup cladding materials are depicted. Finally, conclusions are drawn. (author). 14 refs., 11 tabs., 82 figs.

  17. Advanced LWR Nuclear Fuel Cladding Development

    International Nuclear Information System (INIS)

    Bragg-Sitton, S.; Griffith, G.

    2012-01-01

    The Advanced Light Water Reactor (LWR) Nuclear Fuel Development Research and Development (R and D) Pathway encompasses strategic research focused on improving reactor core economics and safety margins through the development of an advanced fuel cladding system. To achieve significant operating improvements while remaining within safety boundaries, significant steps beyond incremental improvements in the current generation of nuclear fuel are required. Fundamental enhancements are required in the areas of nuclear fuel composition, cladding integrity, and fuel/cladding interaction to allow improved fuel economy via power uprates and increased fuel burn-up allowance while potentially improving safety margin through the adoption of an 'accident tolerant' fuel system that would offer improved coping time under accident scenarios. In a staged development approach, the LWRS program will engage stakeholders throughout the development process to ensure commercial viability of the investigated technologies. Applying minimum performance criteria, several of the top-ranked materials and fabrication concepts will undergo a rigorous series of mechanical, thermal and chemical characterization tests to better define their properties and operating potential in a relatively low-cost, nonnuclear test series. A reduced number of options will be recommended for test rodlet fabrication and in-pile nuclear testing under steady-state, transient and accident conditions. (author)

  18. Cladding failure by local plastic instability

    International Nuclear Information System (INIS)

    Kramer, J.M.; Deitrich, L.W.

    1977-01-01

    Cladding failure is one of the major considerations in analysis of fast-reactor fuel pin behavior during hypothetical accident transients since time, location and nature of failure govern the early post-failure material motion and reactivity feedback. Out-of-Pile transient burst tests of both irradiated and unirradiated fast-reactor cladding show that local plastic instability, or bulging, often precedes rupture. To investigate the details of cladding bulging, a perturbation analysis of the equations governing the large deformation of a cylindrical shell has been developed. The overall deformation history is assumed to consist of a small perturbation epsilon of the radial displacement superimposed on large axisymmetric displacements. Computations have been carried out using high temperature properties of stainless steel in conjunction with various constitutive theories, including a generalization of the Endochronic Theory of Plasticity in which both time-independent and time-dependent plastic strains are modeled. Although the results of the calculations are all qualitatively similar, it appears that modeling of both time-independent and time-dependent plastic strains is necessary to interpret the transient burst test results. Sources for bulge formation that have been considered include initial geometric imperfections and thermal perturbations due to either eccentric fuel pellets or non-symmetric cooling. (Auth.)

  19. Uranium industry annual, 1986

    International Nuclear Information System (INIS)

    1987-01-01

    Uranium industry data collected in the EIA-858 survey provide a comprehensive statistical characterization of annual activities of the industry and include some information about industry plans over the next several years. This report consists of two major sections. The first addresses uranium raw materials activities and covers the following topics: exploration activities and expenditures, resources and reserves, mine production of uranium, production of uranium concentrate, and industry employment. The second major section is concerned with the following uranium marketing activities: uranium purchase commitments, uranium prices, procurement arrangements, uranium imports and exports, enrichment services, inventories, secondary market activities utility market requirements and related topics

  20. Uranium Industry. Annual 1984

    International Nuclear Information System (INIS)

    Lawrence, M.S.S.

    1985-01-01

    This report provides a statistical description of activities of the US uranium industry during 1984 and includes a statistical profile of the status of the industry at the end of 1984. It is based on the results of an Energy Information Administration (EIA) survey entitled ''Uranium Industry Annual Survey'' (Form EIA-858). The principal findings of the survey are summarized under two headings - Uranium Raw Materials Activities and Uranium Marketing Activities. The first heading covers exploration and development, uranium resources, mine and mill production, and employment. The second heading covers uranium deliveries and delivery commitments, uranium prices, foreign trade in uranium, inventories, and other marketing activities. 32 figs., 48 tabs

  1. The interdiffusion behaviour of uranium and aluminum

    Energy Technology Data Exchange (ETDEWEB)

    Le Claire, A D; Bear, I J

    1952-03-05

    Results are given for measurements made on the rate, as a function of temperature and pressure, of the increase in width of the diffusion zone formed by the interdiffusion of uranium and aluminium and of the relative penetration of uranium into aluminium and of aluminium into uranium. The experiments are unusual among diffusion meaurements in showing a marked effect of pressure on the rates of diffusion. Between 570 deg and 640 deg C (the eutectic temperature) the rate of growth of the diffusion zone increases regularly with increase of temperature or of pressure, at least up to the maximum pressure of 10 tons/sq.in. employed in the experiments. Below 570 deg C the rate of diffusion, rising with the rise in pressure at constant temperature increases discontinuously when the pressure reaches a value a little above 1 1/4 tons/sq.in. by a factor which increases, as the temperature is lowered, to a maximum value at 540 deg C. It then remains constant at this maximum value for temperatures below 540 deg C. The accelerating effect of pressure on the rate of interdiffusion is then most apparent when the temperature is below 540 deg C and the pressure raised from below to above about 1 1/4 tons/sq.in. The penetration of the uranium into the aluminium is about 2 1/4 times faster than that of the aluminium into the uranium. The results of a microscopical and X-ray examination of the diffusion zone are presented, and a discussion is given of the possible origins of the pressure effect and of the importance of the work in an understanding of the ''pimpling'' of aluminium clad uranium bars. (auth)

  2. Study of reactions between fuel (mixed oxide (UPu)Osub(2-x)) and cladding (stainless-steel) in reactors: influence of iodine compounds

    International Nuclear Information System (INIS)

    Aubert, Michel.

    1976-03-01

    The influence of iodine compounds on the development of the oxide-cladding reaction was examined. The action of iodine, cesium and cesium iodide on type 316 stainless was determined in the presence or absence of uranium oxide or mixed uranium-plutonium oxide type fuel in a closed system, isothermal or with a temperature gradient. The study of the stainless steel iodine reactions was developed in particular. These experiments showed that cesium combines with uranium oxide to give cesium uranate Cs 2 U 2 O 7 ; it is not unreasonable to suppose that cesium urano-plutonate Cs 2 (U,Pu) 2 O 7 could be formed inside the pile. It was then shown that cesium iodide in the presence of sufficiently non-stoichiometric mixed oxide could contribute towards the degradation of the stainless steel cladding. Under these conditions the reaction is accompained by a transport of manganese, chromium and iron into the hot parts of the fuel by a Van-Arkel type mechanism. This might explain the presence of metallic precipitates in the fuel, but the role assigned to molybdenum iodide in the same phenomenon is considered unlikely. Finally it is proposed to deposit a thin layer of manganese metal on the inner surface of the cladding in order to minimize the action of fission products (CsI, Te) [fr

  3. Microstructure and Mechanical Properties of Laser Clad and Post-cladding Tempered AISI H13 Tool Steel

    Science.gov (United States)

    Telasang, Gururaj; Dutta Majumdar, Jyotsna; Wasekar, Nitin; Padmanabham, G.; Manna, Indranil

    2015-05-01

    This study reports a detailed investigation of the microstructure and mechanical properties (wear resistance and tensile strength) of hardened and tempered AISI H13 tool steel substrate following laser cladding with AISI H13 tool steel powder in as-clad and after post-cladding conventional bulk isothermal tempering [at 823 K (550 °C) for 2 hours] heat treatment. Laser cladding was carried out on AISI H13 tool steel substrate using a 6 kW continuous wave diode laser coupled with fiber delivering an energy density of 133 J/mm2 and equipped with a co-axial powder feeding nozzle capable of feeding powder at the rate of 13.3 × 10-3 g/mm2. Laser clad zone comprises martensite, retained austenite, and carbides, and measures an average hardness of 600 to 650 VHN. Subsequent isothermal tempering converted the microstructure into one with tempered martensite and uniform dispersion of carbides with a hardness of 550 to 650 VHN. Interestingly, laser cladding introduced residual compressive stress of 670 ± 15 MPa, which reduces to 580 ± 20 MPa following isothermal tempering. Micro-tensile testing with specimens machined from the clad zone across or transverse to cladding direction showed high strength but failure in brittle mode. On the other hand, similar testing with samples sectioned from the clad zone parallel or longitudinal to the direction of laser cladding prior to and after post-cladding tempering recorded lower strength but ductile failure with 4.7 and 8 pct elongation, respectively. Wear resistance of the laser surface clad and post-cladding tempered samples (evaluated by fretting wear testing) registered superior performance as compared to that of conventional hardened and tempered AISI H13 tool steel.

  4. Electrically heated ex-reactor pellet-cladding interaction (PCI) simulations utilizing irradiated Zircaloy cladding

    International Nuclear Information System (INIS)

    Barner, J.O.; Fitzsimmons, D.E.

    1985-02-01

    In a program sponsored by the Fuel Systems Research Branch of the US Nuclear Regulatory Commission, a series of six electrically heated fuel rod simulation tests were conducted at Pacific Northwest Laboratory. The primary objective of these tests was to determine the susceptibility of irradiated pressurized-water reactor (PWR) Zircaloy-4 cladding to failures caused by pellet-cladding mechanical interaction (PCMI). A secondary objective was to acquire kinetic data (e.g., ridge growth or relaxation rates) that might be helpful in the interpretation of in-reactor performance results and/or the modeling of PCMI. No cladding failures attributable to PCMI occurred during the six tests. This report describes the testing methods, testing apparatus, fuel rod diametral strain-measuring device, and test matrix. Test results are presented and discussed

  5. Modeling Thermal and Stress Behavior of the Fuel-clad Interface in Monolithic Fuel Mini-plates

    International Nuclear Information System (INIS)

    Miller, Gregory K.; Medvedev, Pavel G.; Burkes, Douglas E.; Wachs, Daniel M.

    2010-01-01

    As part of the Global Threat Reduction Initiative, a fuel development and qualification program is in process with the objective of qualifying very high density low enriched uranium fuel that will enable the conversion of high performance research reactors with operational requirements beyond those supported with currently available low enriched uranium fuels. The high density of the fuel is achieved by replacing the fuel meat with a single monolithic low enriched uranium-molybdenum fuel foil. Doing so creates differences in the mechanical and structural characteristics of the fuel plate because of the planar interface created by the fuel foil and cladding. Furthermore, the monolithic fuel meat will dominate the structural properties of the fuel plate rather than the aluminum matrix, which is characteristic of dispersion fuel types. Understanding the integrity and behavior of the fuel-clad interface during irradiation is of great importance for qualification of the new fuel, but can be somewhat challenging to determine with a single technique. Efforts aimed at addressing this problem are underway within the fuel development and qualification program, comprised of modeling, as-fabricated plate characterization, and post-irradiation examination. An initial finite element analysis model has been developed to investigate worst-case scenarios for the basic monolithic fuel plate structure, using typical mini-plate irradiation conditions in the Advanced Test Reactor. Initial analysis shows that the stress normal to the fuel-clad interface dominates during irradiation, and that the presence of small, rounded delaminations at the interface is not of great concern. However, larger and/or fuel-clad delaminations with sharp corners can create areas of concern, as maximum principal cladding stress, strain, displacement, and peak fuel temperature are all significantly increased. Furthermore, stresses resulting from temperature gradients that cause the plate to bow or buckle in

  6. Impact of reactor water chemistry on cladding performance

    International Nuclear Information System (INIS)

    Cox, B.

    1997-01-01

    Water chemistry may have a major impact on fuel cladding performance in PWRs. If the saturation temperature on the surface of fuel cladding is exceeded, either because of the thermal hydraulics of the system, or because of crud deposition, then LiOH concentration can occur within thick porous oxide films on the cladding. This can degrade the protective film and accelerate the corrosion rate of the cladding. If sufficient boric acid is also present in the coolant then these effects may be mitigated. This is normally the case through most of any reactor fuel cycle. Extensive surface boiling may disrupt this equilibrium because of the volatility of boric acid in steam. Under such conditions severe cladding corrosion can ensue. The potential for such effects on high burnup cladding in CANDU reactors, where bone acid is not present in the primary coolant, is discussed. (author)

  7. Impact of reactor water chemistry on cladding performance

    Energy Technology Data Exchange (ETDEWEB)

    Cox, B. [University of Toronto, Centre for Nuclear Engineering, Toronto, Ontario (Canada)

    1997-07-01

    Water chemistry may have a major impact on fuel cladding performance in PWRs. If the saturation temperature on the surface of fuel cladding is exceeded, either because of the thermal hydraulics of the system, or because of crud deposition, then LiOH concentration can occur within thick porous oxide films on the cladding. This can degrade the protective film and accelerate the corrosion rate of the cladding. If sufficient boric acid is also present in the coolant then these effects may be mitigated. This is normally the case through most of any reactor fuel cycle. Extensive surface boiling may disrupt this equilibrium because of the volatility of boric acid in steam. Under such conditions severe cladding corrosion can ensue. The potential for such effects on high burnup cladding in CANDU reactors, where bone acid is not present in the primary coolant, is discussed. (author)

  8. Research Progress on Laser Cladding Amorphous Coatings on Metallic Substrates

    Directory of Open Access Journals (Sweden)

    CHEN Ming-hui

    2017-01-01

    Full Text Available The microstructure and property of amorphous alloy as well as the limitations of the traditional manufacturing methods for the bulk amorphous alloy were briefly introduced in this paper.Combined with characteristics of the laser cladding technique,the research status of the laser cladding Fe-based,Zr-based,Ni-based,Cu-based and Al-based amorphous coatings on the metal substrates were mainly summarized.The effects of factors such as laser processing parameter,micro-alloying element type and content and reinforcing phase on the laser cladding amorphous coatings were also involved.Finally,the main problems and the future research directions of the composition design and control of the laser-cladded amorphous coating,the design and optimization of the laser cladding process,and the basic theory of the laser cladding amorphous coatings were also put forward finally.

  9. Hygrothermal performance of ventilated wooden cladding

    Energy Technology Data Exchange (ETDEWEB)

    Nore, Kristine

    2009-10-15

    This project contributes to more accurate design guidelines for high-performance building envelopes by analysis of hygrothermal performance of ventilated wooden cladding. Hygrothermal performance is defined by cladding temperature and moisture conditions, and subsequently by risk of degradation. Wood cladding is the most common facade material used in rural and residential areas in Norway. Historically, wooden cladding design varied in different regions in Norway. This was due to both climatic variations and the logistical distance to materials and craftspeople. The rebuilding of Norwegian houses in the 1950s followed central guidelines where local climate adaptation was often not evaluated. Nowadays we find some technical solutions that do not withstand all climate exposures. The demand for thermal comfort and also energy savings has changed hygrothermal condition of the building envelopes. In well-insulated wall assemblies, the cladding temperature is lower compared to traditional walls. Thus the drying out potential is smaller, and the risk of decay may be higher. The climate change scenario indicates a warmer and wetter future in Norway. Future buildings should be designed to endure harsher climate exposure than at present. Is there a need for refined climate differentiated design guidelines for building enclosures? As part of the Norwegian research programme 'Climate 2000', varieties of wooden claddings have been investigated on a test house in Trondheim. The aim of this investigation was to increase our understanding of the relation between microclimatic conditions and the responding hygrothermal performance of wooden cladding, according to orientation, design of ventilation gap, wood material quality and surface treatment. The two test facades, facing east and west have different climate exposure. Hourly measurements of in total 250 sensors provide meteorological data; temperature, radiation, wind properties, relative humidity, and test house data

  10. Uranium price reporting systems

    International Nuclear Information System (INIS)

    1987-09-01

    This report describes the systems for uranium price reporting currently available to the uranium industry. The report restricts itself to prices for U 3 O 8 natural uranium concentrates. Most purchases of natural uranium by utilities, and sales by producers, are conducted in this form. The bulk of uranium in electricity generation is enriched before use, and is converted to uranium hexafluoride, UF 6 , prior to enrichment. Some uranium is traded as UF 6 or as enriched uranium, particularly in the 'secondary' market. Prices for UF 6 and enriched uranium are not considered directly in this report. However, where transactions in UF 6 influence the reported price of U 3 O 8 this influence is taken into account. Unless otherwise indicated, the terms uranium and natural uranium used here refer exclusively to U 3 O 8 . (author)

  11. Uranium Industry Annual, 1992

    International Nuclear Information System (INIS)

    1993-01-01

    The Uranium Industry Annual provides current statistical data on the US uranium industry for the Congress, Federal and State agencies, the uranium and electric utility industries, and the public. The feature article, ''Decommissioning of US Conventional Uranium Production Centers,'' is included. Data on uranium raw materials activities including exploration activities and expenditures, resources and reserves, mine production of uranium, production of uranium concentrate, and industry employment are presented in Chapter 1. Data on uranium marketing activities including domestic uranium purchases, commitments by utilities, procurement arrangements, uranium imports under purchase contracts and exports, deliveries to enrichment suppliers, inventories, secondary market activities, utility market requirements, and uranium for sale by domestic suppliers are presented in Chapter 2

  12. Uranium Industry Annual, 1992

    Energy Technology Data Exchange (ETDEWEB)

    1993-10-28

    The Uranium Industry Annual provides current statistical data on the US uranium industry for the Congress, Federal and State agencies, the uranium and electric utility industries, and the public. The feature article, ``Decommissioning of US Conventional Uranium Production Centers,`` is included. Data on uranium raw materials activities including exploration activities and expenditures, resources and reserves, mine production of uranium, production of uranium concentrate, and industry employment are presented in Chapter 1. Data on uranium marketing activities including domestic uranium purchases, commitments by utilities, procurement arrangements, uranium imports under purchase contracts and exports, deliveries to enrichment suppliers, inventories, secondary market activities, utility market requirements, and uranium for sale by domestic suppliers are presented in Chapter 2.

  13. 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

  14. Analyses on Silicide Coating for LOCA Resistant Cladding

    Energy Technology Data Exchange (ETDEWEB)

    Sweidan, Faris B.; Lee, You Ho; Ryu, Ho Jin [KAIST, Daejeon (Korea, Republic of)

    2015-10-15

    A particular focus of accident-tolerant fuel has been cladding due to the rapid high-temperature oxidation of zirconium-based cladding with the evolution of H2 when steam is a reactant. Some key features of the coated cladding include high-temperature resistance to oxidation, lower processing temperatures, and a high melting point of the coating. Zirconium alloys exhibit a reasonably high melting temperature, so a coating for the cladding is appealing if the coating increases the high-temperature resistance to oxidation. In this case, the cladding is protected from complete oxidation. The cladding coating involves the application of zirconium silicide onto Zr-based cladding. Zirconium silicide coating is expected to produce a glassy layer that becomes more protective at elevated temperature. For this reason, silicide coatings on cladding offer the potential for improved reliability at normal operating temperatures and at the higher transient temperatures encountered during accidents. Although ceramic coatings are brittle and may have weak points to be used as coating materials, several ceramic coatings were successful and showed adherent behavior and high resistance to oxidation. In this study, the oxidation behavior of zirconium silicide and its oxidation kinetics are analyzed. Zirconium silicide is a new suggested material to be used as coatings on existing Zr-based cladding alloys, the aim of this study is to evaluate if zirconium silicide is applicable to be used, so they can be more rapidly developed using existing cladding technology with some modifications. These silicide coatings are an attractive alternative to the use of coatings on zirconium claddings or to the lengthy development of monolithic ceramic or ceramic composite claddings and coatings.

  15. Analyses on Silicide Coating for LOCA Resistant Cladding

    International Nuclear Information System (INIS)

    Sweidan, Faris B.; Lee, You Ho; Ryu, Ho Jin

    2015-01-01

    A particular focus of accident-tolerant fuel has been cladding due to the rapid high-temperature oxidation of zirconium-based cladding with the evolution of H2 when steam is a reactant. Some key features of the coated cladding include high-temperature resistance to oxidation, lower processing temperatures, and a high melting point of the coating. Zirconium alloys exhibit a reasonably high melting temperature, so a coating for the cladding is appealing if the coating increases the high-temperature resistance to oxidation. In this case, the cladding is protected from complete oxidation. The cladding coating involves the application of zirconium silicide onto Zr-based cladding. Zirconium silicide coating is expected to produce a glassy layer that becomes more protective at elevated temperature. For this reason, silicide coatings on cladding offer the potential for improved reliability at normal operating temperatures and at the higher transient temperatures encountered during accidents. Although ceramic coatings are brittle and may have weak points to be used as coating materials, several ceramic coatings were successful and showed adherent behavior and high resistance to oxidation. In this study, the oxidation behavior of zirconium silicide and its oxidation kinetics are analyzed. Zirconium silicide is a new suggested material to be used as coatings on existing Zr-based cladding alloys, the aim of this study is to evaluate if zirconium silicide is applicable to be used, so they can be more rapidly developed using existing cladding technology with some modifications. These silicide coatings are an attractive alternative to the use of coatings on zirconium claddings or to the lengthy development of monolithic ceramic or ceramic composite claddings and coatings

  16. Clad fiber capacitor and method of making same

    Science.gov (United States)

    Tuncer, Enis

    2012-12-11

    A clad capacitor and method of manufacture includes assembling a preform comprising a ductile, electrically conductive fiber; a ductile, electrically insulating cladding positioned on the fiber; and a ductile, electrically conductive sleeve positioned over the cladding. One or more preforms are then bundled, heated and drawn along a longitudinal axis to decrease the diameter of the ductile components of the preform and fuse the preform into a unitized strand.

  17. Pellet-clad interaction in water reactor fuels

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2004-07-01

    The aim of this seminar is was to draw up a comprehensive picture of the pellet clad interaction and its impact on the fuel rod. This document is a detailed abstract of the papers presented during the following five sessions: industrial goals, fuel material behaviour in PCI situation, cladding behaviour relevant to PCI, in pile rod behaviour and modelling of the mechanical interaction between pellet and cladding. (A.L.B.)

  18. Pellet-clad interaction in water reactor fuels

    International Nuclear Information System (INIS)

    2004-01-01

    The aim of this seminar is was to draw up a comprehensive picture of the pellet clad interaction and its impact on the fuel rod. This document is a detailed abstract of the papers presented during the following five sessions: industrial goals, fuel material behaviour in PCI situation, cladding behaviour relevant to PCI, in pile rod behaviour and modelling of the mechanical interaction between pellet and cladding. (A.L.B.)

  19. Provision by the uranium and uranium products

    International Nuclear Information System (INIS)

    Elagin, Yu.P.

    2005-01-01

    International uranium market is converted from the buyer market into the seller market. The prices of uranium are high and the market attempts to adapt to changing circumstances. The industry of uranium enrichment satisfies the increasing demands but should to increase ots capacities. On the whole the situation is not stable and every year may change the existing position [ru

  20. Uranium recovery from slags of metallic uranium

    International Nuclear Information System (INIS)

    Fornarolo, F.; Frajndlich, E.U.C.; Durazzo, M.

    2006-01-01

    The Center of the Nuclear Fuel of the Institute of Nuclear Energy Research - IPEN finished the program of attainment of fuel development for research reactors the base of Uranium Scilicet (U 3 Si 2 ) from Hexafluoride of Uranium (UF 6 ) with enrichment 20% in weight of 235 U. In the process of attainment of the league of U 3 Si 2 we have as Uranium intermediate product the metallic one whose attainment generates a slag contend Uranium. The present work shows the results gotten in the process of recovery of Uranium in slags of calcined slags of Uranium metallic. Uranium the metallic one is unstable, pyrophoricity and extremely reactive, whereas the U 3 O 8 is a steady oxide of low chemical reactivity, what it justifies the process of calcination of slags of Uranium metallic. The calcination of the Uranium slag of the metallic one in oxygen presence reduces Uranium metallic the U 3 O 8 . Experiments had been developed varying it of acid for Uranium control and excess, nitric molar concentration gram with regard to the stoichiometric leaching reaction of temperature of the leaching process. The 96,0% income proves the viability of the recovery process of slags of Uranium metallic, adopting it previous calcination of these slags in nitric way with low acid concentration and low temperature of leaching. (author)

  1. Computer analysis of elongation of the WWER fuel rod claddings

    International Nuclear Information System (INIS)

    Scheglov, A.; Proselkov, V.

    2008-01-01

    In this paper description of mechanisms influencing changes of the WWER fuel cladding length and axial forces influencing fuel and cladding are presented. It is shown that shortening of the fuel claddings in case of high burnup can be explained by the change of the fuel and cladding reference state caused by reduction of the fuel rod power level - during reactor outages. It is noted that the presented calculated data are to be reviewed and interpreted as the preliminary results; further work is needed for their confirmation. (authors)

  2. Laser cladding of Zr on Mg for improved corrosion properties

    International Nuclear Information System (INIS)

    Subramanian, R.; Sircar, S.; Mazumder, J.

    1989-01-01

    This paper reports the results of laser cladding of Mg-2wt%Zr, and Mg-5wt%Zr powder mixture onto magnesium. The microstructure of the laser clad was studied. From the microstructural study, the epitaxial regrowth of the clad region on the underlying substrate was observed. Martensite plates of different size were observed in transmission electron microscope for MG-2wt%Zr and Mg-5wt%Zr laser clad. The corrosion properties of the laser clad were evaluated in sea water (3.5% NaCl). The position of the laser claddings in the galvanic series of metals in sea water, the anodic polarization characteristics of the laser claddings and the protective nature and the stability of the passivating film formed have been determined. The formation of pits on the surface of the laser clad subjected to corrosion is reported. The corrosion properties of the laser claddings are compared with that of the commercially used magnesium alloy AZ91B

  3. Rectangular-cladding silicon slot waveguide with improved nonlinear performance

    Science.gov (United States)

    Huang, Zengzhi; Huang, Qingzhong; Wang, Yi; Xia, Jinsong

    2018-04-01

    Silicon slot waveguides have great potential in hybrid silicon integration to realize nonlinear optical applications. We propose a rectangular-cladding hybrid silicon slot waveguide. Simulation result shows that, with a rectangular-cladding, the slot waveguide can be formed by narrower silicon strips, so the two-photon absorption (TPA) loss in silicon is decreased. When the cladding material is a nonlinear polymer, the calculated TPA figure of merit (FOMTPA) is 4.4, close to the value of bulk nonlinear polymer of 5.0. This value confirms the good nonlinear performance of rectangular-cladding silicon slot waveguides.

  4. Stainless steel clad for light water reactor fuels. Final report

    International Nuclear Information System (INIS)

    Rivera, J.E.; Meyer, J.E.

    1980-07-01

    Proper reactor operation and design guidelines are necessary to assure fuel integrity. The occurrence of fuel rod failures for operation in compliance with existing guidelines suggests the need for more adequate or applicable operation/design criteria. The intent of this study is to develop such criteria for light water reactor fuel rods with stainless steel clad and to indicate the nature of uncertainties in its development. The performance areas investigated herein are: long term creepdown and fuel swelling effects on clad dimensional changes and on proximity to clad failure; and short term clad failure possibilities during up-power ramps

  5. Determination of plastic anisotropy of zirconium alloys cladding

    International Nuclear Information System (INIS)

    Yamshchikov, N.V.; Prasolov, P.F.; Shestak, V.E.

    1991-01-01

    Method for determining plastic anisotropy of zurconium alloy cladding is described. It is based on consideration of material as a combination of transversal crystallites with known distribution over orientations. Such approach enables to describe cladding resistance to plastic deformation at arbitrary stressed state, using the results of texture investigations and uniaxial tests of samples, cut out of claddings along three directions. Plastic anisotropy of fuel element claddings 9.15 and 13.6 mm in diameter up to several percents of plastic deformation is shown

  6. Uranium enrichment

    International Nuclear Information System (INIS)

    Mohrhauer, H.

    1982-01-01

    The separation of uranium isotopes in order to enrich the fuel for light water reactors with the light isotope U-235 is an important part of the nuclear fuel cycle. After the basic principals of isotope separation the gaseous diffusion and the centrifuge process are explained. Both these techniques are employed on an industrial scale. In addition a short review is given on other enrichment techniques which have been demonstrated at least on a laboratory scale. After some remarks on the present situation on the enrichment market the progress in the development and the industrial exploitation of the gas centrifuge process by the trinational Urenco-Centec organisation is presented. (orig.)

  7. Basic research on high-uranium density fuels for research and test reactors

    International Nuclear Information System (INIS)

    Ugajin, M.; Itoh, A.; Akabori, M.

    1992-01-01

    High-uranium density fuels, uranium silicides (U 3 Si 2 , U 3 Si) and U 6 Me-type uranium alloys (Me = Fe, Mn, Ni), were prepared and examined metallurgically as low-enriched uranium (LEU) fuels for research and test reactors. Miniature aluminum-dispersion plate-type fuel (miniplate) and aluminum-clad disk-type fuel specimens were fabricated and subjected to the neutron irradiation in JMTR (Japan Materials Testing Reactor). Fuel-aluminum compatibility tests were conducted to elucidate the extent of reaction and to identify reaction products. The relative stability of the fuels in an aluminum matrix was established at 350degC or above. Experiments were also performed to predict the chemical form of the solid fission-products in the uranium silicide (U 3 Si 2 ) simulating a high burnup anticipated for reactor service. (author)

  8. Uranium conversion; Urankonvertering

    Energy Technology Data Exchange (ETDEWEB)

    Oliver, Lena; Peterson, Jenny; Wilhelmsen, Katarina [Swedish Defence Research Agency (FOI), Stockholm (Sweden)

    2006-03-15

    FOI, has performed a study on uranium conversion processes that are of importance in the production of different uranium compounds in the nuclear industry. The same conversion processes are of interest both when production of nuclear fuel and production of fissile material for nuclear weapons are considered. Countries that have nuclear weapons ambitions, with the intention to produce highly enriched uranium for weapons purposes, need some degree of uranium conversion capability depending on the uranium feed material available. This report describes the processes that are needed from uranium mining and milling to the different conversion processes for converting uranium ore concentrate to uranium hexafluoride. Uranium hexafluoride is the uranium compound used in most enrichment facilities. The processes needed to produce uranium dioxide for use in nuclear fuel and the processes needed to convert different uranium compounds to uranium metal - the form of uranium that is used in a nuclear weapon - are also presented. The production of uranium ore concentrate from uranium ore is included since uranium ore concentrate is the feed material required for a uranium conversion facility. Both the chemistry and principles or the different uranium conversion processes and the equipment needed in the processes are described. Since most of the equipment that is used in a uranium conversion facility is similar to that used in conventional chemical industry, it is difficult to determine if certain equipment is considered for uranium conversion or not. However, the chemical conversion processes where UF{sub 6} and UF{sub 4} are present require equipment that is made of corrosion resistant material.

  9. Issues in uranium availability

    International Nuclear Information System (INIS)

    Schanz, J.J. Jr.; Adams, S.S.; Gordon, R.L.

    1982-01-01

    The purpose of this publication is to show the process by which information about uranium reserves and resources is developed, evaluated and used. The following three papers in this volume have been abstracted and indexed for the Energy Data Base: (1) uranium reserve and resource assessment; (2) exploration for uranium in the United States; (3) nuclear power, the uranium industry, and resource development

  10. Australian uranium industry

    Energy Technology Data Exchange (ETDEWEB)

    Warner, R K

    1976-04-01

    Various aspects of the Australian uranium industry are discussed including the prospecting, exploration and mining of uranium ores, world supply and demand, the price of uranium and the nuclear fuel cycle. The market for uranium and the future development of the industry are described.

  11. Irradiated uranium reprocessing

    International Nuclear Information System (INIS)

    Gal, I.

    1961-12-01

    Task concerned with reprocessing of irradiated uranium covered the following activities: implementing the method and constructing the cell for uranium dissolving; implementing the procedure for extraction of uranium, plutonium and fission products from radioactive uranium solutions; studying the possibilities for using inorganic ion exchangers and adsorbers for separation of U, Pu and fission products

  12. Uranium processing and properties

    CERN Document Server

    2013-01-01

    Covers a broad spectrum of topics and applications that deal with uranium processing and the properties of uranium Offers extensive coverage of both new and established practices for dealing with uranium supplies in nuclear engineering Promotes the documentation of the state-of-the-art processing techniques utilized for uranium and other specialty metals

  13. Recovering uranium from phosphates

    Energy Technology Data Exchange (ETDEWEB)

    Bergeret, M [Compagnie de Produits Chimiques et Electrometallurgiques Pechiney-Ugine Kuhlmann, 75 - Paris (France)

    1981-06-01

    Processes for the recovery of the uranium contained in phosphates have today become competitive with traditional methods of working uranium sources. These new possibilities will make it possible to meet more rapidly any increases in the demand for uranium: it takes ten years to start working a new uranium deposit, but only two years to build a recovery plant.

  14. Uranium enrichment plans

    International Nuclear Information System (INIS)

    Gagne, R.W.; Thomas, D.C.

    1977-01-01

    The status of existing uranium enrichment contracts in the US is reviewed and expected natural uranium requirements for existing domestic uranium enrichment contracts are evaluated. Uncertainty in natural uranium requirements associated with requirements-type and fixed-commitment type contracts is discussed along with implementation of variable tails assay

  15. Uranium enrichment plans

    International Nuclear Information System (INIS)

    Thomas, D.C.; Gagne, R.W.

    1978-01-01

    The following topics are covered: the status of the Government's existing uranium enrichment services contracts, natural uranium requirements based on the latest contract information, uncertainty in predicting natural uranium requirements based on uranium enrichment contracts, and domestic and foreign demand assumed in enrichment planning

  16. All fiber cladding mode stripper with uniform heat distribution and high cladding light loss manufactured by CO2 laser ablation

    Science.gov (United States)

    Jebali, M. A.; Basso, E. T.

    2018-02-01

    Cladding mode strippers are primarily used at the end of a fiber laser cavity to remove high-power excess cladding light without inducing core loss and beam quality degradation. Conventional manufacturing methods of cladding mode strippers include acid etching, abrasive blasting or laser ablation. Manufacturing of cladding mode strippers using laser ablation consist of removing parts of the cladding by fused silica ablation with a controlled penetration and shape. We present and characterize an optimized cladding mode stripper design that increases the cladding light loss with a minimal device length and manufacturing time. This design reduces the localized heat generation by improving the heat distribution along the device. We demonstrate a cladding mode stripper written on a 400um fiber with cladding light loss of 20dB, with less than 0.02dB loss in the core and minimal heating of the fiber and coating. The manufacturing process of the designed component is fully automated and takes less than 3 minutes with a very high throughput yield.

  17. DECONTAMINATION OF ZIRCALOY SPENT FUEL CLADDING HULLS

    International Nuclear Information System (INIS)

    Rudisill, T; John Mickalonis, J

    2006-01-01

    The reprocessing of commercial spent nuclear fuel (SNF) generates a Zircaloy cladding hull waste which requires disposal as a high level waste in the geologic repository. The hulls are primarily contaminated with fission products and actinides from the fuel. During fuel irradiation, these contaminants are deposited in a thin layer of zirconium oxide (ZrO 2 ) which forms on the cladding surface at the elevated temperatures present in a nuclear reactor. Therefore, if the hulls are treated to remove the ZrO 2 layer, a majority of the contamination will be removed and the hulls could potentially meet acceptance criteria for disposal as a low level waste (LLW). Discard of the hulls as a LLW would result in significant savings due to the high costs associated with geologic disposal. To assess the feasibility of decontaminating spent fuel cladding hulls, two treatment processes developed for dissolving fuels containing zirconium (Zr) metal or alloys were evaluated. Small-scale dissolution experiments were performed using the ZIRFLEX process which employs a boiling ammonium fluoride (NH 4 F)/ammonium nitrate (NH 4 NO 3 ) solution to dissolve Zr or Zircaloy cladding and a hydrofluoric acid (HF) process developed for complete dissolution of Zr-containing fuels. The feasibility experiments were performed using Zircaloy-4 metal coupons which were electrochemically oxidized to produce a thin ZrO 2 layer on the surface. Once the oxide layer was in place, the ease of removing the layer using methods based on the two processes was evaluated. The ZIRFLEX and HF dissolution processes were both successful in removing a 0.2 mm (thick) oxide layer from Zircaloy-4 coupons. Although the ZIRFLEX process was effective in removing the oxide layer, two potential shortcomings were identified. The formation of ammonium hexafluorozirconate ((NH 4 ) 2 ZrF 6 ) on the metal surface prior to dissolution in the bulk solution could hinder the decontamination process by obstructing the removal of

  18. Uranium industry annual 1985

    International Nuclear Information System (INIS)

    1986-11-01

    This report consists of two major sections. The first addresses uranium raw materials activities and covers the following topics: exploration activities and expenditures, resources and reserves, mine production of uranium, production of uranium concentrate, and industry employment. The second major section is concerned with the following uranium marketing activities: uranium purchase commitments, uranium prices, procurement arrangements, uranium imports and exports, enrichment services, inventories, secondary market activities, utility market requirements, and related topics. A glossary and appendices are included to assist the reader in interpreting the substantial array of statistical data in this report and to provide background information about the survey

  19. Uranium industry framework

    International Nuclear Information System (INIS)

    Riley, K.

    2008-01-01

    The global uranium market is undergoing a major expansion due to an increase in global demand for uranium, the highest uranium prices in the last 20 years and recognition of the potential greenhouse benefits of nuclear power. Australia holds approximately 27% of the world's uranium resources (recoverable at under US$80/kg U), so is well placed to benefit from the expansion in the global uranium market. Increasing exploration activity due to these factors is resulting in the discovery and delineation of further high grade uranium deposits and extending Australia's strategic position as a reliable and safe supplier of low cost uranium.

  20. Rheological evaluation of pretreated cladding removal waste

    International Nuclear Information System (INIS)

    McCarthy, D.; Chan, M.K.C.; Lokken, R.O.

    1986-01-01

    Cladding removal waste (CRW) contains concentrations of transuranic (TRU) elements in the 80 to 350 nCi/g range. This waste will require pretreatment before it can be disposed of as glass or grout at Hanford. The CRW will be pretreated with a rare earth strike and solids removal by centrifugation to segregate the TRU fraction from the non-TRU fraction of the waste. The centrifuge centrate will be neutralized with sodium hydroxide. This neutralized cladding removal waste (NCRW) is expected to be suitable for grouting. The TRU solids removed by centrifugation will be vitrified. The goal of the Rheological Evaluation of Pretreated Cladding Removal Waste Program was to evaluate those rheological and transport properties critical to assuring successful handling of the NCRW and TRU solids streams and to demonstrate transfers in a semi-prototypic pumping environment. This goal was achieved by a combination of laboratory and pilot-scale evaluations. The results obtained during these evaluations were correlated with classical rheological models and scaled-up to predict the performance that is likely to occur in the full-scale system. The Program used simulated NCRW and TRU solid slurries. Rockwell Hanford Operations (Rockwell) provided 150 gallons of simulated CRW and 5 gallons of simulated TRU solid slurry. The simulated CRW was neutralized by Pacific Northwest Laboratory (PNL). The physical and rheological properties of the NCRW and TRU solid slurries were evaluated in the laboratory. The properties displayed by NCRW allowed it to be classified as a pseudoplastic or yield-pseudoplastic non-Newtonian fluid. The TRU solids slurry contained very few solids. This slurry exhibited the properties associated with a pseudoplastic non-Newtonian fluid

  1. PCI resistant light water reactor fuel cladding

    International Nuclear Information System (INIS)

    Foster, J.P.; Sabol, G.P.

    1988-01-01

    A tubular nuclear fuel element cladding tube is described, the fuel element cladding tube forming the entire fuel element cladding and consisting of: a single continuous wall, the single continuous wall consisting of a single alloy selected from the group consisting of zirconium base alloys, A, B, C, D, and E; the single continuous wall characterized by a cold worked and stress relieved microstructure throughout; wherein the zirconium base alloy A contains 0.2 - 0.6 w/o Sn, 0.03 - 0.11 w/o sum of Fe and Cr, section 600 ppm O and section 1500 ppm total impurities; the zirconium base alloy B contains 0.1 - 0.6 w/oo Sn, 0.04 - 0.24 w/o Fe, 0.05 - 0.15 w/o Cr, section 0.08 w/o Ni, section 600 ppm O and section 1500 ppm total impurities; the zirconium base alloy C contains 1.2 - 1.7 w/o Sn, 0.04 - 0.24 w/o Fe, 0.05 - 0.15 w/o Cr, section 0.08 w/o Ni, section 600 ppm O, and section 1500 ppm total impurities; the zirconium base alloy D contains 0.15 - 0.6 w/o Sn, 0.15 - 0.5 w/o Fe, section 600 ppm O, and section 1500 ppm total impurities; and the zirconium base alloy E contains 0.4 - 0.6 w/o Sn, 0.1 - 0.3 w/o Fe, 0.03 - 0.07 w/o Ni, section 600 ppm O, and section 1500 ppm total impurities

  2. Analysis of pellet cladding mechanical interaction using computational simulation

    Energy Technology Data Exchange (ETDEWEB)

    Berretta, José R.; Suman, Ricardo B.; Faria, Danilo P.; Rodi, Paulo A., E-mail: jose.berretta@marinha.mil.br [Centro Tecnológico da Marinha em São Paulo (CTMSP), São Paulo, SP (Brazil); Giovedi, Claudia, E-mail: claudia.giovedi@labrisco.usp.br [Universidade de Sao Paulo (LabRisco/USP), São Paulo, SP (Brazil). Laboratório de Análise, Avaliação e Gerenciamento de Riscos

    2017-07-01

    During the operation of Pressurized Water Reactors (PWR), specifically under power transients, the fuel pellet experiences many phenomena, such as swelling and thermal expansion. These dimensional changes in the fuel pellet can enable occurrence of contact it and the cladding along the fuel rod. Thus, pellet cladding mechanical interaction (PCMI), due this contact, induces stress increase at the contact points during a period, until the accommodation of the cladding to the stress increases. This accommodation occurs by means of the cladding strain, which can produce failure, if the fuel rod deformation is permanent or the burst limit of the cladding is reached. Therefore, the mechanical behavior of the cladding during the occurrence of PCMI under power transients shall be investigated during the fuel rod design. Considering the Accident Tolerant Fuel program which aims to develop new materials to be used as cladding in PWR, one important design condition to be evaluated is the cladding behavior under PCMI. The purpose of this paper is to analyze the effects of the PCMI on a typical PWR fuel rod geometry with stainless steel cladding under normal power transients using computational simulation (ANSYS code). The PCMI was analyzed considering four geometric situations at the region of interaction between pellet and cladding. The first case, called “perfect fuel model” was used as reference for comparison. In the second case, it was considered the occurrence of a pellet crack with the loss of a chip. The goal for the next two cases was that a pellet chip was positioned into the gap of pellet-cladding, in the situations described in the first two cases. (author)

  3. Determination of uranium metal concentration in irradiated fuel storage basin sludge using selective dissolution

    International Nuclear Information System (INIS)

    Delegard, C.H.; Sinkov, S.I.; Chenault, J.W.; Schmidt, A.J.; Pool, K.N.; Welsh, T.L.

    2014-01-01

    Irradiated uranium metal fuel was stored underwater in the K East and K West storage basins at the US Department of Energy Hanford Site. The uranium metal under damaged cladding reacted with water to generate hydrogen gas, uranium oxides, and spalled uranium metal particles which intermingled with other particulates to form sludge. While the fuel has been removed, uranium metal in the sludge remains hazardous. An expeditious routine method to analyze 0.03 wt% uranium metal in the presence of >30 wt% total uranium was needed to support safe sludge management and processing. A selective dissolution method was designed based on the rapid uranium oxide dissolution but very low uranium metal corrosion rates in hot concentrated phosphoric acid. The uranium metal-bearing heel from the phosphoric acid step then is rinsed before the uranium metal is dissolved in hot concentrated nitric acid for analysis. Technical underpinnings of the selective dissolution method, including the influence of sludge components, were investigated to design the steps and define the reagents, quantities, concentrations, temperatures, and times within the selective dissolution analysis. Tests with simulant sludge proved the technique feasible. Tests with genuine sludge showed a 0.0028 ± 0.0037 wt% (at one standard deviation) uranium metal analytical background, a 0.011 wt% detection limit, and a 0.030 wt% quantitation limit in settled (wet) sludge. In tests using genuine K Basin sludge spiked with uranium metal at concentrations above the 0.030 wt% ± 25 % (relative) quantitation limit, uranium metal recoveries averaged 99.5 % with a relative standard deviation of 3.5 %. (author)

  4. Reduction of uranium hexafluoride to uranium tetrafluoride

    International Nuclear Information System (INIS)

    Chang, I.S.; Do, J.B.; Choi, Y.D.; Park, M.H.; Yun, H.H.; Kim, E.H.; Kim, Y.W.

    1982-01-01

    The single step continuous reduction of uranium hexafluoride (UF 6 ) to uranium tetrafluoride (UF 4 ) has been investigated. Heat required to initiate and maintain the reaction in the reactor is supplied by the highly exothermic reaction of hydrogen with a small amount of elemental fluorine which is added to the uranium hexafluoride stream. When gases uranium hexafluoride and hydrogen react in a vertical monel pipe reactor, the green product, UF 4 has 2.5g/cc in bulk density and is partly contaminated by incomplete reduction products (UF 5 ,U 2 F 9 ) and the corrosion product, presumably, of monel pipe of the reactor itself, but its assay (93% of UF 4 ) is acceptable for the preparation of uranium metal with magnesium metal. Remaining problems are the handling of uranium hexafluoride, which is easily clogging the flowmeter and gas feeding lines because of extreme sensitivity toward moisture, and a development of gas nozzel for free flow of uranium hexafluoride gas. (Author)

  5. Alloys of uranium and aluminium with low aluminium content

    International Nuclear Information System (INIS)

    Cabane, G.; Englander, M.; Lehmann, J.

    1955-01-01

    Uranium, as obtained after spinning in phase γ, presents an heterogeneous structure with large size grains. The anisotropic structure of the metal leads to an important buckling and surface distortion of the fuel slug which is incompatible with its tubular cladding for nuclear fuel uses. Different treatments have been made to obtain an isotropic structure presenting high thermal stability (laminating, hammering and spinning in phase α) without success. Alloys of uranium and aluminium with low aluminium content present important advantage in respect of non allied uranium. The introduction of aluminium in the form of intermetallic compound (UAl 2 ) gives a better resistance to thermal fatigue. Alloys obtained from raw casting present an improved buckling and surface distortion in respect of pure uranium. This improvement is obtained with uranium containing between 0,15 and 0,5 % of aluminium. An even more improvement in thermal stability is obtained by thermal treatments of these alloys. These new characteristics are explained by the fine dispersion of the UAl 2 particles in uranium. The results after treatments obtained from an alloy slug containing 0,4 % of aluminium show no buckling or surface distortion and no elongation. (M.P.)

  6. Process for producing clad superconductive materials

    International Nuclear Information System (INIS)

    Cass, R.B.; Ott, K.C.; Peterson, D.E.

    1992-01-01

    This patent describes a process for fabricating superconducting composite wire. It comprises placing a superconductive precursor admixture capable of undergoing self propagating combustion in stoichiometric amounts sufficient to form a superconductive product within an oxygen-porous metal tube; sealing one end of the tube; igniting the superconductive precursor admixture whereby the superconductive precursor admixture endburns along the length of the admixture; and cross-section reducing the tube at a rate substantially equal to the rate of burning of the superconductive precursor admixture and at a point substantially planar with the burnfront of the superconductive precursor mixture, whereby a clad superconductive product is formed in situ

  7. Method and device for weld deposit cladding

    International Nuclear Information System (INIS)

    Barger, J.J.

    1977-01-01

    In order to get weld beads of good quality, uniform thickness and faultless transition regions between neighboring beads in weld deposit cladding of metallic workpoeces, it is proposed to use a device in which the electromagnets are arranged adjacent to th zone of molten welding powder and molten metal besides having got suitable supplies for applying the welding powder, the polarity of the magnets being chosen in such a way that the lines of flux between the poles are counteracting the lines of flux surrounding the electrode band because of the welding current. Several variants of arranging the electrodes are presented in detail. (UWI) [de

  8. Zircaloy cladding degradation under repository conditions

    International Nuclear Information System (INIS)

    Santanam, L.; Raghavan, S.; Chin, B.A.

    1990-12-01

    Creep, a potential degradation mechanism of Zircaloy cladding after repository disposal of spent nuclear fuel, has been investigated. The deformation and fracture map methodology has been used to predict maximum allowable initial storage temperatures to achieve a thousand year life without rupture as a function of spent-fuel history. Maximum allowable temperatures are 340 degree C (613 K) for typically stressed rods (70--100 MPa) and 300 degree C (573 K) for highly stressed rods (140--160 MPa). 10 refs., 2 figs

  9. Termination of plastic-clad fiber

    International Nuclear Information System (INIS)

    Nance, W.R.

    1982-03-01

    Optical waveguides are ideal in a nuclear weapon environment because of their resistance to electromagnetic interference. Of the fibers on today's market, plastic-clad silica (PCS) is the most radiation resistant and therfore the best choice. Because terminating PCS is complex, this paper attemps to address the major problems associated with these terminations including selecting the proper connector and optimizing the terminating procedures. The sources of losses in the connectors are summarized and typical loss values are given for four connectors which were tested

  10. CO2 laser cladding of VERSAlloyTM on carbon steel with powder feeding

    International Nuclear Information System (INIS)

    Kim, Jae-Do; Kweon, Jin-Wook

    2007-01-01

    Laser cladding processing with metal powder feeding has been experimented on carbon steel with VERSAlloy TM . A special device for the metal powder feeding was designed and manufactured. By adopting proper cladding parameters, good clad layers and sound metallurgical bonding with the base metal were obtained. Analysis indicates that the micro hardness of clad layer and the heat-affected zone increased with increasing of cladding speed. The experimental results showed that VERSAlloy TM cladded well with carbon steel

  11. Uranium - the world picture

    International Nuclear Information System (INIS)

    Silver, J.M.; Wright, W.J.

    1976-01-01

    The world resources of uranium and the future demand for uranium are discussed. The amount of uranium available depends on the price which users are prepared to pay for its recovery. As the price is increased, there is an incentive to recover uranium from lower grade or more difficult deposits. In view of this, attention is drawn to the development of the uranium industry in Australias

  12. Alloys of uranium and aluminium with low aluminium content; Alliages uranium-aluminium a faible teneur en aluminium

    Energy Technology Data Exchange (ETDEWEB)

    Cabane, G; Englander, M; Lehmann, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1955-07-01

    Uranium, as obtained after spinning in phase {gamma}, presents an heterogeneous structure with large size grains. The anisotropic structure of the metal leads to an important buckling and surface distortion of the fuel slug which is incompatible with its tubular cladding for nuclear fuel uses. Different treatments have been made to obtain an isotropic structure presenting high thermal stability (laminating, hammering and spinning in phase {alpha}) without success. Alloys of uranium and aluminium with low aluminium content present important advantage in respect of non allied uranium. The introduction of aluminium in the form of intermetallic compound (UAl{sub 2}) gives a better resistance to thermal fatigue. Alloys obtained from raw casting present an improved buckling and surface distortion in respect of pure uranium. This improvement is obtained with uranium containing between 0,15 and 0,5 % of aluminium. An even more improvement in thermal stability is obtained by thermal treatments of these alloys. These new characteristics are explained by the fine dispersion of the UAl{sub 2} particles in uranium. The results after treatments obtained from an alloy slug containing 0,4 % of aluminium show no buckling or surface distortion and no elongation. (M.P.)

  13. 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

  14. Characterisation of AGR fuel cladding alloy using secondary ion mass spectrometry

    International Nuclear Information System (INIS)

    Allen, G.C.; Sparry, R.P.; Wild, R.K.

    1987-08-01

    Uranium dioxide fuel used in the Advanced Gas Cooled Reactor (AGR) is contained in a ribbed can of 20wt%Cr/25wt%Ni/Nb stabilised steel. Laboratory circumstances, spall during thermal cycling. To date it has been difficult to identify active material originating from the oxidation product of the cladding alloy in the cooling circuit. In an attempt to solve this problem we have set out to characterise fully a sample of oxide from this source and work is in progress to obtain suitable oxide samples from the surface of a 20%Cr/25%Ni/Nb stainless steel. In view of its high sensitivity and the ability to obtain chemical information from relatively small areas we have sought to use Secondary Ion Mass Spectroscopy (SIMS). (author)

  15. Study of reactions between nuclear fuel and cladding (316 stainless steel) in reactors. Influence of oxygen

    International Nuclear Information System (INIS)

    Otter, Monique.

    1980-12-01

    We have studied oxidation of 316 steel in close contact with oxides (Usub(0,74)Pusub(0,26)O 2 or UO 2 ), the stoichiometry of oxygen ranging from 2.00 to 2.5. Experiments are carried out either in a closed isothermal system or in an opened isothermal system with a fixed oxygen potential of uranium oxide. We have realized a potentiostatic device using a solid state electrotyte galvanic cell. In a closed system, the sensitized austenitic steel shows intergranular and volume oxidation probably enhanced by migration of steel components towards the fuel. Evidence of the usefulness of passivation have been obtained. We conclude that in a fast reactor sensitized cladding steel is oxydized by the constant potential of oxygen of UPuO 2 . Deposits observed in fuel can be explain by evaporation and cyclic transport phenomena that can be differents from VAN-ARKEL mechanism taking place through fission products [fr

  16. Long-range plasmonic waveguides with hyperbolic cladding

    DEFF Research Database (Denmark)

    Babicheva, Viktoriia E.; Shalaginov, Mikhail Y.; Ishii, Satoshi

    2015-01-01

    waveguides. We show that the proposed structures support long-range surface plasmon modes, which exist when the permittivity of the core matches the transverse effective permittivity component of the metamaterial cladding. In this regime, the surface plasmon polaritons of each cladding layer are strongly...

  17. Embedded cladding surface thermocouples on Zircaloy-sheathed heater rods

    International Nuclear Information System (INIS)

    Wilkins, S.C.

    1977-06-01

    Titanium-sheathed Type K thermocouples embedded in the cladding wall of zircaloy-sheathed heater rods are described. These thermocouples constitute part of a program intended to characterize the uncertainty of measurements made by surface-mounted cladding thermocouples on nuclear fuel rods. Fabrication and installation detail, and laboratory testing of sample thermocouple installations are included

  18. Corrosion behavior of duplex and reference cladding in NPP Grohnde

    International Nuclear Information System (INIS)

    Besch, O.A.; Yagnik, S.K.; Eucken, C.M.; Bradley, E.R.

    1996-01-01

    The Nuclear Fuel Industry Research (NFIR) Group undertook a lead test assembly (LTA) program in NPP Grohnde PWR in Germany to assess the corrosion performance of duplex and reference cladding. Two identical 16 by 16 LTAs, each containing 32 peripheral test rods, completed four reactor cycles, reaching a peak rod burnup of 46 MWd/kgU. The results from poolside examinations performed at the end of each cycle, together with power histories and coolant chemistry, are reported. Five different cladding materials were characterized during fabrication. The corrosion performance of the cladding materials was tracked in long-term tests in high-pressure, high-temperature autoclaves. The relative ranking of corrosion behavior in such tests corresponded well with the in-reactor corrosion performance. The extent and distribution of hydriding in duplex and reference specimens during the autoclave testing has been characterized. The in-reactor corrosion data indicate that the low-tin Zircaloy-4 reference cladding, R2, had an improved corrosion resistance compared to high-tin Zircaloy-4 reference cladding, R1. Two types of duplex cladding, D1 (Zr-2.5% Nb) and D2 (Zr-0.4% Fe-0.5% Sn), showed an even further improvement in corrosion resistance compared to R2 cladding. The third duplex cladding, D3 (Zr-4 + 1.0% Nb), had significantly less corrosion resistance, which was inferior to R1. The in-reactor and out-reactor corrosion performances have been ranked

  19. Cladding of Advanced Al Alloys Employing Friction Stir Welding

    NARCIS (Netherlands)

    van der Stelt, A.A.; Bor, Teunis Cornelis; Geijselaers, Hubertus J.M.; Akkerman, Remko; van den Boogaard, Antonius H.

    2013-01-01

    In this paper an advanced solid state cladding process, based on Friction Stir Welding, is presented. The Friction Surface Cladding (FSC) technology enables the deposition of a solid-state coating using filler material on a substrate with good metallurgical bonding. A relatively soft AA1050 filler

  20. Ultimate disposition of aluminum clad spent nuclear fuel in the United States

    International Nuclear Information System (INIS)

    Messick, C.E.; Clark, W.D.; Clapper, M.; Mustin, T.P.

    2001-01-01

    Treatment and disposition of spent nuclear fuel (SNF) in the United States has changed significantly over the last decade due to change in world climate associated with nuclear material. Chemical processing of aluminum based SNF is ending and alternate disposition paths are being developed that will allow for the ultimate disposition of the enriched uranium in this SNF. Existing inventories of aluminum based SNF are currently being stored primarily in water-filled basins at the Savannah River Site (SRS) while these alternate disposition paths are being developed and implemented. Nuclear nonproliferation continues to be a worldwide concern and it is causing a significant influence on the development of management alternatives for SNF. SRS recently completed an environmental impact statement for the management of aluminum clad SNF that selects alternatives for all of the fuels in inventory. The U.S. Department of Energy and SRS are now implementing a dual strategy of processing small quantities of 'problematic' SNF while developing an alternative technology to dispose of the remaining aluminum clad SNF in the proposed monitored geologic repository. (author)

  1. Natural uranium

    International Nuclear Information System (INIS)

    Ammerich, Marc; Frot, Patricia; Gambini, Denis-Jean; Gauron, Christine; Moureaux, Patrick; Herbelet, Gilbert; Lahaye, Thierry; Pihet, Pascal; Rannou, Alain

    2014-08-01

    This sheet belongs to a collection which relates to the use of radionuclides essentially in unsealed sources. Its goal is to gather on a single document the most relevant information as well as the best prevention practices to be implemented. These sheets are made for the persons in charge of radiation protection: users, radioprotection-skill persons, labor physicians. Each sheet treats of: 1 - the radio-physical and biological properties; 2 - the main uses; 3 - the dosimetric parameters; 4 - the measurement; 5 - the protection means; 6 - the areas delimitation and monitoring; 7 - the personnel classification, training and monitoring; 8 - the effluents and wastes; 9 - the authorization and declaration administrative procedures; 10 - the transport; and 11 - the right conduct to adopt in case of incident or accident. This sheet deals specifically with natural uranium

  2. Instrument for measuring fuel cladding strain

    International Nuclear Information System (INIS)

    Billeter, T.R.

    1976-01-01

    Development work to provide instrumentation for the continuous measurement of strain of material specimens such as nuclear fuel cladding has shown that a microwave sensor and associated instrumentation hold promise. The cylindrical sensor body enclosing the specimen results in a coaxial resonator absorbing microwave energy at frequencies dependent upon the diameter of the specimen. Diametral changes of a microinch can be resolved with use of the instrumentation. Very reasonable values of elastic strain were measured at 75 0 F and 1000 0 F for an internally pressurized 20 percent C.W. 316 stainless steel specimen simulating nuclear fuel cladding. The instrument also indicated the creep strain of the same specimen pressurized at 6500 psi and at a temperature of 1000 0 F for a period of 700 hours. Although the indicated strain appears greater than actual, the sensor/specimen unit experienced considerable oxidation even though an inert gas purge persisted throughout the test duration. By monitoring at least two modes of resonance, the measured strain was shown to be nearly independent of sensor temperature. To prevent oxidation, a second test was performed in which the specimen/sensor units were contained in an evacuated enclosure. The strain of the two prepressurized specimens as indicated by the microwave instrumentation agreed very closely with pre- and post-test measurements obtained with use of a laser interferometer

  3. Material Selection for Accident Tolerant Fuel Cladding

    International Nuclear Information System (INIS)

    Pint, Bruce A.; Terrani, Kurt A.; Yamamoto, Yukinori; Snead, Lance Lewis

    2015-01-01

    Alternative cladding materials to Zr-based alloys are being investigated for accident tolerance, which can be defined as > 100X improvement (compared to Zr-based alloys) in oxidation resistance to steam or steam-H 2 environments at ≥1473 K (1200°C) for short times. After reviewing a wide range of candidates, current steam oxidation testing is being conducted on Mo, MAX phases and FeCrAl alloys. Recently reported low mass losses for Mo in steam at 800°C could not be reproduced. Both FeCrAl and MAX phase Ti 2 AlC form a protective alumina scale in steam. However, commercial Ti 2 AlC that was not single phase, formed a much thicker oxide at 1200°C in steam and significant TiO 2 , and therefore Ti 2 AlC may be challenging to form as a cladding or a coating. Alloy development for FeCrAl is seeking to maintain its steam oxidation resistance to 1475°C, while reducing its Cr content to minimize susceptibility to irradiation-assisted α' formation. The composition effects and critical limits to retaining protective scale formation at > 1400°C are still being evaluated.

  4. Chemical interaction of fuel and cladding tubes

    International Nuclear Information System (INIS)

    Kirihara, Tomoo; Yamawaki, Michio; Obata, Naomi; Handa, Muneo.

    1983-01-01

    It was attempted to take up the behavior of nuclear fuel in cores and summarize it by the expert committee on the irradiation behavior of nuclear fuel from fiscal 1978 to fiscal 1980 from the following viewpoints. The behavior of nuclear fuel in cores has been treated separately according to each reactor type, accordingly this point is reconsidered. The clearly understood points and the uncertain points are discriminated. It is made more easily understandable for people in other fields of atomic energy. This report is that of the group on the chemical interaction, and the first report of this committee. The chemical interaction as the behavior of fuel in cores is in the unseparable relation to the mechanical interaction, but this relation is not included in this report. The chemical interaction of fuel and cladding tubes under irradiation shows different phenomena in LWRs and FBRs, and is called SCC and FCC, respectively. But this point of causing the difference must be understood to grasp the behavior of fuel. The mutual comparison of oxide fuels for FBRs and LWRs, the stress corrosion cracking of zircaloy tubes, and fuel-cladding chemical interaction in FBRs are reported. (Kako, I.)

  5. Material Selection for Accident Tolerant Fuel Cladding

    Energy Technology Data Exchange (ETDEWEB)

    Pint, Bruce A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrani, Kurt A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Yamamoto, Yukinori [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Snead, Lance Lewis [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-01-01

    Alternative cladding materials to Zr-based alloys are being investigated for accident tolerance, which can be defined as > 100X improvement (compared to Zr-based alloys) in oxidation resistance to steam or steam-H2 environments at ≥ 1200°C for short times. After reviewing a wide range of candidates, current steam oxidation testing is being conducted on Mo, MAX phases and FeCrAl alloys. Recently reported low mass losses for Mo in steam at 800°C could not be reproduced. Both FeCrAl and MAX phase Ti2AlC form a protective alumina scale in steam. However, commercial Ti2AlC that was not single phase, formed a much thicker oxide at 1200°C in steam and significant TiO2, and therefore Ti2AlC may be challenging to form as a cladding or a coating. Alloy development for FeCrAl is seeking to maintain its steam oxidation resistance to 1475°C, while reducing its Cr content to minimize susceptibility to irradiation-assisted α´ formation. The composition effects and critical limits to retaining protective scale formation at > 1400°C are still being evaluated.

  6. Fuel cladding tube leak detection device

    International Nuclear Information System (INIS)

    Naito, Makoto.

    1992-01-01

    The device of the present invention can detect even a minute leakage or a continuous leakage during reactor operation. That is, the device of the present invention comprises a detector for analyzing nuclides of gases incorporated in a gas waste processing system, and a calculation device connected to the detector and detecting leakage from a fuel cladding tube by calculation for variation coefficient of long-life nuclides. By using theses devices, radioactivity contained in gases incorporated in the gas waste processing system is analyzed for the nuclides. Among the analized nuclides, if the amount of the long-life nuclides exceeds a predetermined value, it is judged as leakage of the fuel cladding tube. For example, the long-life nuclides include Xe-133. The device of the present invention can certainly detect occurrence of leakage even when it is minute or continues leakage. Accordingly, countermeasures can be taken in an early stage, thereby enabling to contribute improvement for the safety of a nuclear power plant. (I.S.)

  7. The ballooning of fuel cladding tubes: theory and experiment

    International Nuclear Information System (INIS)

    Shewfelt, R.S.W.

    1988-01-01

    Under some conditions, fuel clad ballooning can result in considerable strain before rupture. If ballooning were to occur during a loss-of-coolant accident (LOCA), the resulting substantial blockage of the sub-channel would restrict emergency core cooling. However, circumferential temperature gradients that would occur during a LOCA may significantly limit the average strain at failure. Understandably, the factors that control ballooning and rupture of fuel clad are required for the analysis of a LOCA. Considerable international effort has been spent on studying the deformation of Zircaloy fuel cladding under conditions that would occur during a LOCA. This effort has established a reasonable understanding of the factors that control the ballooning, failure time, and average failure strain of fuel cladding. In this paper, both the experimental and theoretical studies of the fuel clad ballooning are reviewed. (author)

  8. Cladding modes of optical fibers: properties and applications

    International Nuclear Information System (INIS)

    Ivanov, Oleg V; Nikitov, Sergei A; Gulyaev, Yurii V

    2006-01-01

    One of the new methods of fiber optics uses cladding modes for controlling propagation of radiation in optical fibers. This paper reviews the results of studies on the propagation, excitation, and interaction of cladding modes in optical fibers. The resonance between core and cladding modes excited by means of fiber Bragg gratings, including tilted ones, is analyzed. Propagation of cladding modes in microstructured fibers is considered. The most frequently used method of exciting cladding modes is described, based on the application of long-period fiber gratings. Examples are presented of long-period gratings used as sensors and gain equalizers for fiber amplifiers, as well as devices for coupling light into and out of optical fibers. (instruments and methods of investigation)

  9. Cladding failure probability modeling for risk evaluations of fast reactors

    International Nuclear Information System (INIS)

    Mueller, C.J.; Kramer, J.M.

    1987-01-01

    This paper develops the methodology to incorporate cladding failure data and associated modeling into risk evaluations of liquid metal-cooled fast reactors (LMRs). Current US innovative designs for metal-fueled pool-type LMRs take advantage of inherent reactivity feedback mechanisms to limit reactor temperature increases in response to classic anticipated-transient-without-scram (ATWS) initiators. Final shutdown without reliance on engineered safety features can then be accomplished if sufficient time is available for operator intervention to terminate fission power production and/or provide auxiliary cooling prior to significant core disruption. Coherent cladding failure under the sustained elevated temperatures of ATWS events serves as one indicator of core disruption. In this paper we combine uncertainties in cladding failure data with uncertainties in calculations of ATWS cladding temperature conditions to calculate probabilities of cladding failure as a function of the time for accident recovery

  10. Reduction of Bragg-grating-induced coupling to cladding modes

    DEFF Research Database (Denmark)

    Berendt, Martin Ole; Bjarklev, Anders Overgaard; Soccolich, C.E.

    1999-01-01

    gratings in a depressed-cladding fiber are compared with simulations. The model gives good agreement with the measured transmission spectrum and accounts for the pronounced coupling to asymmetrical cladding modes, even when the grating is written with the smallest possible blaze. The asymmetry causing...... this is accounted for by the unavoidable attenuation of the UV light. It is found for the considered fiber designs that a high numerical-aperture fiber increases the spectral separation between the Bragg resonance and the onset of cladding-mode losses. A depressed-cladding fiber reduces the coupling strength......We discuss fiber designs that have been suggested for the reduction of Bragg-grating induced coupling to cladding modes. The discussion is based on a theoretical approach that includes the effect of asymmetry in the UV-induced index grating, made by UV-side writing. Experimental results from...

  11. Study on process of laser cladded nuclear valve parts

    International Nuclear Information System (INIS)

    Zhang Chunliang

    2000-01-01

    The microstructure and performances of the Co-base alloy coatings that are formed by laser cladding, plasma spurt welding and arc surfacing on the nuclear valve-sealing surface have been studied and compared. The combination costs of laser cladding, plasma spurt welding and arc, surfacing have been analyzed and compared. The results showed that the laser cladding processing has the advantages of high efficiency, low energy cost, a little machining allowance, high rate of finished products and low combination cost, compared with plasma spurt welding processing and arc surfacing processing. The laser cladding technology can improve the qualities of nuclear valve parts and increase their service life. Therefore, the laser cladding processing is a new technology with developing potential

  12. Cladding failure probability modeling for risk evaluations of fast reactors

    International Nuclear Information System (INIS)

    Mueller, C.J.; Kramer, J.M.

    1987-01-01

    This paper develops the methodology to incorporate cladding failure data and associated modeling into risk evaluations of liquid metal-cooled fast reactors (LMRs). Current U.S. innovative designs for metal-fueled pool-type LMRs take advantage of inherent reactivity feedback mechanisms to limit reactor temperature increases in response to classic anticipated-transient-without-scram (ATWS) initiators. Final shutdown without reliance on engineered safety features can then be accomplished if sufficient time is available for operator intervention to terminate fission power production and/or provide auxiliary cooling prior to significant core disruption. Coherent cladding failure under the sustained elevated temperatures of ATWS events serves as one indicator of core disruption. In this paper we combine uncertainties in cladding failure data with uncertainties in calculations of ATWS cladding temperature conditions to calculate probabilities of cladding failure as a function of the time for accident recovery. (orig.)

  13. Test system to simulate transient overpower LMFBR cladding failure

    International Nuclear Information System (INIS)

    Barrus, H.G.; Feigenbutz, L.V.

    1981-01-01

    One of the HEDL programs has the objective to experimentally characterize fuel pin cladding failure due to cladding rupture or ripping. A new test system has been developed which simulates a transient mechanically-loaded fuel pin failure. In this new system the mechanical load is prototypic of a fuel pellet rapidly expanding against the cladding due to various causes such as fuel thermal expansion, fuel melting, and fuel swelling. This new test system is called the Fuel Cladding Mechanical Interaction Mandrel Loading Test (FCMI/MLT). The FCMI/MLT test system and the method used to rupture cladding specimens very rapidly to simulate a transient event are described. Also described is the automatic data acquisition and control system which is required to control the startup, operation and shutdown of the very fast tests, and needed to acquire and store large quantities of data in a short time

  14. Polarization characteristics of double-clad elliptical fibers.

    Science.gov (United States)

    Zhang, F; Lit, J W

    1990-12-20

    A scalar variational analysis based on a Gaussian approximation of the fundamental mode of a double-clad elliptical fiber with a depressed inner cladding is studied. The polarization properties and graphic results are presented; they are given in terms of three parameters: the ratio of the major axis to the minor axis of the core, the ratio of the inner cladding major axis to the core major axis, and the difference between the core index and the inner cladding index. The variations of both the spot size and the field intensity with core ellipticity are examined. It is shown that high birefringence and dispersion-free orthogonal polarization modes can be obtained within the single-mode region and that the field intensity distribution may be more confined to the fiber center than in a single-clad elliptical fiber.

  15. Uranium management activities

    International Nuclear Information System (INIS)

    Jackson, D.; Marshall, E.; Sideris, T.; Vasa-Sideris, S.

    2001-01-01

    One of the missions of the Department of Energy's (DOE) Oak Ridge Office (ORO) has been the management of the Department's uranium materials. This mission has been accomplished through successful integration of ORO's uranium activities with the rest of the DOE complex. Beginning in the 1980's, several of the facilities in that complex have been shut down and are in the decommissioning process. With the end of the Cold War, the shutdown of many other facilities is planned. As a result, inventories of uranium need to be removed from the Department facilities. These inventories include highly enriched uranium (HEU), low enriched uranium (LEU), normal uranium (NU), and depleted uranium (DU). The uranium materials exist in different chemical forms, including metals, oxides, solutions, and gases. Much of the uranium in these inventories is not needed to support national priorities and programs. (author)

  16. Application of Coating Technology for Accident Tolerant Fuel Cladding

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyun-Gil; Kim, Il-Hyun; Jung, Yang-Il; Park, Dong-Jun; Park, Jeong-Yong; Koo, Yang-Hyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    To commercialize the ATF cladding concepts, various factors are considered, such as safety under normal and accident conditions, economy for the fuel cycle, and developing development challenges, and schedule. From the proposed concepts, it is known that the cladding coating, FeCrAl alloy, and Zr-Mo claddings are considered as a near/mid-term application, whereas the SiC material is considered as a long-term application. Among them, the benefit of cladding coating on Zr-based alloys is the fuel cycle economy regarding the manufacturing, neutron cross section, and high tritium permeation characteristics. However, the challenge of cladding coating on Zr-based alloys is the lower oxidation resistance and mechanical strength at high-temperature than other concepts. Another important point is the adhesion property between the Zr-based alloy and coating materials. As an improved coating technology compared to a previous study, a 3D laser coating technology supplied with Cr powders is considered to make a coated cladding because it is possible to make a coated layer on the tubular cladding surface by controlling the 3-diminational axis. We are systematically studying the laser beam power, inert gas flow, cooling of the cladding tube, and powder control as key points to develop 3D laser coating technology. After Cr-coating on the Zr-based cladding, ring compression and ring tensile tests were performed to evaluate the adhesion property between a coated layer and Zr-based alloy tube at room temperature (RT), and a high-temperature oxidation test was conducted to evaluate the oxidation behavior at 1200 .deg. C of the coated tube samples. A 3D laser coating method supplied with Cr powders was developed to decrease the high-temperature oxidation rate in a steam environment through a systematic study for various coating parameters, and a Cr-coated Zircaloy-4 cladding tube of 100 mm in length to the axial direction can be successfully manufactured.

  17. Uranium industry annual 1998

    International Nuclear Information System (INIS)

    1999-01-01

    The Uranium Industry Annual 1998 (UIA 1998) provides current statistical data on the US uranium industry's activities relating to uranium raw materials and uranium marketing. It contains data for the period 1989 through 2008 as collected on the Form EIA-858, ''Uranium Industry Annual Survey.'' Data provides a comprehensive statistical characterization of the industry's activities for the survey year and also include some information about industry's plans and commitments for the near-term future. Data on uranium raw materials activities for 1989 through 1998, including exploration activities and expenditures, EIA-estimated reserves, mine production of uranium, production of uranium concentrate, and industry employment, are presented in Chapter 1. Data on uranium marketing activities for 1994 through 2008, including purchases of uranium and enrichment services, enrichment feed deliveries, uranium fuel assemblies, filled and unfilled market requirements, and uranium inventories, are shown in Chapter 2. The methodology used in the 1998 survey, including data edit and analysis, is described in Appendix A. The methodologies for estimation of resources and reserves are described in Appendix B. A list of respondents to the ''Uranium Industry Annual Survey'' is provided in Appendix C. The Form EIA-858 ''Uranium Industry Annual Survey'' is shown in Appendix D. For the readers convenience, metric versions of selected tables from Chapters 1 and 2 are presented in Appendix E along with the standard conversion factors used. A glossary of technical terms is at the end of the report. 24 figs., 56 tabs

  18. Uranium industry annual 1994

    International Nuclear Information System (INIS)

    1995-01-01

    The Uranium Industry Annual 1994 (UIA 1994) provides current statistical data on the US uranium industry's activities relating to uranium raw materials and uranium marketing during that survey year. The UIA 1994 is prepared for use by the Congress, Federal and State agencies, the uranium and nuclear electric utility industries, and the public. It contains data for the 10-year period 1985 through 1994 as collected on the Form EIA-858, ''Uranium Industry Annual Survey.'' Data collected on the ''Uranium Industry Annual Survey'' (UIAS) provide a comprehensive statistical characterization of the industry's activities for the survey year and also include some information about industry's plans and commitments for the near-term future. Where aggregate data are presented in the UIA 1994, care has been taken to protect the confidentiality of company-specific information while still conveying accurate and complete statistical data. A feature article, ''Comparison of Uranium Mill Tailings Reclamation in the United States and Canada,'' is included in the UIA 1994. Data on uranium raw materials activities including exploration activities and expenditures, EIA-estimated resources and reserves, mine production of uranium, production of uranium concentrate, and industry employment are presented in Chapter 1. Data on uranium marketing activities, including purchases of uranium and enrichment services, and uranium inventories, enrichment feed deliveries (actual and projected), and unfilled market requirements are shown in Chapter 2

  19. Uranium: a basic evaluation

    International Nuclear Information System (INIS)

    Crull, A.W.

    1978-01-01

    All energy sources and technologies, including uranium and the nuclear industry, are needed to provide power. Public misunderstanding of the nature of uranium and how it works as a fuel may jeopardize nuclear energy as a major option. Basic chemical facts about uranium ore and uranium fuel technology are presented. Some of the major policy decisions that must be made include the enrichment, stockpiling, and pricing of uranium. Investigations and lawsuits pertaining to uranium markets are reviewed, and the point is made that oil companies will probably have to divest their non-oil energy activities. Recommendations for nuclear policies that have been made by the General Accounting Office are discussed briefly

  20. Uranium health physics

    International Nuclear Information System (INIS)

    1980-01-01

    This report contains the papers delivered at the Summer School on Uranium Health Physics held in Pretoria on the 14 and 15 April 1980. The following topics were discussed: uranium producton in South Africa; radiation physics; internal dosimetry and radiotoxicity of long-lived uranium isotopes; uranium monitoring; operational experience on uranium monitoring; dosimetry and radiotoxicity of inhaled radon daughters; occupational limits for inhalation of radon-222, radon-220 and their short-lived daughters; radon monitoring techniques; radon daughter dosimeters; operational experience on radon monitoring; and uranium mill tailings management

  1. Uranium: one utility's outlook

    International Nuclear Information System (INIS)

    Gass, C.B.

    1983-01-01

    The perspective of the Arizona Public Service Company (APS) on the uncertainty of uranium as a fuel supply is discussed. After summarizing the history of nuclear power and the uranium industries, a projection is made for the future uranium market. An uncrtain uranium market is attributed to various determining factors that include international politics, production costs, non-commercial government regulation, production-company stability, and questionable levels of uranium sales. APS offers its solutions regarding type of contract, choice of uranium producers, pricing mechanisms, and aids to the industry as a whole. 5 references, 10 figures, 1 table

  2. Bragg grating induced cladding mode coupling due to asymmetrical index modulation in depressed cladding fibers

    DEFF Research Database (Denmark)

    Berendt, Martin Ole; Grüne-Nielsen, Lars; Soccolich, C.F.

    1998-01-01

    to reduce this problem. None of these designs seems to give complete solutions. In particular, the otherwise promising depressed cladding design gives a pronounced coupling to one LP01 mode, this has been referred to as a Ghost grating. To find the modes of the fiber we have established a numerical mode......UV-written Bragg gratings find wide spread use as wavelength selective components. In reflection high extinction ratios are routinely obtained. However, coupling to cladding modes gives excess loss on the short wavelength side of the main reflection. Different fiber-designs have been proposed......-solver based on the staircase-approximation method. The Bragg grating causes coupling between the fundamental LP01 mode and higher order LP1p modes that satisfy phase-matching. The coupling strength is determined by the overlap integral of the LP01, the LP1p mode, and the UV-induced index perturbation. For LP0...

  3. Recovery of uranium from crude uranium tetrafluoride

    International Nuclear Information System (INIS)

    Ghosh, S.K.; Bellary, M.P.; Keni, V.S.

    1994-01-01

    An innovative process has been developed for recovery of uranium from crude uranium tetrafluoride cake. The process is based on direct dissolution of uranium tetrafluoride in nitric acid in presence of aluminium hydroxide and use of solvent extraction for removal of fluorides and other bulk impurities to make uranium amenable for refining. It is a simple process requiring minimum process step and has advantage of lesser plant corrosion. This process can be applied for processing of uranium tetrafluoride generated from various sources like uranium by-product during thorium recovery from thorium concentrate, first stage product of uranium recovery from phosphoric acid by OPPA process and off grade uranium tetrafluoride material. The paper describes the details of the process developed and demonstrated on bench and pilot scale and its subsequent modification arising out of bulky solid waste generation. The modified process uses a lower quantity of aluminium hydroxide by allowing a lower dissolution of uranium per cycle and recycles the undissolved material to the next cycle, maintaining the overall recovery at high level. This innovation has reduced the solid waste generated by a factor of four at the cost of a slightly larger dissolution vessel and its increased corrosion rate. (author)

  4. Recovery of uranium from crude uranium tetrafluoride

    Energy Technology Data Exchange (ETDEWEB)

    Ghosh, S K; Bellary, M P; Keni, V S [Chemical Engineering Division, Bhabha Atomic Research Centre, Mumbai (India)

    1994-06-01

    An innovative process has been developed for recovery of uranium from crude uranium tetrafluoride cake. The process is based on direct dissolution of uranium tetrafluoride in nitric acid in presence of aluminium hydroxide and use of solvent extraction for removal of fluorides and other bulk impurities to make uranium amenable for refining. It is a simple process requiring minimum process step and has advantage of lesser plant corrosion. This process can be applied for processing of uranium tetrafluoride generated from various sources like uranium by-product during thorium recovery from thorium concentrate, first stage product of uranium recovery from phosphoric acid by OPPA process and off grade uranium tetrafluoride material. The paper describes the details of the process developed and demonstrated on bench and pilot scale and its subsequent modification arising out of bulky solid waste generation. The modified process uses a lower quantity of aluminium hydroxide by allowing a lower dissolution of uranium per cycle and recycles the undissolved material to the next cycle, maintaining the overall recovery at high level. This innovation has reduced the solid waste generated by a factor of four at the cost of a slightly larger dissolution vessel and its increased corrosion rate. (author). 4 refs., 1 fig., 3 tabs.

  5. Experimental approach for adhesion strength of ATF cladding

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Donghyun; Kim, Hyochan; Yang, Yongsik; In, Wangkee [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Haksung [Hanyang University, Seoul (Korea, Republic of)

    2016-10-15

    The quality of a coating depends on the quality of its adhesion bond strength between the coating and the underlying substrate. Therefore, it is essential to evaluate the adhesion properties of the coating. There are many available test methods for the evaluation of coatings adhesion bond strength. Considering these restrictions of the coated cladding, the scratch test is useful for evaluation of adhesion properties compared to other methods. The purpose of the present study is to analyze the possibility of adhesion bond strength evaluation of ATF coated cladding by scratch testing on coatings cross sections. Experimental approach for adhesion strength of ATF coated cladding was investigated in the present study. The scratch testing was chosen as a testing method. Uncoated zircaloy-4 tube was employed as a reference and plasma spray and arc ion coating were selected as a ATF coated claddings for comparison. As a result, adhesion strengths of specimens affect the measured normal and tangential forces. For the future, the test will be conducted for CrAl coated cladding by laser coating, which is the most promising ATF cladding. Computational analysis with finite element method will also be conducted to analyze a stress distribution in the cladding tube.

  6. Investigations on dry sliding of laser cladded aluminum bronze

    Directory of Open Access Journals (Sweden)

    Freiße Hannes

    2016-01-01

    Full Text Available The aim of this study was to investigate the tribological behaviour of laser cladded aluminum bronze tool surfaces for dry metal forming. In a first part of this work a process window for cladding aluminum bronze on steel substrate was investigated to ensure a low dilution. Therefore, the cladding speed, the powder feed rate, the laser power and the distance between the process head and the substrate were varied. The target of the second part was to investigate the influence of different process parameters on the tribological behaviour of the cladded tracks. The laser claddings were carried out on both aluminum bronze and cold work tool steel as substrate materials. Two different particle sizes of the cladding powder material were used. The cladding speed was varied and a post-processing laser remelting treatment was applied. It is shown that the tribological behaviour of the surface in a dry oscillating ball-on-plate test is highly dependent on the substrate material. In the third part a deep drawing tool was additively manufactured by direct laser deposition. Furthermore, the tool was applied to form circular cups with and without lubrication.

  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. Bending of pipes with inconel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Nachpitz, Leonardo; Menezes, Carlos Eduardo B; Vieira, Carlos R. Tavares [Primus Processamento de Tubos S.A. (PROTUBO), Macae, RJ (Brazil)

    2009-07-01

    The high-frequency induction bending process, using API pipes coated with Inconel 625 reconciled to a mechanical transformation for a higher degree of resistance, was developed through a careful specification and control of the manufacturing parameters and inherent heat treatments. The effects of this technology were investigated by a qualification process consisting of a sequence of tests and acceptance criteria typically required by the offshore industry, and through the obtained results was proved the effectiveness of this entire manufacturing process, without causing interference in the properties and the quality of the inconel cladding, adding a gain of resistance to the base material, guaranteed by the requirements of the API 5L Standard. (author)

  9. Laser cladding of bioactive glass coatings.

    Science.gov (United States)

    Comesaña, R; Quintero, F; Lusquiños, F; Pascual, M J; Boutinguiza, M; Durán, A; Pou, J

    2010-03-01

    Laser cladding by powder injection has been used to produce bioactive glass coatings on titanium alloy (Ti6Al4V) substrates. Bioactive glass compositions alternative to 45S5 Bioglass were demonstrated to exhibit a gradual wetting angle-temperature evolution and therefore a more homogeneous deposition of the coating over the substrate was achieved. Among the different compositions studied, the S520 bioactive glass showed smoother wetting angle-temperature behavior and was successfully used as precursor material to produce bioactive coatings. Coatings processed using a Nd:YAG laser presented calcium silicate crystallization at the surface, with a uniform composition along the coating cross-section, and no significant dilution of the titanium alloy was observed. These coatings maintain similar bioactivity to that of the precursor material as demonstrated by immersion in simulated body fluid. Copyright 2009 Acta Materialia Inc. Published by Elsevier Ltd. All rights reserved.

  10. Solution treatment of fast reactor claddings

    International Nuclear Information System (INIS)

    Miura, Makoto; Nagaki, Hiroshi; Koyama, Masahiro

    1974-01-01

    The fuel cladding tubes for Joyo (experimental FBR) are required to be a material corresponding to AISI Type 316 and cold-rolled after solution treatment. It is necessary to have no precipitation of carbide and to make the grain size smaller than ASTM No.6. It is very difficult to obtain the fine grains without the precipitation, however. In this connection, the behavior of carbide solution at high temperature and the annealing behavior of the material cold-worked and solution-treated were studied. The following matters are described: the solid solubility line of AISI Type 316; the behavior of carbide at solution treatment temperature; and the annealing behavior of the cold-worked material. (Mori, K.)

  11. Plasma spheroidizing and cladding of powders

    International Nuclear Information System (INIS)

    Petrunichev, V.A.; Averin, V.V.; Sorokin, L.M.; Koroleva, E.B.

    1987-01-01

    Arc and high-frequency plasmatrons are used for spheroidizing nickel and chromium-base alloy particles. Different plasma-forming medium compositions are used in the arc variant and the effect of these media on the plasma treatment product is demonstrated. For a high-frequency plasmatron, a long time of plasma contact with the powder leads to the transfer of the part of the material from the treated particles into vaporous state with subsequent condensation at the outlet from the discharge zone. Results of investigations into the formation of metal coatings on oxide and carbide particles during plasma-arc action are also presented. Representative data on the output of particles with coating are obtained and factors, providing for the optimal particle cladding conditions, are indicated

  12. Uranium production

    International Nuclear Information System (INIS)

    Jones, J.Q.

    1981-01-01

    The domestic uranium industry is in a state of stagflation. Costs continue to rise while the market for the product remains stagnant. During the last 12 months, curtailments and closures of mines and mills have eliminated over 5000 jobs in the industry, plus many more in those industries that furnish supplies and services. By January 1982, operations at four mills and the mines that furnish them ore will have been terminated. Other closures may follow, depending on cost trends, duration of current contracts, the degree to which mills have been amortized, the feasibility of placing mines on standby, the grade of the ore, and many other factors. Open-pit mines can be placed on standby without much difficulty, other than the possible cost of restoration before all the ore has been removed. There are a few small, dry, underground mines that could be mothballed; however, the major underground producers are wet sandstone mines that in most cases could not be reopened after a prolonged shutdown; mills can be mothballed for several years. Figure 8 shows the location of all the production centers in operation, as well as those that have operated or are on standby. Table 1 lists the same production centers plus those that have been deferred, showing nominal capacity of conventional mills in tons of ore per calendar day, and the industry production rate for those mills as of October 1, 1981

  13. Uranium mining in Australia

    International Nuclear Information System (INIS)

    Anon.

    1980-01-01

    Known uranium deposits and the companies involved in uranium mining and exploration in Australia are listed. The status of the development of the deposits is outlined and reasons for delays to mining are given

  14. Uranium Processing Facility

    Data.gov (United States)

    Federal Laboratory Consortium — An integral part of Y‑12's transformation efforts and a key component of the National Nuclear Security Administration's Uranium Center of Excellence, the Uranium...

  15. Uranium in Niger

    International Nuclear Information System (INIS)

    Gabelmann, E.

    1978-03-01

    This document presents government policy in the enhancement of uranium resources, existing mining companies and their productions, exploitation projects and economical outcome related to the uranium mining and auxiliary activities [fr

  16. Price of military uranium

    International Nuclear Information System (INIS)

    Klimenko, A.V.

    1998-01-01

    The theoretical results about optimum strategy of use of military uranium confirmed by systems approach accounts are received. The numerical value of the system approach price of the highly enriched military uranium also is given

  17. Uranium market and resources

    International Nuclear Information System (INIS)

    Capus, G.; Arnold, Th.

    2004-01-01

    The controversy about the extend of the uranium resources worldwide is still important, this article sheds some light on this topic. Every 2 years IAEA and NEA (nuclear energy agency) edit an inventory of uranium resources as reported by contributing countries. It appears that about 4.6 millions tons of uranium are available at a recovery cost less than 130 dollars per kg of uranium and a total of 14 millions tons of uranium can be assessed when including all existing or supposed resources. In fact there is enough uranium to sustain a moderate growth of the park of nuclear reactors during next decades and it is highly likely that the volume of uranium resources can allow a more aggressive development of nuclear energy. It is recalled that a broad use of the validated breeder technology can stretch the durability of uranium resources by a factor 50. (A.C.)

  18. Uranium from phosphate ores

    International Nuclear Information System (INIS)

    Hurst, F.J.

    1983-01-01

    The following topics are described briefly: the way phosphate fertilizers are made; how uranium is recovered in the phosphate industry; and how to detect covert uranium recovery operations in a phsophate plant

  19. Industrial realities: Uranium

    International Nuclear Information System (INIS)

    Thiron, H.

    1990-01-01

    In this special issue are examined ores and metals in France and in the world for 1988. The chapter on uranium gives statistical data on the uranium market: Demand, production, prices and reserves [fr

  20. Brazilian uranium deposits

    International Nuclear Information System (INIS)

    Santos, L.C.S. dos.

    1985-01-01

    Estimatives of uranium reserves carried out in Figueira, Itataia, Lagoa Real and Espinharas, in Brazil are presented. The samples testing allowed to know geological structures, and the characteristics of uranium mineralization. (M.C.F.) [pt

  1. Uranium mining in Australia

    International Nuclear Information System (INIS)

    Anon.

    1984-01-01

    The mining of uranium in Australia is criticised in relation to it's environmental impact, economics and effects on mine workers and Aborigines. A brief report is given on each of the operating and proposed uranium mines in Australia

  2. Experiments for evaluation of corrosion to develop storage criteria for interim dry storage of aluminum-alloy clad spent nuclear fuel

    International Nuclear Information System (INIS)

    Peacock, H.B.; Sindelar, R.L.; Lam, P.S.; Murphy, T.H.

    1994-01-01

    The technical bases for specification of limits to environmental exposure conditions to avoid excessive degradation are being developed for storage criteria for dry storage of highly-enriched, aluminum-clad spent nuclear fuels owned by the US Department of Energy. Corrosion of the aluminum cladding is a limiting degradation mechanism (occurs at lowest temperature) for aluminum exposed to an environment containing water vapor. Attendant radiation fields of the fuels can lead to production of nitric acid in the presence of air and water vapor and would exacerbate the corrosion of aluminum by lowering the pH of the water solution. Laboratory-scale specimens are being exposed to various conditions inside an autoclave facility to measure the corrosion of the fuel matrix and cladding materials through weight change measurements and metallurgical analysis. In addition, electrochemical corrosion tests are being performed to supplement the autoclave testing by measuring differences in the general corrosion and pitting corrosion behavior of the aluminum cladding alloys and the aluminum-uranium fuel materials in water solutions

  3. Estimation of penetration depth of fission products in cladding Hull

    International Nuclear Information System (INIS)

    Kim, Hee Moon; Jung, Yang Hong; Yoo, Byong Ok; Choo, Yong Sun; Hong, Kwon Pyo

    2005-01-01

    A disposal and a reprocessing for spent fuel rod with high burnup need de-cladding procedure. Pellet in this rod has been separated from a cladding hull to reduce a radioactivity of hull by chemical and mechanical methods. But fission products and actinides(U,Pu) still remain inside of cladding hull by chemical bonding and fission spike, which is called as 'contamination'. More specific removal of this contamination would have been considered. In this study, the sorts of fission products and penetration depth in hull were observed by EPMA test. To analyze this behavior, SRIM 2000 code was also used as energies of fission products and an oxide thickness of hull

  4. Fuel compliance model for pellet-cladding mechanical interaction

    International Nuclear Information System (INIS)

    Shah, V.N.; Carlson, E.R.

    1985-01-01

    This paper describes two aspects of fuel pellet deformation that play significant roles in determining maximum cladding hoop strains during pellet-cladding mechanical interaction: compliance of fragmented fuel pellets and influence of the pellet end-face design on the transmission of axial compressive force in the fuel stack. The latter aspect affects cladding ridge formation and explains several related observations that cannot be explained by the hourglassing model. An empirical model, called the fuel compliance model and representing the above aspects of fuel deformation, has been developed using the results from two Halden experiments and incorporated into the FRAP-T6 fuel performance code

  5. Deep-probe metal-clad waveguide biosensors

    DEFF Research Database (Denmark)

    Skivesen, Nina; Horvath, Robert; Thinggaard, S.

    2007-01-01

    Two types of metal-clad waveguide biosensors, so-called dip-type and peak-type, are analyzed and tested. Their performances are benchmarked against the well-known surface-plasmon resonance biosensor, showing improved probe characteristics for adlayer thicknesses above 150-200 nm. The dip-type metal-clad...... waveguide sensor is shown to be the best all-round alternative to the surface-plasmon resonance biosensor. Both metal-clad waveguides are tested experimentally for cell detection, showing a detection linut of 8-9 cells/mm(2). (c) 2006 Elsevier B.V. All rights reserved....

  6. Fabrication of oxide dispersion strengthened ferritic clad fuel pins

    International Nuclear Information System (INIS)

    Zirker, L.R.; Bottcher, J.H.; Shikakura, S.; Tsai, C.L.

    1991-01-01

    A resistance butt welding procedure was developed and qualified for joining ferritic fuel pin cladding to end caps. The cladding are INCO MA957 and PNC ODS lots 63DSA and 1DK1, ferritic stainless steels strengthened by oxide dispersion, while the end caps are HT9 a martensitic stainless steel. With adequate parameter control the weld is formed without a residual melt phase and its strength approaches that of the cladding. This welding process required a new design for fuel pin end cap and weld joint. Summaries of the development, characterization, and fabrication processes are given for these fuel pins. 13 refs., 6 figs., 1 tab

  7. WWER water chemistry related to fuel cladding behaviour

    Energy Technology Data Exchange (ETDEWEB)

    Kysela, J; Zmitko, M [Nuclear Research Inst. plc., Rez (Czech Republic); Vrtilkova, V [Nuclear Fuel Inst., Prague (Czech Republic)

    1997-02-01

    Operational experience in WWER primary water chemistry and corrosion related to the fuel cladding is reviewed. Insignificant corrosion of fuel cladding was found which is caused by good corrosion resistance of Zr1Nb material and relatively low coolant temperature at WWER-440 reactor units. The differences in water chemistry control is outlined and an attention to the question of compatibility of Zircaloys with WWER water chemistry is given. Some results of research and development in field of zirconium alloy corrosion behaviour are discussed. Experimental facility for in-pile and out-of-pile cladding material corrosion testing is shown. (author). 14 refs, 5 figs, 3 tabs.

  8. Uranium mining in Australia

    International Nuclear Information System (INIS)

    Mackay, G.A.

    1978-01-01

    Western world requirements for uranium based on increasing energy consumption and a changing energy mix, will warrant the development of Australia's resources. By 1985 Australian mines could be producing 9500 tonnes of uranium oxide yearly and by 1995 the export value from uranium could reach that from wool. In terms of benefit to the community the economic rewards are considerable but, in terms of providing energy to the world, Australias uranium is vital

  9. Radiation damage of uranium

    International Nuclear Information System (INIS)

    Lazarevic, Dj.

    1966-11-01

    Study of radiation damage covered the following: Kinetics of electric resistance of uranium and uranium alloy with 1% of molybdenum dependent on the second phase and burnup rate; Study of gas precipitation and diffusion of bubbles by transmission electron microscopy; Numerical analysis of the influence of defects distribution and concentration on the rare gas precipitation in uranium; study of thermal sedimentation of uranium alloy with molybdenum; diffusion of rare gas in metal by gas chromatography method

  10. Bicarbonate leaching of uranium

    International Nuclear Information System (INIS)

    Mason, C.

    1998-01-01

    The alkaline leach process for extracting uranium from uranium ores is reviewed. This process is dependent on the chemistry of uranium and so is independent on the type of mining system (conventional, heap or in-situ) used. Particular reference is made to the geochemical conditions at Crownpoint. Some supporting data from studies using alkaline leach for remediation of uranium-contaminated sites is presented

  11. Bicarbonate leaching of uranium

    Energy Technology Data Exchange (ETDEWEB)

    Mason, C.

    1998-12-31

    The alkaline leach process for extracting uranium from uranium ores is reviewed. This process is dependent on the chemistry of uranium and so is independent on the type of mining system (conventional, heap or in-situ) used. Particular reference is made to the geochemical conditions at Crownpoint. Some supporting data from studies using alkaline leach for remediation of uranium-contaminated sites is presented.

  12. Uranium in fossil bones

    International Nuclear Information System (INIS)

    Koul, S.L.

    1978-01-01

    An attempt has been made to determine the uranium content and thus the age of certain fossil bones Haritalyangarh (Himachal Pradesh), India. The results indicate that bones rich in apatite are also rich in uranium, and that the radioactivity is due to radionuclides in the uranium series. The larger animals apparently have a higher concentration of uranium than the small. The dating of a fossil jaw (elephant) places it in the Pleistocene. (Auth.)

  13. Recent irradiation tests of uranium-plutonium-zirconium metal fuel elements

    International Nuclear Information System (INIS)

    Pahl, R.G.; Lahm, C.E.; Villarreal, R.; Hofman, G.L.; Beck, W.N.

    1986-09-01

    Uranium-Plutonium-Zirconium metal fuel irradiation tests to support the ANL Integral Fast Reactor concept are discussed. Satisfactory performance has been demonstrated to 2.9 at.% peak burnup in three alloys having 0, 8, and 19 wt % plutonium. Fuel swelling measurements at low burnup in alloys to 26 wt % plutonium show that fuel deformation is primarily radial in direction. Increasing the plutonium content in the fuel diminishes the rate of fuel-cladding gap closure and axial fuel column growth. Chemical redistribution occurs by 2.1 at.% peak burnup and generally involves the inward migration of zirconium and outward migration of uranium. Fission gas release to the plenum ranges from 46% to 56% in the alloys irradiated to 2.9 at.% peak burnup. No evidence of deleterious fuel-cladding chemical or mechanical interaction was observed

  14. Method for converting uranium oxides to uranium metal

    International Nuclear Information System (INIS)

    Duerksen, W.K.

    1988-01-01

    A method for converting uranium oxide to uranium metal is described comprising the steps of heating uranium oxide in the presence of a reducing agent to a temperature sufficient to reduce the uranium oxide to uranium metal and form a heterogeneous mixture of a uranium metal product and oxide by-products, heating the mixture in a hydrogen atmosphere at a temperature sufficient to convert uranium metal in the mixture to uranium hydride, cooling the resulting uranium hydride-containing mixture to a temperature sufficient to produce a ferromagnetic transition in the uranium hydride, magnetically separating the cooled uranium hydride from the mixture, and thereafter heating the separated uranium hydride in an inert atmosphere to a temperature sufficient to convert the uranium hydride to uranium metal

  15. Design of Matched Cladding Fiber with UV-sensitive Cladding for Minimization of Claddingmode Losses in Fiber Bragg Gratings

    DEFF Research Database (Denmark)

    Nielsen, Mads Lønstrup; Berendt, Martin Ole; Bjarklev, Anders Overgaard

    2000-01-01

    The effect on the Bragg-grating-induced cladding-mode coupling of varying the extent of the photosensitive region in a step-index fiber is analyzed. We introduce a figure of merit for the suppression of cladding-mode loss and compare different matched cladding fiber designs. It is found to be adv......The effect on the Bragg-grating-induced cladding-mode coupling of varying the extent of the photosensitive region in a step-index fiber is analyzed. We introduce a figure of merit for the suppression of cladding-mode loss and compare different matched cladding fiber designs. It is found...... to be advantageous to increase the extent of the photosensitive region. However, no significant improvement is obtained by extending the photosensitive region more than approximately 10 mu m into the cladding. This result is not in agreement with a simple analysis that neglects UV absorption, which suggests...... that the radius of the photosensitive region should be close to twice as large. (C) 2000 Academic Press....

  16. The Fabrication Problem Of U3Si2-Al Fuel With Uranium High Loading

    International Nuclear Information System (INIS)

    Supardjo

    1996-01-01

    The quality of U 3 Si 2 -Al dispersion fuel product is the main aim for each fabricator. Low loading of uranium fuel element is easily fabricated, but with the increased, uranium loading, homogeneity of uranium distribution is difficult to achieve and it always formed white spots, blister, and dogboning in the fuel plates. The problem can be eliminated by the increasing treatment of the fuel/Al powder. The precise selection of fuel/Al particles diameter is needed indeed to make easier in the homogeneous process of powder and the porosities arrangement in the fuel plates. The increasing of uranium loading at constant meat thickness will increase the meat hardness, therefore to withdraw the dogboning forming, the use of harder cladding materials is necessity

  17. Microbial accumulation of uranium

    International Nuclear Information System (INIS)

    Zhang Wei; Dong Faqin; Dai Qunwei

    2005-01-01

    The mechanism of microbial accumulation of uranium and the effects of some factors (including pH, initial uranium concentration, pretreatment of bacteria, and so on) on microbial accumulation of uranium are discussed briefly. The research direction and application prospect are presented. (authors)

  18. Uranium energy dependence

    International Nuclear Information System (INIS)

    Erkes, P.

    1981-06-01

    Uranium supply and demand as projected by the Uranium Institute is discussed. It is concluded that for the industrialized countries, maximum energy independence is a necessity. Hence it is necessary to achieve assurance of supply for uranium used in thermal power reactors in current programs and eventually to move towards breeders

  19. Australian uranium today

    International Nuclear Information System (INIS)

    Fisk, B.

    1978-01-01

    The subject is covered in sections, entitled: Australia's resources; Northern Territory uranium in perspective; the government's decision [on August 25, 1977, that there should be further development of uranium under strictly controlled conditions]; Government legislation; outlook [for the Australian uranium mining industry]. (U.K.)

  20. Uranium resources, 1983

    International Nuclear Information System (INIS)

    1983-01-01

    The specific character of uranium as energy resources, the history of development of uranium resources, the production and reserve of uranium in the world, the prospect regarding the demand and supply of uranium, Japanese activity of exploring uranium resources in foreign countries and the state of development of uranium resources in various countries are reported. The formation of uranium deposits, the classification of uranium deposits and the reserve quantity of each type are described. As the geological environment of uranium deposits, there are six types, that is, quartz medium gravel conglomerate deposit, the deposit related to the unconformity in Proterozoic era, the dissemination type magma deposit, pegmatite deposit and contact deposit in igneaus rocks and metamorphic rocks, vein deposit, sandstone type deposit and the other types of deposit. The main features of respective types are explained. The most important uranium resources in Japan are those in the Tertiary formations, and most of the found reserve belongs to this type. The geological features, the state of yield and the scale of the deposits in Ningyotoge, Tono and Kanmon Mesozoic formation are reported. Uranium minerals, the promising districts in the world, and the matters related to the exploration and mining of uranium are described. (Kako, I.)

  1. Recycling of reprocessed uranium

    International Nuclear Information System (INIS)

    Randl, R.P.

    1987-01-01

    Since nuclear power was first exploited in the Federal Republic of Germany, the philosophy underlying the strategy of the nuclear fuel cycle has been to make optimum use of the resource potential of recovered uranium and plutonium within a closed fuel cycle. Apart from the weighty argument of reprocessing being an important step in the treatment and disposal of radioactive wastes, permitting their optimum ecological conditioning after the reprocessing step and subsequent storage underground, another argument that, no doubt, carried weight was the possibility of reducing the demand of power plants for natural uranium. In recent years, strategies of recycling have emerged for reprocessed uranium. If that energy potential, too, is to be exploited by thermal recycling, it is appropriate to choose a slightly different method of recycling from the one for plutonium. While the first generation of reprocessed uranium fuel recycled in the reactor cuts down natural uranium requirement by some 15%, the recycling of a second generation of reprocessed, once more enriched uranium fuel helps only to save a further three per cent of natural uranium. Uranium of the second generation already carries uranium-232 isotope, causing production disturbances, and uranium-236 isotope, causing disturbances of the neutron balance in the reactor, in such amounts as to make further fabrication of uranium fuel elements inexpedient, even after mixing with natural uranium feed. (orig./UA) [de

  2. 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

  3. Characteristics of Ni-based coating layer formed by laser and plasma cladding processes

    International Nuclear Information System (INIS)

    Xu Guojian; Kutsuna, Muneharu; Liu Zhongjie; Zhang Hong

    2006-01-01

    The clad layers of Ni-based alloy were deposited on the SUS316L stainless plates by CO 2 laser and plasma cladding processes. The smooth clad bead was obtained by CO 2 laser cladding process. The phases of clad layer were investigated by an optical microscope, scanning electron microscopy (SEM), X-ray diffractometer (XRD), electron probe microanalysis (EPMA) and energy-dispersive spectrometer (EDS). The microstructures of clad layers belonged to a hypereutectic structure. Primary phases consist of boride CrB and carbide Cr 7 C 3 . The eutectic structure consists of Ni + CrB or Ni + Cr 7 C 3 . Compared with the plasma cladding, the fine microstructures, low dilutions, high Vickers hardness and excellent wear resistance were obtained by CO 2 laser cladding. All that show the laser cladding process has a higher efficiency and good cladding quality

  4. Measurement of thermal conductivity of uranium metal using transient plane source technique

    International Nuclear Information System (INIS)

    Subramanian, G.G.S.; Bapuji, T.; Panneerselvam, G.; Antony, M.P.; Nagarajan, K.

    2012-01-01

    Thermo physical properties of fuel, cladding and structural materials play a significant role in the reactor operation. Thermal conductivity is one of the most important physical properties of the fuel which determines the maximum linear heat rating of the fuel in a reactor. As part of this study, the thermal conductivity of uranium metal was measured using a transient plane source (TPS) by Hot-disc method

  5. Study on microstructure and high temperature wear resistance of laser cladded nuclear valve clack

    International Nuclear Information System (INIS)

    Zhang Chunliang; Chen Zichen

    2002-01-01

    Laser cladding of Co-base alloy on the nuclear valve-sealing surface are performed with a 5 kW CO 2 transverse flowing laser. The microstructure and the high temperature impact-slide wear resistance of the laser cladded coating and the plasma cladded coating are studied. The results show that the microstructure, the dilution rate and the high temperature impact-slide wear resistance of the laser cladded coating have obvious advantages over the spurt cladding processing

  6. Nuclear fuel clad clothed with burnable poison and obtainment process

    International Nuclear Information System (INIS)

    Diez, P.; Netter, P.

    1994-01-01

    This clad has preferentially on its inner surface a boron compound such boron carbide or boron nitrogen deposited by Chemical Vapor Deposition or by Physical Vapor Deposition without any temperature elevation injurious to its mechanical properties. 3 figs

  7. Oxidation during reflood of reactor core with melting cladding

    Energy Technology Data Exchange (ETDEWEB)

    Siefken, L.J.; Allison, C.M.; Davis, K.L. [and others

    1995-09-01

    Models were recently developed and incorporated into the SCDAP/RELAP5 code for calculating the oxidation of fuel rods during cladding meltdown and reflood. Experiments have shown that a period of intense oxidation may occur when a hot partially oxidized reactor core is reflooded. This paper offers an explanation of the cladding meltdown and oxidation processes that cause this intense period of oxidation. Models for the cladding meltdown and oxidation processes are developed. The models are assessed by simulating a severe fuel damage experiment that involved reflood. The models for cladding meltdown and oxidation were found to improve calculation of the temperature and oxidation of fuel rods during the period in which hot fuel rods are reflooded.

  8. A Multi-Scale Modeling of Laser Cladding Process (Preprint)

    National Research Council Canada - National Science Library

    Cao, J; Choi, J

    2006-01-01

    Laser cladding is an additive manufacturing process that a laser generates a melt-pool on the substrate material while a second material, as a powder or a wire form, is injected into that melt-pool...

  9. Neutron-induced helium implantation in GCFR cladding

    International Nuclear Information System (INIS)

    Yamada, H.; Poeppel, R.B.; Sevy, R.H.

    1980-10-01

    The neutron-induced implantation of helium atoms on the exterior surfaces of the cladding of a prototypic gas-cooled fast reactor (GCFR) has been investigated analytically. A flux of recoil helium particles as high as 4.2 x 10 10 He/cm 2 .s at the cladding surface has been calculated at the peak power location in the core of a 300-MWe GCFR. The calculated profile of the helium implantation rates indicates that although some helium is implanted as deep as 20 μm, more than 99% of helium particles are implanted in the first 2-μm-deep layer below the cladding surface. Therefore, the implanted helium particles should mainly affect surface properties of the GCFR cladding

  10. Composite polymer: Glass edge cladding for laser disks

    Science.gov (United States)

    Powell, H.T.; Wolfe, C.A.; Campbell, J.H.; Murray, J.E.; Riley, M.O.; Lyon, R.E.; Jessop, E.S.

    1987-11-02

    Large neodymium glass laser disks for disk amplifiers such as those used in the Nova laser require an edge cladding which absorbs at 1 micrometer. This cladding prevents edge reflections from causing parasitic oscillations which would otherwise deplete the gain. Nova now utilizes volume-absorbing monolithic-glass claddings which are fused at high temperature to the disks. These perform quite well but are expensive to produce. Absorbing glass strips are adhesively bonded to the edges of polygonal disks using a bonding agent whose index of refraction matches that of both the laser and absorbing glass. Optical finishing occurs after the strips are attached. Laser disks constructed with such claddings have shown identical gain performance to the previous Nova disks and have been tested for hundreds of shots without significant degradation. 18 figs.

  11. Composite polymer-glass edge cladding for laser disks

    Science.gov (United States)

    Powell, Howard T.; Riley, Michael O.; Wolfe, Charles R.; Lyon, Richard E.; Campbell, John H.; Jessop, Edward S.; Murray, James E.

    1989-01-01

    Large neodymium glass laser disks for disk amplifiers such as those used in the Nova laser require an edge cladding which absorbs at 1 micrometer. This cladding prevents edge reflections from causing parasitic oscillations which would otherwise deplete the gain. Nova now utilizes volume-absorbing monolithic-glass claddings which are fused at high temperature to the disks. These perform quite well but are expensive to produce. Absorbing glass strips are adhesively bonded to the edges of polygonal disks using a bonding agent whose index of refraction matches that of both the laser and absorbing glass. Optical finishing occurs after the strips are attached. Laser disks constructed with such claddings have shown identical gain performance to the previous Nova disks and have been tested for hundreds of shots without significant degradation.

  12. Semipolar III-nitride laser diodes with zinc oxide cladding.

    Science.gov (United States)

    Myzaferi, Anisa; Reading, Arthur H; Farrell, Robert M; Cohen, Daniel A; Nakamura, Shuji; DenBaars, Steven P

    2017-07-24

    Incorporating transparent conducting oxide (TCO) top cladding layers into III-nitride laser diodes (LDs) improves device design by reducing the growth time and temperature of the p-type layers. We investigate using ZnO instead of ITO as the top cladding TCO of a semipolar (202¯1) III-nitride LD. Numerical modeling indicates that replacing ITO with ZnO reduces the internal loss in a TCO clad LD due to the lower optical absorption in ZnO. Lasing was achieved at 453 nm with a threshold current density of 8.6 kA/cm 2 and a threshold voltage of 10.3 V in a semipolar (202¯1) III-nitride LD with ZnO top cladding.

  13. Siemens advance PWR fuel assemblies (HTP) and cladding

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  14. Cladding axial elongation models for FRAP-T6

    International Nuclear Information System (INIS)

    Shah, V.N.; Carlson, E.R.; Berna, G.A.

    1983-01-01

    This paper presents a description of the cladding axial elongation models developed at the Idaho National Engineering Laboratory (INEL) for use by the FRAP-T6 computer code in analyzing the response of fuel rods during reactor transients in light water reactors (LWR). The FRAP-T6 code contains models (FRACAS-II subcode) that analyze the structural response of a fuel rod including pellet-cladding-mechanical-interaction (PCMI). Recently, four models were incorporated into FRACAS-II to calculate cladding axial deformation: (a) axial PCMI, (b) trapped fuel stack, (c) fuel relocation, and (d) effective fuel thermal expansion. Comparisons of cladding axial elongation measurements from two experiments with the corresponding FRAP-T6 calculations are presented

  15. PROCESS OF RECOVERING URANIUM

    Science.gov (United States)

    Carter, J.M.; Larson, C.E.

    1958-10-01

    A process is presented for recovering uranium values from calutron deposits. The process consists in treating such deposits to produce an oxidlzed acidic solution containing uranium together with the following imparities: Cu, Fe, Cr, Ni, Mn, Zn. The uranium is recovered from such an impurity-bearing solution by adjusting the pH of the solution to the range 1.5 to 3.0 and then treating the solution with hydrogen peroxide. This results in the precipitation of uranium peroxide which is substantially free of the metal impurities in the solution. The peroxide precipitate is then separated from the solution, washed, and calcined to produce uranium trioxide.

  16. HYDRIDE-RELATED DEGRADATION OF SNF CLADDING UNDER REPOSITORY CONDITIONS

    International Nuclear Information System (INIS)

    McCoy, K.

    2000-01-01

    The purpose and scope of this analysis/model report is to analyze the degradation of commercial spent nuclear fuel (CSNF) cladding under repository conditions by the hydride-related metallurgical processes, such as delayed hydride cracking (DHC), hydride reorientation and hydrogen embrittlement, thereby providing a better understanding of the degradation process and clarifying which aspects of the process are known and which need further evaluation and investigation. The intended use is as an input to a more general analysis of cladding degradation

  17. Second phase precipitation in irradiated Type 316 stainless steel cladding

    International Nuclear Information System (INIS)

    Hales, J.W.

    1978-05-01

    Differences in the phase composition of FFTF fuel cans following irradiation in the General Electric Test Reactor compared to HEDL fuel cans prompted laboratory studies to be conducted using cladding from the same lots of material used to fabricate the fuel pins and on cladding sections removed from the plenum area of the irradiated fuel pins to help establish the cause of the observed differences

  18. Explosion Clad for Upstream Oil and Gas Equipment

    Science.gov (United States)

    Banker, John G.; Massarello, Jack; Pauly, Stephane

    2011-01-01

    Today's upstream oil and gas facilities frequently involve the combination of high pressures, high temperatures, and highly corrosive environments, requiring equipment that is thick wall, corrosion resistant, and cost effective. When significant concentrations of CO2 and/or H2S and/or chlorides are present, corrosion resistant alloys (CRA) can become the material of choice for separator equipment, piping, related components, and line pipe. They can provide reliable resistance to both corrosion and hydrogen embrittlement. For these applications, the more commonly used CRA's are 316L, 317L and duplex stainless steels, alloy 825 and alloy 625, dependent upon the application and the severity of the environment. Titanium is also an exceptional choice from the technical perspective, but is less commonly used except for heat exchangers. Explosion clad offers significant savings by providing a relatively thin corrosion resistant alloy on the surface metallurgically bonded to a thick, lower cost, steel substrate for the pressure containment. Developed and industrialized in the 1960's the explosion cladding technology can be used for cladding the more commonly used nickel based and stainless steel CRA's as well as titanium. It has many years of proven experience as a reliable and highly robust clad manufacturing process. The unique cold welding characteristics of explosion cladding reduce problems of alloy sensitization and dissimilar metal incompatibility. Explosion clad materials have been used extensively in both upstream and downstream oil, gas and petrochemical facilities for well over 40 years. The explosion clad equipment has demonstrated excellent resistance to corrosion, embrittlement and disbonding. Factors critical to insure reliable clad manufacture and equipment design and fabrication are addressed.

  19. Explosion Clad for Upstream Oil and Gas Equipment

    International Nuclear Information System (INIS)

    Banker, John G.; Massarello, Jack; Pauly, Stephane

    2011-01-01

    Today's upstream oil and gas facilities frequently involve the combination of high pressures, high temperatures, and highly corrosive environments, requiring equipment that is thick wall, corrosion resistant, and cost effective. When significant concentrations of CO 2 and/or H 2 S and/or chlorides are present, corrosion resistant alloys (CRA) can become the material of choice for separator equipment, piping, related components, and line pipe. They can provide reliable resistance to both corrosion and hydrogen embrittlement. For these applications, the more commonly used CRA's are 316L, 317L and duplex stainless steels, alloy 825 and alloy 625, dependent upon the application and the severity of the environment. Titanium is also an exceptional choice from the technical perspective, but is less commonly used except for heat exchangers. Explosion clad offers significant savings by providing a relatively thin corrosion resistant alloy on the surface metallurgically bonded to a thick, lower cost, steel substrate for the pressure containment. Developed and industrialized in the 1960's the explosion cladding technology can be used for cladding the more commonly used nickel based and stainless steel CRA's as well as titanium. It has many years of proven experience as a reliable and highly robust clad manufacturing process. The unique cold welding characteristics of explosion cladding reduce problems of alloy sensitization and dissimilar metal incompatibility. Explosion clad materials have been used extensively in both upstream and downstream oil, gas and petrochemical facilities for well over 40 years. The explosion clad equipment has demonstrated excellent resistance to corrosion, embrittlement and disbonding. Factors critical to insure reliable clad manufacture and equipment design and fabrication are addressed.

  20. Cladding for transverse-pumped solid-state laser

    Science.gov (United States)

    Byer, Robert L. (Inventor); Fan, Tso Y. (Inventor)

    1989-01-01

    In a transverse pumped, solid state laser, a nonabsorptive cladding surrounds a gain medium. A single tranverse mode, namely the Transverse Electromagnetic (TEM) sub 00 mode, is provided. The TEM sub 00 model has a cross sectional diameter greater than a transverse dimension of the gain medium but less than a transverse dimension of the cladding. The required size of the gain medium is minimized while a threshold for laser output is lowered.