WorldWideScience

Sample records for t-111 clad uranium

  1. Irradiation of three T-111 clad uranium nitride fuel pins for 8070 hours at 990 C (1815 F)

    Science.gov (United States)

    Slaby, J. G.; Siegel, B. L.; Gedeon, L.; Galbo, R. J.

    1973-01-01

    The design and successful operation of three tantalum alloy (Ta-8W-2Hf) clad uranium mononitride (UN) fuel pins irradiated for 8070 hr at 990 C (1815 F) is described. Two pin diameters having measured burnups of 0.47 and 0.90 uranium atom percent were tested. No clad failures or swelling was detected; however, postirradiation clad samples tested failed with 1 percent strain. The fuel density decrease was 2 percent, and the fission gas release was less than 0.05 percent. Isotropic fuel swelling, which averaged about 0.5 percent, was less than fuel pin assembly clearances. Thus the clad was not strained. Thermocouples with a modified hot zone operated at average temperatures to 1100 C (2012 F) without failure. Factors that influence the ability to maintain uniform clad temperature as well as the results of the heat transfer calculations are discussed.

  2. 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.

  3. 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

  4. 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

  5. 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.

  6. Experiences in the fabrication of aluminium clad metallic uranium fuel

    International Nuclear Information System (INIS)

    Vijayaraghavan, R.

    1989-01-01

    With a view to achieve self-sufficiency and self-reliance in the fabrication of metallic natural uranium fuel, a full fledged fuel fabrication facility was set up in 1958. Based on the then technical information available and the development work carried out, a flow-sheet for the fabrication of metallic uranium fuel, starting from uranium ingots, was worked out and the first fuel element was successfully fabricated in June 1959. More than half the first charge for the initial criticality of CIRUS, a 40 MWt research reactor at Trombay, was fabricated and supplied. Since then, this facility has been regularly catering to the replacement fuel and component requirements for CIRUS. The fuel for Dhruva, a 100 MWt research reactor at Trombay, is in the form of a cluster consisting of 7 fuel pins as compared to the rigid single fuel element for CIRUS. The fabrication process chosen for making this fuel is more or less on the same lines as that has been followed for CIRUS fuel. However, because of the smaller diameter of uranium metal rod, higher length to diameter ratio, configurations of fins on aluminium sheaths, cluster assembly etc., extensive development work was required to be undertaken for optimising various production parameters. Several prototype fuel clusters of different designs were made and subjected to rigorous out-of-pile and in-pile testing. Based on the reliable satisfactory performance, type II-B SPT cluster design was finally frozen for production. Stringent quality control is of prime importance in ensuring good performance of the fuel in the reactor. Hence, appropriate quality control measures have been adopted at various stages of fuel fabrication to ensure conformance with the specifications. (author) 9 refs., 11 figs., 1 tab

  7. 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

  8. Tank 241-T-111 characterization report

    International Nuclear Information System (INIS)

    Simpson, B.C.

    1994-09-01

    In late October and early November of 1991, single-shell tank (SST) 241-T-111 (on the Organic Watch List) was sampled and analyses were conducted on the materials obtained to complete Hanford Federal Facility Agreement and Consent Order (Tri-Party Agreement) Milestone M-10-00 to sample and analyze two cores from each tank. Other objectives that these measurements and inventory estimates support are as follows: Obtain estimates of both the concentration and total quantity of key analytes relating to safety issues, such as organics and radionuclides; Provided input to risk assessment-based disposal decisions for the waste; and Implement physical property measurements, such as rheology, bulk density, and particle size. These measurements are necessary for the design and fabrication of retrieval, pretreatment, and vitrification systems. The purpose of the tank characterization report is to describe and characterize the waste in SST 241-T-111, based on information from various sources. This report summarizes the available information regarding the waste in tank 241-T-111, and arranges this information in a format useful to support management and technical decisions concerning this waste tank

  9. Effect of nitrogen on tensile properties and structures of T-111 (tantalum, 8 percent tungsten, 2 percent hafnium) tubing

    Science.gov (United States)

    Buzzard, R. J.; Metroka, R. R.

    1973-01-01

    The effect of controlled nitrogen additions was evaluated on the mechanical properties of T-111 (Ta-8W-2Hf) fuel pin cladding material proposed for use in a lithium-cooled nuclear reactor concept. Additions of 80 to 1125 ppm nitrogen resulted in increased strengthening of T-111 tubular section test specimens at temperatures of 25 to 1200 C. Homogeneous distributions of up to 500 ppm nitrogen did not seriously decrease tensile ductility. Both single and two-phase microstructures, with hafnium nitride as the second phase, were evaluated in this study.

  10. Theoretical analysis of swelling characteristics of cylindrical uranium dioxide fuel pins with a niobium - 1-percent-zirconium clad

    Science.gov (United States)

    Saltsman, J. F.

    1973-01-01

    The relations between clad creep strain and fuel volume swelling are shown for cylindrical UO2 fuel pins with a Nb-1Zr clad. These relations were obtained by using the computer code CYGRO-2. These clad-strain - fuel-volume-swelling relations may be used with any fuel-volume-swelling model, provided the fuel volume swelling is isotropic and independent of the clad restraints. The effects of clad temperature (over a range from 118 to 1642 K (2010 to 2960 R)), pin diameter, clad thickness and central hole size in the fuel have been investigated. In all calculations the irradiation time was 500 hours. The burnup rate was varied.

  11. Short term creep rupture predictions for tantalum alloy T-111

    International Nuclear Information System (INIS)

    Stephens, J.J.

    1991-01-01

    A knowledge of the short term creep rupture behavior of Tantalum alloy T-111 is necessary to predict device integrity in the heat source section of Radioisotope Thermoelectric Generators (RTGs) at the end of service life, in the event of a fuel fire. High pressures exist in RTGs near the end of service life, these are caused by gas generation resulting from radioactive decay of the nuclear fuel. The internal pressure exerts a significant hoop stress on the T-111 alloy structural containment member. This paper analyses the short term creep behavior (rupture times up to ∼2x10 3 hrs.) of cold worked (CW) T-111 alloy, using the existing data of Stephenson (1967). Corellations for the time to rupture, time to 1% strain and minimum creep rate have been obtained from this data using multivariable linear regression analysis. These results are compared to other short term rupture data for T-111 alloy. Finally, at the stress/temperature levels relevant to the RTG fuel fire scenario near the end of service life, the rupture time correlation for T-111 alloy predicts a rupture time of approximately 100 hrs

  12. Plutonium-238 dioxide/T-111 compatibility studies

    International Nuclear Information System (INIS)

    Jones, G.J.; Selle, J.E.; Teaney, P.E.

    1975-01-01

    The tantalum-base alloy, T-111, is an ideal radioisotope encapsulant from the aspect of mechanical properties, but unfortunately undergoes severe oxygen embrittlement during long-term, high-temperature exposure to PuO 2 . A study was undertaken in an effort to improve T-111/PuO 2 compatibility by testing the hypothesis that reduction of fuel stoichiometry to the range PuO 1 . 75 to PuO 1 . 8 would suspend the embrittlement process by producing a state of thermodynamic equilibrium within the capsule. Test temperatures ranged from 773 to 1373 0 K, with aging times of 60 days, 240 days, and 2 y. The desired reaction did not proceed rapidly enough at the aging temperatures to stop T-111 embrittlement. Capsules heated above 1573 0 K for 1 hr showed no signs of embrittlement during aging, even after 2 y at temperatures as high as 1173 0 K. Results with test specimens employing pelletized fuel indicated the solid-state diffusion of oxygen from the fuel to the T-111 was the dominating transport process. In nonpretreated capsules oxygen diffusion in T-111 was the rate-controlling process. Pretreatment does result in the desired thermodynamic equilibrium at temperatures up to at least 1173 0 K. (auth)

  13. Uranium

    International Nuclear Information System (INIS)

    Anon.

    1980-01-01

    Developments in the the uranium industry in Australia that took place during the quarter ended 30 June 1980 are reviewed. These include uranium mine production and uranium exploration. Prices for uranium oxide and uranium hexafluoride as at the end of June 1980 and figures for U 3 O 8 production and export from 1978 to March 1980 are listed

  14. 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

  15. 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

  16. 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

  17. 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.)

  18. 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

  19. 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)

  20. Uranium

    International Nuclear Information System (INIS)

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

    1986-01-01

    Production of uranium oxide in Australia for 1983 was 3786 t(3211 t U). Uranium exports for 1983 were 3273 t U 3 O 8 at an average f.o.b. value of $41.02/lb U 3 O 8 . Private exploration expenditure for uranium in Australia during the 1982-83 fiscal year was $36.5 million, 35% less than in 1981-82. In November 1983, the Government decided that uranium mining would be allowed only at the existing Ranger and Nabarlek mines and at the proposed Olympic Dam mine. Australia's Reasonably Assured Resources of uranium recoverable at less than US$80/kg U as at December 1983, totalled 474 000 t U. Australia's total now represents 30% of the Western world's low-cost RAR. In addition Australia has 235 000 t U in the low-cost Estimated Additional Resources Category 1, which represents 31% of the Western world's resources in this category

  1. Uranium

    International Nuclear Information System (INIS)

    Villarreal, E.

    1986-01-01

    After the increase in oil prices in 1973, several European countries increased their power programs. As a result some uranium mining companies from the FRG, Spain and France invested in exploration of radioactive minerals in Colombia hoping to find uranium resources needed to fuel European reactors. In the article a historic review of foreign investment in uranium in Colombia is made; some recommendations about joint-venture contracts used to regulate the work of the foreign companies are included. The four companies involved in exploration left the country in the early eighties, due to the difficulties in finding a large deposit and the difficult world situation of nuclear power

  2. 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

  3. Uranium*

    Science.gov (United States)

    Grenthe, Ingmar; Drożdżyński, Janusz; Fujino, Takeo; Buck, Edgar C.; Albrecht-Schmitt, Thomas E.; Wolf, Stephen F.

    Uranium compounds have been used as colorants since Roman times (Caley, 1948). Uranium was discovered as a chemical element in a pitchblende specimen by Martin Heinrich Klaproth, who published the results of his work in 1789. Pitchblende is an impure uranium oxide, consisting partly of the most reduced oxide uraninite (UO2) and partly of U3O8. Earlier mineralogists had considered this mineral to be a complex oxide of iron and tungsten or of iron and zinc, but Klaproth showed by dissolving it partially in strong acid that the solutions yielded precipitates that were different from those of known elements. Therefore he concluded that it contained a new element (Mellor, 1932); he named it after the planet Uranus, which had been discovered in 1781 by William Herschel, who named it after the ancient Greek deity of the Heavens.

  4. Addition of oxygen to and distribution of oxides in tantalum alloy T-111 at low concentrations

    Science.gov (United States)

    Stecura, S.

    1975-01-01

    Oxygen was added at 820 and 990 C at an oxygen pressure of about .0003 torr. The technique permitted predetermined and reproducible oxygen doping of the tantalum alloy (T-111). Based on the temperature dependency of the doping reaction, it was concluded that the initial rates of oxygen pickup are probably controlled by solution of oxygen into the T-111 lattice. Although hafnium oxides are more stable than those of tantalum or tungsten, analyses of extracted residues indicate that the tantalum and tungsten oxides predominate in the as-doped specimens, presumably because of the higher concentrations of tantalum and tungsten in the alloy. However, high-temperature annealing promotes gettering of dissolved oxygen and oxygen from other oxides to form hafnium oxides. Small amounts of tantalum and tungsten oxides were still present after high temperature annealing. Tungsten oxide (WO3) volatilizes slightly from the surface of T-111 at 990 C but not at 820 C. The vaporization of WO3 has no apparent effect on the doping reaction.

  5. Calculation of radiation induced swelling of uranium mononitride using the digital computer program CYGRO 2

    Science.gov (United States)

    Davison, H. W.; Fiero, I. B.

    1971-01-01

    Fuel volume swelling and clad diametral creep strains were calculated for five fuel pins, clad with either T-111 (Ta-8W-2.4Hf) or PWC-11 (Nb-1Zr-0.1C). The fuel pins were irradiated to burnups between 2.7 and 4.6%. Clad temperatures were between 1750 and 2400 F (1228 and 1589 K). The maximum percentage difference between calculated and experimentally measured values of volumetric fuel swelling is 60%.

  6. 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

  7. 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

  8. 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

  9. The T111I variant in the endothelial lipase gene and risk of coronary heart disease in three independent populations

    DEFF Research Database (Denmark)

    Jensen, M. K.; Rimm, E. B.; Mukamal, K. J.

    2009-01-01

    of coronary heart disease (CHD) in the general population is unclear. The objective of this study was to address the associations of the T111I variant with plasma lipids and risk of CHD in three independent prospective studies of generally healthy men and women. The T111I variant was genotyped in case......-control studies of CHD nested within the Diet, Cancer, and Health study with 998 cases, Nurses' Health Study with 241 cases, and Health Professionals Follow-up Study with 262 cases. The minor allele frequency in the combined pool of controls was 0.29. The T111I variant was not associated with HDL-C or any other...... lipid and lipoprotein measures. Compared with wildtype homozygotes, the pooled estimate for risk of CHD was 0.95 (0.85-1.06) per T111I allele. Our analysis among healthy Caucasian men and women from three independent studies does not support an association between the T111I variant and HDL-C, other...

  10. 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

  11. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Peacock, H.B. Jr.

    1999-10-21

    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.

  13. 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

  14. 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...

  15. 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)

  16. 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

  17. 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

  18. 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.

  19. Regarding the chemistry of metallic uranium stored in steel drums

    International Nuclear Information System (INIS)

    Wood, D.H.; Snowden, S.A.; Howe, H.J. Jr.; Thomas, L.L.; Moon, D.W.; Gregg, H.R.; Miller, P.E.

    1994-01-01

    This letter describes the chemistry leading to an unexpected release of chemical energy that occured during the opening of a steel drum containing co-extruded, iron-clad, depleted-uranium rods. (orig.). Short communication

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

    Energy Technology Data Exchange (ETDEWEB)

    Sindelar, R.L.; Peacock, H.B. Jr.; Lam, P.S.; Iyer, N.C.; Louthan, M.R. Jr.; Murphy, J.R. [Westinghouse Savannah River Co., Aiken, SC (United States)

    1996-08-01

    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.

  1. Study of the uranium-zirconium diffusion

    International Nuclear Information System (INIS)

    Adda, Y.; Mairy, C.; Bouchet, P.

    1957-01-01

    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.)

  2. Aerogel-clad optical fiber

    Science.gov (United States)

    Sprehn, Gregory A.; Hrubesh, Lawrence W.; Poco, John F.; Sandler, Pamela H.

    1997-01-01

    An optical fiber is surrounded by an aerogel cladding. For a low density aerogel, the index of refraction of the aerogel is close to that of air, which provides a high numerical aperture to the optical fiber. Due to the high numerical aperture, the aerogel clad optical fiber has improved light collection efficiency.

  3. Ice-clad volcanoes

    Science.gov (United States)

    Waitt, Richard B.; Edwards, B.R.; Fountain, Andrew G.; Huggel, C.; Carey, Mark; Clague, John J.; Kääb, Andreas

    2015-01-01

    An icy volcano even if called extinct or dormant may be active at depth. Magma creeps up, crystallizes, releases gas. After decades or millennia the pressure from magmatic gas exceeds the resistance of overlying rock and the volcano erupts. Repeated eruptions build a cone that pokes one or two kilometers or more above its surroundings - a point of cool climate supporting glaciers. Ice-clad volcanic peaks ring the northern Pacific and reach south to Chile, New Zealand, and Antarctica. Others punctuate Iceland and Africa (Fig 4.1). To climb is irresistible - if only “because it’s there” in George Mallory’s words. Among the intrepid ascents of icy volcanoes we count Alexander von Humboldt’s attempt on 6270-meter Chimborazo in 1802 and Edward Whymper’s success there 78 years later. By then Cotopaxi steamed to the north.

  4. 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.

  5. Metal-clad waveguide sensors

    DEFF Research Database (Denmark)

    Skivesen, Nina

    by Qi et al [Zm Qi et al, Sens. Actuators B 81, 2002] before, however the sensing principle we present results in a broad detection range from gasses to solid materials and is different from the principle suggested by Qi et al with a highlylimited detection range. Metal-clad waveguide sensors......, where single cell detection isshown by use of the metal-clad waveguide sensors.......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...

  6. 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

  7. 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

  8. Waste Tank Vapor Program: Vapor space characterization of waste tank 241-T-111. Results from samples collected on January 20, 1995

    International Nuclear Information System (INIS)

    Klinger, G.S.; Clauss, T.W.; Ligotke, M.W.; Pool, K.H.; McVeety, B.D.; Olsen, K.B.; Bredt, O.P.; Fruchter, J.S.; Goheen, S.C.

    1995-10-01

    This document presents the details of the inorganic and organic analysis that was performed on samples from the headspace of Hanford waste tank 241-T-111. The results described were obtained to support the safety and toxicological evaluations. A summary of the results for the inorganic and organic analytes is included, as well as, a detailed description of the results which appears in the text

  9. 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.

  10. A comparative study of zircaloy and stainless steel as cladding material for pressurised water reactor

    International Nuclear Information System (INIS)

    Ganguly, C.; Roy, P.R.

    1977-01-01

    A comparative study of cladding material, stainless steel (Type 304, 347/348) and zircaloy (2 and 4) for oxide fuels of uranium, plutonium and mixed uranium-plutonium in pressurised light water cooled and moderated thermal reactor is made based on factors like: (i) neutrone economics, (ii) corrosion resistance, (iii) chemical interaction with fuel, impurities and fission products, (iv) pellet clad mechanical interaction- tensile, creep and fatigue properties, and (v) overall performance. Though the physical and mechanical properties and also the in-reactor performance record of stainless steel is better than zircaloy, the better neutron economics with zircaloy, because of its twelve times less neutron absorption cross section, overrides all the advantages of stainless steel. As a result all P.W.R.s are switching over to Zircaloy claddings. To achieve the failure rate as low as those for stainless steel the zircaloy fuel elements need modifications like use of smaller diameter pins with greater wall thickness, graphite of siloxane coating of clad internal surface, fuel pin pre-pressurisation, vacuum hot drying of pellets prior to loading and final end capping, use of short (length/diameter = 1 maximum), dished, chamferred, high density fuel pellets with uniformly distributed pores and enough diametral gap between pellet and clad. (author)

  11. 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])

  12. 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.)

  13. Uranium demand

    International Nuclear Information System (INIS)

    Roux, A.J.A.

    1976-01-01

    The estimated world resources of uranium as well as the estimated consumption of uranium over the next 25 years are briefly discussed. Attention is given to the prospecting of uranium in South Africa and elsewhere in the world

  14. 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)

  15. 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

  16. 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

  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. 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

  19. 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.

  20. 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)

  1. Fracture Toughness Of Zircaloy Claddings

    International Nuclear Information System (INIS)

    Bertsch, J.; Hoffelner, W

    2003-01-01

    Zirconium-based alloys (Zircaloy) have been used as cladding material in Light Water Reactors for many years. During fabrication, or in in-reactor service, crack-type defects can be formed, posing questions regarding mechanical integrity. As claddings change their mechanical properties (mainly toughness) during service as a result of irradiation-induced degradation, oxidation and hydride formation, it is essential for integrity considerations to provide parameters for the assessment of the influence of flaws on rupture behaviour. Usually, fracture-mechanics parameters are employed such as the fracture toughness, K IC , or, for high plastic strains, the J-integral, JIC. The applicability of these parameters is, however, limited by the dimensions of the samples (e.g. thickness). In claddings with a wall thickness of below 1 mm, determination of toughness necessitates an extension of the J-integral concept. A method based on the traditional J-approach, but applicable to thin-walled structures, is presented in this paper. (author)

  2. 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

  3. 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.

  4. 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.

  5. 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...

  6. 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.

  7. 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

  8. 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.

  9. Rigorous modeling of cladding modes in photonic crystal fibers

    DEFF Research Database (Denmark)

    Rindorf, Lars Henning; Bang, Ole

    We study the cladding modes of a photonic crystal fiber (PCF) with a finite size cladding using a finite element method. The cladding consists of seven rings of air holes with bulk silica outside.......We study the cladding modes of a photonic crystal fiber (PCF) with a finite size cladding using a finite element method. The cladding consists of seven rings of air holes with bulk silica outside....

  10. 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

  11. 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.

  12. Nuclear fuel elements having a composite cladding

    Science.gov (United States)

    Gordon, Gerald M.; Cowan, II, Robert L.; Davies, John H.

    1983-09-20

    An improved nuclear fuel element is disclosed for use in the core of nuclear reactors. The improved nuclear fuel element has a composite cladding of an outer portion forming a substrate having on the inside surface a metal layer selected from the group consisting of copper, nickel, iron and alloys of the foregoing with a gap between the composite cladding and the core of nuclear fuel. The nuclear fuel element comprises a container of the elongated composite cladding, a central core of a body of nuclear fuel material disposed in and partially filling the container and forming an internal cavity in the container, an enclosure integrally secured and sealed at each end of said container and a nuclear fuel material retaining means positioned in the cavity. The metal layer of the composite cladding prevents perforations or failures in the cladding substrate from stress corrosion cracking or from fuel pellet-cladding interaction or both. The substrate of the composite cladding is selected from conventional cladding materials and preferably is a zirconium alloy.

  13. Australia's uranium

    International Nuclear Information System (INIS)

    Hampson, D.C.

    1980-01-01

    The subject is discussed as follows: structure of the uranium industry in Australia (export policies; development of mining programme; table of export contracts approved by Australian government, 1972; government policy towards the industry 1972-75 and since 1975); reserves (table of Australia's major uranium deposits; estimated world resources of uranium, excluding USSR, Eastern Europe and China; comparison of exploration expenditures and discovery of uranium in Australia and the USA); enrichment; resource potential; future demand (table of nuclear power reactors above 30 MW in operation or under construction, mid-1979; projection of Australian uranium production to 1990); government and union action. (U.K.)

  14. TEC – Thin Environmental Cladding

    Directory of Open Access Journals (Sweden)

    Alan Tomasi

    2015-05-01

    Full Text Available Permasteelisa Group developed with Fiberline Composites a new curtain wall system (Thin Environmental Cladding or TEC, making use of pultruded GFRP (Glass Fiber Reinforced Polymer material instead of traditional aluminum. Main advantages using GFRP instead of aluminum are the increased thermal performance and the limited environmental impact. Selling point of the selected GFRP resin is the light transmission, which results in pultruded profiles that allow the visible light to pass through them, creating great aesthetical effects. However, GFRP components present also weaknesses, such as high acoustic transmittance (due to the reduced weight and anisotropy of the material, low stiffness if compared with aluminum (resulting in higher facade deflection and sensible fire behavior (as combustible material. This paper will describe the design of the TEC-facade, highlighting the functional role of glass within the facade concept with regards to its acoustic, structural, aesthetics and fire behavior.

  15. SELECTIVE SEPARATION OF URANIUM FROM FERRITIC STAINLESS STEELS

    Science.gov (United States)

    Beaver, R.J.; Cherubini, J.H.

    1963-05-14

    A process is described for separating uranium from a nuclear fuel element comprising a uranium-containing core and a ferritic stainless steel clad by heating said element in a non-carburizing atmosphere at a temperature in the range 850-1050 un. Concent 85% C, rapidly cooling the heated element through the temperature range 815 un. Concent 85% to 650 EC to avoid annealing said steel, and then contacting the cooled element with an aqueous solution of nitric acid to selectively dissolve the uranium. (AEC)

  16. Detection of fatigue cracks in cladded blocks

    International Nuclear Information System (INIS)

    Singh, G.P.; Cervantes, R.A.; Manning, R.C.; Takama, S.

    1986-01-01

    A nuclear reactor pressure vessel (RPV) operates at high temperatures. Feedwater nozzles are susceptible to thermal fatigue; and, after a large number of plant startup/shutdown cycles, thermal fatigue cracking may be initiated at these nozzles. In order to address this problem, ultrasonic data were acquired from five cladded specimens with overall approximate 4-mm thick stainless steel cladding; the specimens contained one fatigue crack each. The study evaluates the application of signal processing and pattern recognition methods to discriminate between base metal-to-clad interface signals and fatigue crack signals. Details are presented

  17. CALCULATION OF STRESS AND DEFORMATION IN FUEL ROD CLADDING DURING PELLET-CLADDING INTERACTION

    Directory of Open Access Journals (Sweden)

    Dávid Halabuk

    2015-12-01

    Full Text Available The elementary parts of every fuel assembly, and thus of the reactor core, are fuel rods. The main function of cladding is hermetic separation of nuclear fuel from coolant. The fuel rod works in very specific and difficult conditions, so there are high requirements on its reliability and safety. During irradiation of fuel rods, a state may occur when fuel pellet and cladding interact. This state is followed by changes of stress and deformations in the fuel cladding. The article is focused on stress and deformation analysis of fuel cladding, where two fuels are compared: a fresh one and a spent one, which is in contact with cladding. The calculations are done for 4 different shapes of fuel pellets. It is possible to evaluate which shape of fuel pellet is the most appropriate in consideration of stress and deformation forming in fuel cladding, axial dilatation of fuel, and radial temperature distribution in the fuel rod, based on the obtained results.

  18. 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.)

  19. 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

  20. 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.

  1. 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

  2. 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.

  3. 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....

  4. Plasmonic waveguides cladded by hyperbolic metamaterials.

    Science.gov (United States)

    Ishii, Satoshi; Shalaginov, Mikhail Y; Babicheva, Viktoriia E; Boltasseva, Alexandra; Kildishev, Alexander V

    2014-08-15

    Strongly anisotropic media with hyperbolic dispersion can be used for claddings of plasmonic waveguides (PWs). In order to analyze the fundamental properties of such waveguides, we analytically study 1D waveguides arranged from a hyperbolic metamaterial (HMM) in a HMM-Insulator-HMM (HIH) structure. We show that HMM claddings give flexibility in designing the properties of HIH waveguides. Our comparative study on 1D PWs reveals that HIH-type waveguides can have a higher performance than MIM or IMI waveguides.

  5. Uranium prospecting and uranium supply

    International Nuclear Information System (INIS)

    Kegel, K.E.

    1975-01-01

    Following of short historical survey, estimations of the uranium resources of the western world made in the middle of 1975 are presented and interpreted. The most common methods of prospecting and exploration of the mines and of production and processing of the uranium eres are described. A short survey of the situation of supply and demand is supplemented by a description of the activities of the two German companies in the field of uranium supply. (UA/AK) [de

  6. TEC – Thin Environmental Cladding

    Directory of Open Access Journals (Sweden)

    Alan Tomasi

    2014-06-01

    Full Text Available Corresponding author: Alan Tomasi, Group R&D Project Manager, Permasteelisa S.p.A., viale E. Mattei 21/23 | 31029 Vittorio Veneto, Treviso, Italy. Tel.: +39 0438 505207; E-mail: a.tomasi@permasteelisagroup.com; www.permasteelisagroup.com Permasteelisa Group developed with Fiberline Composites a new curtain wall system (Thin Environmental Cladding or TEC, making use of pultruded GFRP (Glass Fiber Reinforced Polymer material instead of traditional aluminum. Main advantages using GFRP instead of aluminum are the increased thermal performance and the limited environmental impact. Selling point of the selected GFRP resin is the light transmission, which results in pultruded profiles that allow the visible light to pass through them, creating great aesthetical effects. However, GFRP components present also weaknesses, such as high acoustic transmittance (due to the reduced weight and anisotropy of the material, low stiffness if compared with aluminum (resulting in higher facade deflection and sensible fire behavior (as combustible material. This paper will describe the design of the TEC-facade, highlighting the functional role of glass within the facade concept with regards to its acoustic, structural, aesthetics and fire behavior.

  7. 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.)

  8. 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

  9. 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

  10. 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.

  11. 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)

  12. 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

  13. 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)

  14. Potential effects of gallium on cladding materials

    Energy Technology Data Exchange (ETDEWEB)

    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.

  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. 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

  17. 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

  18. 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

  19. 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)

  20. 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)

  1. 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....

  2. 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

  3. 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

  4. 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.)

  5. 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

  6. 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

  7. 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.

  8. 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.

  9. 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.'

  10. Functionally graded materials produced by laser cladding

    NARCIS (Netherlands)

    Pei, Y.T.; Hosson, J.Th.M. De

    2000-01-01

    AlSi40 functionally graded materials (FGMs) were produced by a one-step laser cladding process on cast Al-alloy substrate as a possible solution for interfacial problems often present in laser coatings. The microstructure of the FGMs consists of a large amount of silicon primary particles surrounded

  11. Crud deposits on zircaloy-clad fuel

    International Nuclear Information System (INIS)

    Lister, D.H.

    1980-05-01

    The information on in-reactor corrosion of Zircaloy fuel cladding and crud deposition on fuel generated by Atomic Energy of Canada Limited up to 1979 has been reviewed elsewhere. This report is a summary of the crud deposition part of that review. (auth)

  12. 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)

  13. 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)

  14. Composite cladding for nuclear fuel elements

    International Nuclear Information System (INIS)

    Armijo, J.S.

    1975-01-01

    The composite cladding described is for nuclear reactors and comprises a zirconium alloy substrate, a metallurgically bonded metal barrier on the inner surface of the substrate, and a metallurgically bonded internal coating on the inner surface of the metal barrier. The metal of the barrier is selected from aluminium, niobium, copper, nickel, stainless steel or iron. The internal coating is zirconium alloy [fr

  15. 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.

  16. Dissolution of aluminium-cladded fuel elements

    International Nuclear Information System (INIS)

    Bernhard, G.; Boessert, W.; Hladik, O.; Schwarzbach, R.

    1984-01-01

    In the molybdenum production plant at Rossendorf (AMOR) short-term irradiated aluminium-cladded fuel elements from the Rossendorf research reactor RFR are dissolved for the purpose of molybdenum 99 production. The dissolution behaviour of these fuel elements and the appropriate dissolver are described. (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. Plasmonic waveguides cladded by hyperbolic metamaterials

    DEFF Research Database (Denmark)

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

    2014-01-01

    Strongly anisotropic media with hyperbolic dispersion can be used for claddings of plasmonic waveguides (PWs). In order to analyze the fundamental properties of such waveguides, we analytically study 1D waveguides arranged from a hyperbolic metamaterial (HMM) in a HMM-Insulator-HMM (HIH) structur...

  19. 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)

  20. Thick tool steel coatings with laser cladding

    NARCIS (Netherlands)

    Ocelik, V.; de Oliveira, U.; De Hosson, J. Th. M.; DeHosson, JTM; Brebbia, CA; Nishida, SI

    2007-01-01

    This paper concentrates on thick and crack-free laser clad coatings (up to 3 mm). The coating material is a chromium-molybdenum-tungsten-vanadium alloyed high-speed steel that shows high wear resistance, high compressive strength, good toughness, very good dimensional stability on heat treatment and

  1. 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

  2. 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

  3. 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

  4. 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

  5. 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.

  6. 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

  7. 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.

  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. 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)

  10. Uranium concentrates

    International Nuclear Information System (INIS)

    Boutonnet, G.

    1985-01-01

    Fabrication processes and main characteristics of the yellow coke are described. Choice of acidic or alkaline process, chemical reactions reagents, installation and efficiency are examined. Investment, cost and production are given. Extraction of uranium from phosphoric acid is briefly described [fr

  11. 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

  12. 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.

  13. 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)

  14. 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

  15. Microstructure of laser cladded martensitic stainless steel

    CSIR Research Space (South Africa)

    Van Rooyen, C

    2006-08-01

    Full Text Available of austenitic solidification. Table 4 - Chemical composition of the laser cladded martensitic stainless steel in the dendritic and interdendritic areas Material Area C* Cr Ni Mn Si Mo Dendritic 0.3 12.8 0.15 0.7 0.65 0.02 Fe211-1 (420) Off... alloy steels and are shown in Table 4. Ms (ºC) = 550 – 350C – 40Mn - 20Cr – 10Mo – 17Ni – 8W – 35V – 10Cu + 15Co + 30Al (Eq 3) Table 5 - Ms temperatures of laser cladded martensitic stainless steel Material Ms Dendritic area (ºC) Ms...

  16. Development of metal-clad filled evacuated panel superinsulation

    Energy Technology Data Exchange (ETDEWEB)

    Wilkes, K.E.; Strizak, J.P.; Weaver, F.J. [Oak Ridge National Lab., TN (United States); Besser, J.E.; Smith, D.L. [Aladdin Industries, Inc. (United States)

    1997-03-01

    This Cooperative Research and Development Agreement (CRADA) was between Aladdin Industries, Inc. and Lockheed Martin Energy Research Corp. The purpose of the CRADA was to determine the thermal performance of various metal claddings used to encapsulate Filled Evacuated Panel (FEP) superinsulation and to optimize the cost versus thermal performance of the claddings. A FEP superinsulation is a new type of superinsulation with the potential for saving large amounts of energy in buildings, building equipment, transportation (refrigerated railcars and trucks), industrial applications, etc. The major disadvantage of metal claddings for FEPs is the heat loss through the cladding caused by the high thermal conductivity of most metals. In smaller FEPs, this heat loss can degrade the overall performance of the FEP by factors of two or more as compared with polymer-clad FEPs. On the other hand, metal claddings are essentially impermeable to ambient air, whereas polymer claddings are not. Thus, the longevity and reliability of metal-clad FEPs are much superior to polymer-clad FEPs. In addition, because of the very low vapor pressure of metals as compared to polymers, metal-clad FEPs can achieve and operate at lower internal pressures. These lower pressures allow use of less expensive and/or higher performance filler materials.

  17. Characterization and Leach Testing for PUREX Cladding Waste Sludge (Group 3) and REDOX Cladding Waste Sludge (Group 4) Actual Waste Sample Composites

    Energy Technology Data Exchange (ETDEWEB)

    Snow, Lanee A.; Buck, Edgar C.; Casella, Amanda J.; Crum, Jarrod V.; Daniel, Richard C.; Draper, Kathryn E.; Edwards, Matthew K.; Fiskum, Sandra K.; Jagoda, Lynette K.; Jenson, Evan D.; Kozelisky, Anne E.; MacFarlan, Paul J.; Peterson, Reid A.; Swoboda, Robert G.

    2009-02-13

    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.(a) The testing program was subdivided into logical increments. The bulk water-insoluble solid wastes that are anticipated to be delivered to the Waste Treatment and Immobilization Plant (WTP) were identified according to type such that the actual waste testing could be targeted to the relevant categories. Eight broad waste groupings were defined. Samples available from the 222S archive were identified and obtained for testing. The actual wastetesting program included homogenizing the samples by group, characterizing the solids and aqueous phases, and performing parametric leaching tests. Two of the eight defined groups—plutonium-uranium extraction (PUREX) cladding waste sludge (Group 3, or CWP) and reduction-oxidation (REDOX) cladding waste sludge (Group 4, or CWR)—are the subjects of this report. Both the Group 3 and 4 waste composites were anticipated to be high in gibbsite, requiring caustic leaching. Characterization of the composite Group 3 and Group 4 waste samples confirmed them to be high in gibbsite. The focus of the Group 3 and 4 testing was on determining the behavior of gibbsite during caustic leaching. The waste-type definition, archived sample conditions, homogenization activities, characterization (physical, chemical, radioisotope, and crystal habit), and caustic leaching behavior as functions of time, temperature, and hydroxide concentration are discussed in this report. Testing was conducted according to TP-RPP-WTP-467.

  18. Conditioning of nuclear cladding wastes by melting

    International Nuclear Information System (INIS)

    Puyou, M.; Jouan, A.; Jacquet-Francillon, N.

    1991-01-01

    This paper discusses a cold-crucible induction melting process to condition cladding waste from irradiated fast breeder reactor fuel. The process has been developed by the CEA at Marcoule (France) as part of a major R and D program. It has been qualified at industrial scale on nonradioactive waste, and at laboratory scale on radioactive waste: several radioactive ingots have been produced from actual stainless steel or zircaloy hulls. The results confirm the numerous advantages of this containment method

  19. 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.

  20. Alloy development for high burnup cladding (PWR)

    International Nuclear Information System (INIS)

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

    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

  1. PREPARATION OF URANIUM MONOSULFIDE

    Science.gov (United States)

    Yoshioka, K.

    1964-01-28

    A process is given for preparing uranium monosulfide from uranium tetrafluoride dissolved in molten alkali metal chloride. A hydrogen-hydrogen sulfide gas mixture passed through the solution precipitates uranium monosulfide. (AEC)

  2. Recovery of uranium

    International Nuclear Information System (INIS)

    Clemens, D.H.; Walker, R.W.; Hurwitz, M.J.

    1974-01-01

    A process for recovering uranium from an uranium bearing liquid which comprises contacting the liquid with a crosslinkedvinyl benzyl chloride polymer, thereafter eluting from the resin the uranium in salt form

  3. 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

  4. 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)

  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. 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

  7. Uranium mineralogy

    International Nuclear Information System (INIS)

    Smith, D.K. Jr.

    1984-01-01

    The subject is discussed under the headings: introduction; U 4+ minerals (uraninite; coffinite; brannerite; ningyoite; lermontovite; U 4+ molybdates; U 4+ pyrochlores; U 4+ columbites; petscheckite and liandratite) minerals (uranyl oxide hydrates; uranyl silicates; uranyl phosphates and arsenates; uranyl vanadates; uranyl molybdates; uranyl sulphates; uranyl carbonates; uranyl selenates and tellurates). Appendices contain X-ray data and optical data for uranium minerals. (U.K.)

  8. 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.)

  9. 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.

  10. 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.

  11. 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

  12. 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

  13. Nuclear reactor fuel element with vanadium getter on cladding

    International Nuclear Information System (INIS)

    Johnson, C.E.; Carroll, K.G.

    1977-01-01

    A nuclear reactor fuel element is described which has an outer cladding, a central core of fissionable or mixed fissionable and fertile fuel material and a layer of vanadium as an oxygen getter on the inner surface of the cladding. The vanadium reacts with oxygen released by the fissionable material during irradiation of the core to prevent the oxygen from reacting with and corroding the cladding. Also described is a method for coating the inner surface of small diameter tubes of cladding with a layer of vanadium. 5 claims, 1 figure

  14. 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)

  15. Calculations of stresses in GCFR cladding under normal operating conditions

    International Nuclear Information System (INIS)

    Liu, Y.Y.; Hsieh, T.C.; Billone, M.C.

    1979-11-01

    A modified version of the LIFE-III code, LIFE-GCFR, and classical stress-analysis techniques have been used to calculate the stresses in the GCFR cladding under normal reactor operating conditions. Several types of loadings on the cladding that occur during normal operation have been considered. These include fuel-cladding mechanical interaction, thermal stresses induced by radial and axial temperature gradients, and swelling gradient-induced stresses. The combined and individual effects of these loadings, as well as the effect of creep on cladding stresses, have been assessed

  16. Finite-width plasmonic waveguides with hyperbolic multilayer cladding.

    Science.gov (United States)

    Babicheva, Viktoriia E; Shalaginov, Mikhail Y; Ishii, Satoshi; Boltasseva, Alexandra; Kildishev, Alexander V

    2015-04-20

    Engineering plasmonic metamaterials with anisotropic optical dispersion enables us to tailor the properties of metamaterial-based waveguides. We investigate plasmonic waveguides with dielectric cores and multilayer metal-dielectric claddings with hyperbolic dispersion. Without using any homogenization, we calculate the resonant eigenmodes of the finite-width cladding layers, and find agreement with the resonant features in the dispersion of the cladded waveguides. We show that at the resonant widths, the propagating modes of the waveguides are coupled to the cladding eigenmodes and hence, are strongly absorbed. By avoiding the resonant widths in the design of the actual waveguides, the strong absorption can be eliminated.

  17. 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

  18. 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

  19. 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.

  20. 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.

  1. DECONTAMINATION OF ZIRCALOY SPENT FUEL CLADDING HULLS

    Energy Technology Data Exchange (ETDEWEB)

    Rudisill, T; John Mickalonis, J

    2006-09-27

    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{sub 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{sub 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{sub 4}F)/ammonium nitrate (NH{sub 4}NO{sub 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{sub 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{sub 4}){sub 2}ZrF{sub 6}) on the metal surface prior to dissolution in the bulk solution could hinder the decontamination

  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...

  3. 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)

  4. 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

  5. Photonic lantern with cladding-removable fibers

    Science.gov (United States)

    Sun, Weimin; Yan, Qi; Bi, Yao; Yu, Haijiao; Liu, Xiaoqi; Xue, Jiuling; Tian, He; Liu, Yongjun

    2014-07-01

    Recently, spectral measurement becomes an important tool in astronomy to find exoplanets etc. The fibers are used to transfer light from the focal plate to spectrometers. To get high-resolution spectrum, the input slits of the spectrometers should be as narrow as possible. In opposite, the light spots from the fibers are circle, which diameters are clearly wider than the width of the spectrometer slits. To reduce the energy loss of the fiber-guide star light, many kinds of image slicers were designed and fabricated to transform light spot from circle to linear. Some different setup of fiber slicers are introduced by different research groups around the world. The photonic lanterns are candidates of fiber slicers. Photonic lantern includes three parts: inserted fibers, preform or tubing, taped part of the preform or tubing. Usually the optical fields concentrate in the former-core area, so the light spots are not uniform from the tapered end of the lantern. We designed, fabricated and tested a special kind of photonic lantern. The special fibers consist polymer cladding and doped high-index core. The polymer cladding could be easily removed using acetone bath, while the fiber core remains in good condition. We inserted the pure high-index cores into a pure silica tubing and tapered it. During the tapering process, the gaps between the inserted fibers disappeared. Finally we can get a uniform tapered multimode fiber end. The simulation results show that the longer the taper is, the lower the loss is. The shape of the taper should be controlled carefully. A large-zone moving-flame taper machine was fabricated to make the special photonic lantern. Three samples of photonic lanterns were fabricated and tested. The lanterns with cladding-removable fibers guide light uniform in the tapered ends that means these lanterns could collect more light from those ends.

  6. 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

  7. Analysis of pellet cladding mechanical interaction using computational simulation

    International Nuclear Information System (INIS)

    Berretta, José R.; Suman, Ricardo B.; Faria, Danilo P.; Rodi, Paulo A.; Giovedi, Claudia

    2017-01-01

    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)

  8. 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)

  9. COMPARISON OF CLADDING CREEP RUPTURE MODELS

    Energy Technology Data Exchange (ETDEWEB)

    P. Macheret

    2000-06-12

    The objective of this calculation is to compare several creep rupture correlations for use in calculating creep strain accrued by the Zircaloy cladding of spent nuclear fuel when it has been emplaced in the repository. These correlations are used to calculate creep strain values that are then compared to a large set of experimentally measured creep strain data, taken from four different research articles, making it possible to determine the best fitting correlation. The scope of the calculation extends to six different creep rupture correlations.

  10. 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

  11. 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

  12. 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)

  13. Stainless Steel Cladding Of Structural Steels By CO2 Laser Welding Techniques

    Science.gov (United States)

    Ludovico, A.; Daurelio, G.; Arcamone, O.

    1989-01-01

    Steel cladding processes are usually performed in different ways: hot roll cladding, strip cladding, weld cladding, explosion forming. For the first time, a medium power (2 KW c.w.) CO2 laser was used to clad structural steels (Fe 37C), 3 and 5 mm thick, with austenitic stainless steels (AISI 304 and AISI 316), 0.5 and 1.5 mm thick. The cladding technique we have developed uses the laser penetration welding process.

  14. 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.

  15. PRODUCTION OF URANIUM TETRACHLORIDE

    Science.gov (United States)

    Calkins, V.P.

    1958-12-16

    A process is descrlbed for the production of uranium tetrachloride by contacting uranlum values such as uranium hexafluoride, uranlum tetrafluoride, or uranium oxides with either aluminum chloride, boron chloride, or sodium alumlnum chloride under substantially anhydrous condltlons at such a temperature and pressure that the chlorldes are maintained in the molten form and until the uranium values are completely converted to uranlum tetrachloride.

  16. 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

  17. 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

  18. 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

  19. 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

  20. 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

  1. LANL Experience Rolling Zr-Clad LEU-10Mo Foils for AFIP-7

    Energy Technology Data Exchange (ETDEWEB)

    Hammon, Duncan L. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Clarke, Kester D. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Alexander, David J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Kennedy, Patrick K. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Edwards, Randall L. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Duffield, Andrew N. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Dombrowski, David E. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-05-29

    The cleaning, canning, rolling and final trimming of Low Enriched Uranium-10 wt. pct. Molybdenum (LEU-10Mo) foils for ATR (Advanced Test Reactor) fuel plates to be used in the AFIP-7 (ATR Full Size Plate In Center Flux Trap Position) experiments are summarized. Six Zr-clad foils were produced from two LEU-10Mo castings supplied to Los Alamos National Laboratory (LANL) by Y-12 National Security Complex. Details of cleaning and canning procedures are provided. Hot- and cold-rolling results are presented, including rolling schedules, images of foils in-process, metallography and local compositions of regions of interest, and details of final foil dimensions and process yield. This report was compiled from the slides for the presentation of the same name given by Duncan Hammon on May 12, 2011 at the AFIP-7 Lessons Learned meeting in Salt Lake City, UT, with Los Alamos National Laboratory document number LA-UR 11-02898.

  2. 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

  3. 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

  4. 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)

  5. Qualification of submerged-arc narrow strip cladding process

    International Nuclear Information System (INIS)

    Ayres, P.S.; Gottschling, J.D.; Jeffers, G.K.

    1975-08-01

    An unique narrow strip cladding process for use on both plate and forging material for nuclear components was developed. The qualification testing of this low-heat input process for cladding nuclear components, including those of SA508 Class 2 material is described. The theory that explains the acceptable results of these tests is also given. (auth)

  6. Qualification of submerged-arc narrow strip cladding process

    International Nuclear Information System (INIS)

    Ayres, P.S.; Gottschling, J.D.; Jeffers, G.K.

    1976-03-01

    Babcock and Wilcox has developed an unique narrow strip cladding process for use on both plate and forging material for nuclear components. The qualification testing of this low-heat input process for cladding nuclear components is described, including those of SA508 Class 2 material. The theory that explains the acceptable results of these tests is also given

  7. Laser cladding process development for high carbon steel substrates

    CSIR Research Space (South Africa)

    Lengopeng, T

    2014-11-01

    Full Text Available with an increase in laser power at constant scan speed. The hardness values taken on the HAZ showed a decrease with the increase in the number of clad layers during build-up. Meanwhile, four clad layer build up samples showed similar hardness trend independent...

  8. Interfacial adhesion of laser clad functionally graded materials

    NARCIS (Netherlands)

    Ocelik, V.; Pei, Y.T.; de Hosson, J.T.M.; Popoola, O; Dahotre, NB; Midea, SJ; Kopech, HM

    2003-01-01

    Two functionally graded coatings were prepared by different laser surface engineering techniques. Laser cladding of AlSi40 powder leads to the formation of functionally graded material (FGM) coating on AI-Si cast alloy substrate. Mapping of strain fields near the laser clad track using the digital

  9. 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...

  10. 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.)

  11. 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)

  12. 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.

  13. 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.

  14. 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)

  15. 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

  16. Compact cladding-pumped planar waveguide amplifier and fabrication method

    Science.gov (United States)

    Bayramian, Andy J.; Beach, Raymond J.; Honea, Eric; Murray, James E.; Payne, Stephen A.

    2003-10-28

    A low-cost, high performance cladding-pumped planar waveguide amplifier and fabrication method, for deployment in metro and access networks. The waveguide amplifier has a compact monolithic slab architecture preferably formed by first sandwich bonding an erbium-doped core glass slab between two cladding glass slabs to form a multi-layer planar construction, and then slicing the construction into multiple unit constructions. Using lithographic techniques, a silver stripe is deposited and formed at a top or bottom surface of each unit construction and over a cross section of the bonds. By heating the unit construction in an oven and applying an electric field, the silver stripe is then ion diffused to increase the refractive indices of the core and cladding regions, with the diffusion region of the core forming a single mode waveguide, and the silver diffusion cladding region forming a second larger waveguide amenable to cladding pumping with broad area diodes.

  17. 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)

  18. 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.

  19. Neutronic Analysis on Potential Accident Tolerant Fuel-Cladding Combination U3Si2-FeCrAl

    Directory of Open Access Journals (Sweden)

    Shengli Chen

    2017-01-01

    Full Text Available Neutronic performance is investigated for a potential accident tolerant fuel (ATF, which consists of U3Si2 fuel and FeCrAl cladding. In comparison with current UO2-Zr system, FeCrAl has a better oxidation resistance but a larger thermal neutron absorption cross section. U3Si2 has a higher thermal conductivity and a higher uranium density, which can compensate the reactivity suppressed by FeCrAl. Based on neutronic investigations, a possible U3Si2-FeCrAl fuel-cladding system is taken into consideration. Fundamental properties of the suggested fuel-cladding combination are investigated in a fuel assembly. These properties include moderator and fuel temperature coefficients, control rods worth, radial power distribution (in a fuel rod, and different void reactivity coefficients. The present work proves that the new combination has less reactivity variation during its service lifetime. Although, compared with the current system, it has a little larger deviation on power distribution and a little less negative temperature coefficient and void reactivity coefficient and its control rods worth is less important, variations of these parameters are less important during the service lifetime of fuel. Hence, U3Si2-FeCrAl system is a potential ATF candidate from a neutronic view.

  20. URANIUM DECONTAMINATION

    Science.gov (United States)

    Buckingham, J.S.; Carroll, J.L.

    1959-12-22

    A process is described for reducing the extractability of ruthenium, zirconium, and niobium values into hexone contained in an aqueous nitric acid uranium-containing solution. The solution is made acid-deficient, heated to between 55 and 70 deg C, and at that temperature a water-soluble inorganic thiosulfate is added. By this, a precipitate is formed which carries the bulk of the ruthenium, and the remainder of the ruthenium as well as the zirconium and niobium are converted to a hexone-nonextractable form. The rutheniumcontaining precipitate can either be removed from the solu tion or it can be dissolved as a hexone-non-extractable compound by the addition of sodium dichromate prior to hexone extraction.

  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. Cladding material, tube including such cladding material and methods of forming the same

    Science.gov (United States)

    Garnier, John E.; Griffith, George W.

    2016-03-01

    A multi-layered cladding material including a ceramic matrix composite and a metallic material, and a tube formed from the cladding material. The metallic material forms an inner liner of the tube and enables hermetic sealing of thereof. The metallic material at ends of the tube may be exposed and have an increased thickness enabling end cap welding. The metallic material may, optionally, be formed to infiltrate voids in the ceramic matrix composite, the ceramic matrix composite encapsulated by the metallic material. The ceramic matrix composite includes a fiber reinforcement and provides increased mechanical strength, stiffness, thermal shock resistance and high temperature load capacity to the metallic material of the inner liner. The tube may be used as a containment vessel for nuclear fuel used in a nuclear power plant or other reactor. Methods for forming the tube comprising the ceramic matrix composite and the metallic material are also disclosed.

  3. 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)

  4. 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

  5. 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

  6. 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

  7. 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

  8. Hollow Core Photonic Crystal Fibre Comprising a Fibre Grating in the Cladding and its Applications

    DEFF Research Database (Denmark)

    2010-01-01

    An optical fibre is provided having a fibre cladding around a longitudinally extending optical propagation core. The cladding has a reflection region of a varying refractive index in the longitudinal direction.......An optical fibre is provided having a fibre cladding around a longitudinally extending optical propagation core. The cladding has a reflection region of a varying refractive index in the longitudinal direction....

  9. Cladding development and characterization in the United States

    International Nuclear Information System (INIS)

    Lobsinger, R.J.

    1977-01-01

    During the past decade, an extensive program of cladding development has been carried out at HEDL in support of the FFTF/Breeder Reactor Program. This activity has resulted in the establishment of specification, fabrication, and characterization requirements for breeder reactor fuel cladding. In establishing these requirements, input has been derived from vendor development programs, evaluation studies, and reviews with other national laboratories. The major fabrication requirements of the current specification as derived from the development program and the testing required to assure product conformance and characterize cladding behavior are described

  10. 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

  11. Deep-probe metal-clad waveguide biosensors

    DEFF Research Database (Denmark)

    Skivesen, Nina; Horvath, Robert; Thinggaard, S.

    2007-01-01

    -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.......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...

  12. Long-range plasmonic waveguides with hyperbolic cladding.

    Science.gov (United States)

    Babicheva, Viktoriia E; Shalaginov, Mikhail Y; Ishii, Satoshi; Boltasseva, Alexandra; Kildishev, Alexander V

    2015-11-30

    We study plasmonic waveguides with dielectric cores and hyperbolic multilayer claddings. The proposed design provides better performance in terms of propagation length and mode confinement in comparison to conventional designs, such as metal-insulator-metal and insulator-metal-insulator plasmonic 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 coupled, and the propagation length can be on the order of a millimeter.

  13. 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

  14. Experiments for evaluation of corrosion to develop storage criteria for interim dry storage of aluminum-alloy clad spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Peacock, H.B.; Sindelar, R.L.; Lam, P.S.; Murphy, T.H.

    1994-11-01

    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.

  15. Treatment of Corroded Metallic Uranium Fuel

    International Nuclear Information System (INIS)

    Stridsman, H.; Ekeroth, E.; Hallberg, B.; Hellsten, E.; Lindberg, M.; Nordlinder, S.

    2009-01-01

    This paper describes the extensive planning and treatment of corroded metallic uranium fuel performed in Studsvik in 2007. This included conversion of possible pyrophoric uranium hydride to uranium oxide and separation of intact parts of fuel rods from corrosion products. The first nuclear reactor in Sweden was the R1 reactor in Stockholm, with fuel of natural metallic uranium. After shut-down of the reactor, part of the fuel had been placed in waterproof canisters in an interim storage pool at Studsvik. In 1988, corrosion of the fuel had been discovered in one of the canisters due to water leakage. Exposure of metallic uranium to water causes pyrophoric uranium hydride to be formed, which poses a severe risk of fire or explosion if subjected to oxygen in air. The canister with corroded fuel had then been placed in an outer container to which inert gas with low oxygen concentration was supplied, in order to slowly oxidize hydride. A project to take care of the corroded fuel was set up in 2006, comprising several preparatory steps: research, equipment design, pre-treatment and safety-case demonstrations. Treatment comprised two stages, pre-treatment at the storage location, and final treatment in a hot-cell. The pre-treatment stage was performed in order to study the reaction inside the container with a controlled oxygen supply and regularly control of atmosphere inside the container. In addition, equipment was developed for controlled movement of the container. The results from the pre-treatment stage were used for the planning and acceptance of the transport to and handling in the hot-cell facility in Studsvik. The work performed in the hot-cell consisted of opening the container, sequentially cutting the container and canister and separation of intact parts of fuel rods from the corrosion products and cladding, top and bottoms of aluminium. No evidence of uranium hydride was found throughout the operation. The corrosion products was expected to be in form of

  16. Modeling of the non-monotonous viscoplastic behavior of uranium dioxide

    Science.gov (United States)

    Sauter, F.; Leclercq, S.

    2003-10-01

    In order to evaluate the stress level during pellet cladding mechanical interaction (PCMI), Electricité de France is involved in a large program of investigation of the mechanical properties of both fuel pellets and claddings. In this paper, we focus on the mechanical behavior of uranium dioxide. These pellets exhibit a yield point during strain hardening tests and sigmoı̈dal creep curves, that is inflection points characteristic of a non-monotonous viscoplastic flow in the first stages of the tests. Inspired by Alexander and Haasen's work upon single crystal silicon, we develop a dislocation-based model that is able to describe the viscoplasticity of uranium dioxide in the range of temperature and stress of the PCMI. After developing this model, we introduce it into the Pilvin's polycrystalline approach. The self-consistency of the polycrystalline approach in the case of a non-monotonous viscoplastic flow is demonstrated in an independent article.

  17. 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

  18. 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...

  19. 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

  20. 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

  1. 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

  2. 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

  3. 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

  4. 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

  5. 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

  6. Characteristics Of KRS-5 Fiber With Crystalline Cladding

    Science.gov (United States)

    Kimura, M.; Kachi, S.; Shiroyama, K.

    1986-05-01

    KRS-5 polycrystalline fibers having a core-cladding structure have been fabricated by complex extrusion. The core material is KRS-5 (T1Br-TlI), and the cladding material is KRS-6 (T1Br-T1C1). Through thermal treatment after fiber fabrication, the refractive index profile is changed to the graded index (GI) type. The attenuation loss of these core-cladding GI fibers is 0.2 dB/m at about 10.6 μm (lauching NA = 0.05). Fibers with this core-cladding structure will enable wide applications using far infrared wavelength light, for example in CO2, laser processing machines and spectroscopy equipment, because of their ease in handling.

  7. 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

  8. High Temperature Resistance Claddings for Nuclear Thermal Rockets, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — This program will develop a series of nano-/micro-composite coated nuclear reactor facing components using MesoCoat's CermaCladTM process. This proposed SBIR program...

  9. 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.

  10. 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...

  11. Finite-width plasmonic waveguides with hyperbolic multilayer cladding

    DEFF Research Database (Denmark)

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

    2015-01-01

    Engineering plasmonic metamaterials with anisotropic optical dispersion enables us to tailor the properties of metamaterial-based waveguides. We investigate plasmonic waveguides with dielectric cores and multilayer metal-dielectric claddings with hyperbolic dispersion. Without using any homogeniz...

  12. High Resolution Temperature Estimation During Laser Cladding of Stainless Steel

    Science.gov (United States)

    Devesse, Wim; De Baere, Dieter; Hinderdael, Michaël; Guillaume, Patrick

    Laser cladding is a technique that is used for the coating, repair and production of metallic parts. Material is added to the surface of the part by injecting a flow of powder into a melt pool that is created with a high power laser beam. When the beam scans the surface of the substrate, strong local heating and cooling results. A good knowledge of the temperature distribution history during the laser cladding process is vital to predict and optimize the material properties of the final part. This paper presents a contactless temperature measurement system with high temporal and spatial resolution based on a hyperspectral line camera. High temperature measurements were made during laser cladding of AISI 316L stainless steel. A good correlation is shown between the temperature measurements and microscope images taken after creation of the clad.

  13. Uranium: the recalcitrant commodity

    International Nuclear Information System (INIS)

    Crowson, P.C.F.

    1989-01-01

    During the past year dollar prices of virtually every mineral commodity have risen considerably on the back of strong demand, falling stocks and rising capacity utilization. Uranium has experienced the same economic influences, but measures of conditions in uranium spot markets display drift and even further decline. Does this mean that uranium is not really a commodity after all, or have special factors been at work? Some of the economic features of the uranium market are re-examined in this context. (author)

  14. 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.)

  15. PREPARATION OF URANIUM HEXAFLUORIDE

    Science.gov (United States)

    Lawroski, S.; Jonke, A.A.; Steunenberg, R.K.

    1959-10-01

    A process is described for preparing uranium hexafluoride from carbonate- leach uranium ore concentrate. The briquetted, crushed, and screened concentrate is reacted with hydrogen fluoride in a fluidized bed, and the uranium tetrafluoride formed is mixed with a solid diluent, such as calcium fluoride. This mixture is fluorinated with fluorine and an inert diluent gas, also in a fluidized bed, and the uranium hexafluoride obtained is finally purified by fractional distillation.

  16. 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

  17. 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

  18. 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

  19. Double-clad nuclear-fuel safety rod

    Science.gov (United States)

    McCarthy, W.H.; Atcheson, D.B.

    1981-12-30

    A device for shutting down a nuclear reactor during an undercooling or overpower event, whether or not the reactor's scram system operates properly. This is accomplished by double-clad fuel safety rods positioned at various locations throughout the reactor core, wherein melting of a secondary internal cladding of the rod allows the fuel column therein to shift from the reactor core to place the reactor in a subcritical condition.

  20. Interim report on the creepdown of Zircaloy fuel cladding

    International Nuclear Information System (INIS)

    Hobson, D.O.; Dodd, C.V.

    1977-01-01

    This report describes the creepdown phenomenon in Zircaloy fuel cladding and the methods by which it will be measured and analyzed. Instrumentation for monitoring radial deformation in the cladding is described in detail--in terms of theory, design, and stability. The programs that control the microcomputer are listed, both to document the level of sophistication of the instrumentation and to indicate the flexibility of the test equipment

  1. 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

  2. Special techniques for tensile tests of irradiated zirconium claddings

    International Nuclear Information System (INIS)

    Prokhorov, V.I.; Makarov, O.Ju.; Smirnov, V.P.

    2002-01-01

    Irradiated zirconium alloy claddings possessing property anisotropy should be tested in transverse and longitudinal directions. Such mechanical tests can be performed in conditions of large variety of geometric peculiarities of specimens, supports or grips. The objective of the work is the development of the unified complex of updated special techniques that allow investigation of mechanical pre- and post-irradiation properties of VVER claddings including radiation effect of property anisotropy changes in the same way. (author)

  3. Laser cladding crack repair of austenitic stainless steel

    CSIR Research Space (South Africa)

    Van Rooyen, C

    2009-06-01

    Full Text Available of pressurised vessel Fig. 2: Overlay positional welding Fig. 3: Crack sealing without peening prior to cladding 3 Hammer peening was introduced in order to mechanically seal of squirting water prior to crack sealing. Leaks were reduced from.... Simulated crack repair and overlay cladding with 316L is shown in Fig. 6. Three small leaking pores were observed in the first layer, indicated by arrows. Hammer peening was applied to the first layer to mechanically seal the leaking pores prior...

  4. 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.)

  5. 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.)

  6. 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

  7. 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

  8. Evolutionary developments of advanced PWR nuclear fuels and cladding materials

    International Nuclear Information System (INIS)

    Kim, Kyu-Tae

    2013-01-01

    Highlights: • PWR fuel and cladding materials development processes are provided. • Evolution of PWR advanced fuel in U.S.A. and in Korea is described. • Cutting-edge design features against grid-to-rod fretting and debris are explained. • High performance data of advanced grids, debris filters and claddings are given. -- Abstract: The evolutionary developments of advanced PWR fuels and cladding materials are explained with outstanding design features of nuclear fuel assembly components and zirconium-base cladding materials. The advanced PWR fuel and cladding materials development processes are also provided along with verification tests, which can be used as guidelines for newcomers planning to develop an advanced fuel for the first time. The up-to-date advanced fuels with the advanced cladding materials may provide a high level of economic utilization and reliable performance even under current and upcoming aggressive operating conditions. To be specific, nuclear fuel vendors may achieve high fuel burnup capability of between 45,000 and 65,000 MWD/MTU batch average, overpower thermal margin of as much as 15% and longer cycle length up to 24 months on the one hand and fuel failure rates of around 10 −6 on the other hand. However, there is still a need for better understanding of grid-to-rod fretting wear mechanisms leading to major PWR fuel defects in the world and subsequently a driving force for developing innovative spacer grid designs with zero fretting wear-induced fuel failure

  9. Laser cladding: repairing and manufacturing metal parts and tools

    Science.gov (United States)

    Sexton, Leo

    2003-03-01

    Laser cladding is presently used to repair high volume aerospace, automotive, marine, rail or general engineering components where excessive wear has occurred. It can also be used if a one-off high value component is either required or has been accidentally over-machined. The ultimate application of laser cladding is to build components up from nothing, using a laser cladding system and a 3D CAD drawing of the component. It is thus emerging that laser cladding can be classified as a special case of Rapid Prototyping (RP). Up to this point in time RP was seen, and is still seen, as in intermediately step between the design stage of a component and a finished working product. This can now be extended so that laser cladding makes RP a one-stop shop and the finished component is made from tool-steel or some alloy-base material. The marriage of laser cladding with RP is an interesting one and offers an alternative to traditional tool builders, re-manufacturers and injection mould design/repair industries. The aim of this paper is to discuss the emergence of this new technology, along with the transference of the process out of the laboratory and into the industrial workplace and show it is finding its rightful place in the manufacturing/repair sector. It will be shown that it can be used as a cost cutting, strategic material saver and consequently a green technology.

  10. Real-time laser cladding control with variable spot size

    Science.gov (United States)

    Arias, J. L.; Montealegre, M. A.; Vidal, F.; Rodríguez, J.; Mann, S.; Abels, P.; Motmans, F.

    2014-03-01

    Laser cladding processing has been used in different industries to improve the surface properties or to reconstruct damaged pieces. In order to cover areas considerably larger than the diameter of the laser beam, successive partially overlapping tracks are deposited. With no control over the process variables this conduces to an increase of the temperature, which could decrease mechanical properties of the laser cladded material. Commonly, the process is monitored and controlled by a PC using cameras, but this control suffers from a lack of speed caused by the image processing step. The aim of this work is to design and develop a FPGA-based laser cladding control system. This system is intended to modify the laser beam power according to the melt pool width, which is measured using a CMOS camera. All the control and monitoring tasks are carried out by a FPGA, taking advantage of its abundance of resources and speed of operation. The robustness of the image processing algorithm is assessed, as well as the control system performance. Laser power is decreased as substrate temperature increases, thus maintaining a constant clad width. This FPGA-based control system is integrated in an adaptive laser cladding system, which also includes an adaptive optical system that will control the laser focus distance on the fly. The whole system will constitute an efficient instrument for part repair with complex geometries and coating selective surfaces. This will be a significant step forward into the total industrial implementation of an automated industrial laser cladding process.

  11. A model for hydrogen pickup for BWR cladding materials

    International Nuclear Information System (INIS)

    Hede, G.; Kaiser, U.

    2001-01-01

    It has been observed that rod elongation is driven by the hydrogen pickup but not by corrosion as such. Based on this a non-destructive method to determine clad hydrogen concentration has been developed. The method is based on the observation that there are three different mechanisms behind the rod growth: the effect of neutron irradiation on the Zircaloy microstructure, the volume increase of the cladding as an effect of hydride precipitation and axial pellet-cladding-mechanical-interaction (PCMI). The derived correlation is based on the experience of older cladding materials, inspected at hot-cell laboratories, that obtained high hydrogen levels (above 500 ppm) at lower burnup (assembly burnup below 50 MWd/kgU). Now this experience can be applied, by interpolation, on more modern cladding materials with a burnup beyond 50 MWd/kgU by analysis of the rod growth database of the respective cladding materials. Hence, the method enables an interpolation rather than an extrapolation of present day hydrogen pickup database, which improves the reliability and accuracy. Further, one can get a good estimate of the hydrogen pickup during an ongoing outage based on a non-destructive method. Finally, rod growth measurements are normally performed for a large population of rods, hence giving a good statistics compared to examination of a few rods at a hot cell. (author)

  12. Cladding Attachment Over Thick Exterior Insulating Sheathing

    Energy Technology Data Exchange (ETDEWEB)

    Baker, P. [Building Science Corporation, Somerville, MA (United States); Eng, P. [Building Science Corporation, Somerville, MA (United States); Lepage, R. [Building Science Corporation, Somerville, MA (United States)

    2014-01-01

    The addition of insulation to the exterior of buildings is an effective means of increasing the thermal resistance of both wood framed walls as well as mass masonry wall assemblies. For thick layers of exterior insulation (levels greater than 1.5 inches), the use of wood furring strips attached through the insulation back to the structure has been used by many contractors and designers as a means to provide a convenient cladding attachment location (Straube and Smegal 2009, Pettit 2009, Joyce 2009, Ueno 2010). The research presented in this report is intended to help develop a better understanding of the system mechanics involved and the potential for environmental exposure induced movement between the furring strip and the framing. BSC sought to address the following research questions: 1.What are the relative roles of the mechanisms and the magnitudes of the force that influence the vertical displacement resistance of the system? 2.Can the capacity at a specified deflection be reliably calculated using mechanics based equations? 3.What are the impacts of environmental exposure on the vertical displacement of furring strips attached directly through insulation back to a wood structure?

  13. Cladding Attachment Over Thick Exterior Insulating Sheathing

    Energy Technology Data Exchange (ETDEWEB)

    Baker, P. [Building Science Corporation, Somerville, MA (United States); Eng, P. [Building Science Corporation, Somerville, MA (United States); Lepage, R. [Building Science Corporation, Somerville, MA (United States)

    2014-01-01

    The addition of insulation to the exterior of buildings is an effective means of increasing the thermal resistance of both wood framed walls as well as mass masonry wall assemblies. For thick layers of exterior insulation (levels greater than 1.5 inches), the use of wood furring strips attached through the insulation back to the structure has been used by many contractors and designers as a means to provide a convenient cladding attachment location (Straube and Smegal 2009, Pettit 2009, Joyce 2009, Ueno 2010). The research presented in this report is intended to help develop a better understanding of the system mechanics involved and the potential for environmental exposure induced movement between the furring strip and the framing. BSC sought to address the following research questions: 1. What are the relative roles of the mechanisms and the magnitudes of the force that influence the vertical displacement resistance of the system? 2. Can the capacity at a specified deflection be reliably calculated using mechanics based equations? 3. What are the impacts of environmental exposure on the vertical displacement of furring strips attached directly through insulation back to a wood structure?

  14. 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.

  15. URANIUM SEPARATION PROCESS

    Science.gov (United States)

    Hyde, E.K.; Katzin, L.I.; Wolf, M.J.

    1959-07-14

    The separation of uranium from a mixture of uranium and thorium by organic solvent extraction from an aqueous solution is described. The uranium is separrted from an aqueous mixture of uranium and thorium nitrates 3 N in nitric acid and containing salting out agents such as ammonium nitrate, so as to bring ihe total nitrate ion concentration to a maximum of about 8 N by contacting the mixture with an immiscible aliphatic oxygen containing organic solvent such as diethyl carbinol, hexone, n-amyl acetate and the like. The uranium values may be recovered from the organic phase by back extraction with water.

  16. Uranium uptake by plants

    International Nuclear Information System (INIS)

    Singh, K.P.

    1997-01-01

    This paper highlights the transport of uranium present in the soil to plants. An increase in the uranium content in soil enhances its transport in various parts of plants. The transport of uranium from the soil to the grain follows the order: black gram>maize>lentil>chick-pea>rice>wheat. In certain vegetables and fruits, this order is: spinach>carrot>radish> brinjal>banana>tomato>beet. In vegetables and fruits, the stem reflects minimum percentage of uranium present in the soil. The uranium transport is appreciably high in arecanut plant. The chances of uranium transport to the human organs, are expected to be more through consumption of crops grown in uranium-rich soil. (author)

  17. Effects of cold worked and fully annealed claddings on fuel failure behaviour

    International Nuclear Information System (INIS)

    Saito, Shinzo; Hoshino, Hiroaki; Shiozawa, Shusaku; Yanagihara, Satoshi

    1979-12-01

    Described are the results of six differently heat-treated Zircaloy clad fuel rod tests in NSRR experiments. The purpose of the test is to examine the extent of simulating irradiated claddings in mechanical properties by as-cold worked ones and also the effect of fully annealing on the fuel failure bahaviour in a reactivity initiated accident (RIA) condition. As-cold worked cladding does not properly simulated the embrittlement of the irradiated one in a RIA condition, because the cladding is fully annealed before the fuel failure even in the short transient. Therefore, the fuel behaviour such as fuel failure threshold energy, failure mechanism, cladding deformation and cladding oxidation of the fully annealed cladding fuel, as well as that of the as-cold worked cladding fuel, are not much different from that of the standard stress-relieved cladding fuel. (author)

  18. Corrosion Resistance Evaluation of HANA Claddings in Commercial PWR

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Hee-Hun; Kwon, Oh-Hyun; Kim, Hong-Jin; Yoo, Jong-Sung; Kim, Yong-Hwan [KEPCO NF, Daejeon (Korea, Republic of)

    2014-10-15

    Korea Atomic Energy Research Institute (KAERI) in collaboration with KEPCO Nuclear Fuel (KNF) developed newly-advanced alloy which are named HANA (High-performance Alloy for Nuclear Application) for high burnup PWR nuclear fuel, showed an excellent out-pile corrosion resistance in PWR simulating loop conditions. And in-pile corrosion resistance of HANA claddings, which was examined at the first provisional inspection after -185 FPD of irradiation in the Halden Reactor, and also shown superior to the other references alloy. Also, other researches showed a much better corrosion resistance when compared to the other Zr-based alloy in various corrosion conditions. In this study, the LTA program for newly-developed fuel assembly (HIPER) with the HANA claddings was implemented to justify the performance for 3 cycles of operation schedule in Hanul nuclear power plant. The objective of this study is to compare corrosion properties of reference alloy with HANA claddings loaded in Hanul nuclear power plant.. For the examination procedures, the oxide thickness measurements method and equipment of PSE are described in detail as follow in measurement methods chapter. Finally, based on the above mentioned measurements method, the summarized oxide thickness data obtained from PSE are evaluated for the corrosion resistance in commercial nuclear power plant and some discussion for the corrosion resistance are described. In the past, corrosion resistance of HANA claddings was successfully conducted in test reactor. In this study, the corrosion characteristic of HANA claddings which are applied to HIPER is examined in the commercial nuclear power plant. HANA claddings in the HIPER showed a more improved corrosion resistance than reference alloy claddings and are evaluated well with meeting the oxide thickness criteria.

  19. Geology of hydrothermal uranium deposits

    International Nuclear Information System (INIS)

    Korolev, K.G.; Belov, V.K.; Putilov, G.S.

    1983-01-01

    Geological characteristics of hydrothermal phosphorus-uranium deposits placed in sedimentary, igneous-sedimentary, metamorphic and intrusion formations are presented. Attention is paid to mineral composition, texture and structure of ores, their genesis, tectonics. Geochemical peculiarities of ores and age of molybdenum-uranium and uranium deposits are described. Geological criteria and prospecting features of uranium and uranium-molybdenum deposits are given

  20. Development of advanced LWR fuel cladding

    International Nuclear Information System (INIS)

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

    2000-04-01

    This report describes the results from evaluating the preliminary Zr-based alloys to develop the advanced Zr-based alloys for the nuclear fuel claddings, which should have good corrosion resistance and mechanical properties at high burn-up over 70,000MWD/MTU. It also includes the results from the basic studies for optimizing the processes which are involved in the development of the advanced Zr-based alloys. Ten(10) kinds of candidates for the alloys of which performance is over that of the existing Zircaloy-4 or ZIRLO alloy were selected out of the preliminary alloys of 150 kinds which were newly designed and repeatedly manufactured and evaluated to find out the promising alloys. First of all, the corrosion tests on the preliminary alloys were carried out to evaluate their performance in both pure water and LiOH solution at 360 deg C and in steam at 400 deg C. The tensile tests were performed on the alloys which proved to be good in the corrosion resistance. The creep behaviors were tested at 400 deg C for 10 days with the application of constant load on the samples which showed good performance in the corrosion resistance and tensile properties. The effect of the final heat treatment and A-parameters as well as Sn or Nb on the corrosion resistance, tensile properties, hardness, microstructures of the alloys was evaluated for some alloys interested. The other basic researches on the oxides, electrochemical properties, corrosion mechanism, and the establishment of the phase diagrams of some alloys were also carried out

  1. Development of advanced LWR fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Yong Hwan; Park, S. Y.; Lee, M. H. [and others

    2000-04-01

    This report describes the results from evaluating the preliminary Zr-based alloys to develop the advanced Zr-based alloys for the nuclear fuel claddings, which should have good corrosion resistance and mechanical properties at high burn-up over 70,000MWD/MTU. It also includes the results from the basic studies for optimizing the processes which are involved in the development of the advanced Zr-based alloys. Ten(10) kinds of candidates for the alloys of which performance is over that of the existing Zircaloy-4 or ZIRLO alloy were selected out of the preliminary alloys of 150 kinds which were newly designed and repeatedly manufactured and evaluated to find out the promising alloys. First of all, the corrosion tests on the preliminary alloys were carried out to evaluate their performance in both pure water and LiOH solution at 360 deg C and in steam at 400 deg C. The tensile tests were performed on the alloys which proved to be good in the corrosion resistance. The creep behaviors were tested at 400 deg C for 10 days with the application of constant load on the samples which showed good performance in the corrosion resistance and tensile properties. The effect of the final heat treatment and A-parameters as well as Sn or Nb on the corrosion resistance, tensile properties, hardness, microstructures of the alloys was evaluated for some alloys interested. The other basic researches on the oxides, electrochemical properties, corrosion mechanism, and the establishment of the phase diagrams of some alloys were also carried out.

  2. Uranium of Kazakhstan

    International Nuclear Information System (INIS)

    Tsalyuk, Yu.; Gurevich, D.

    2000-01-01

    Over 25 % of the world's uranium reserves are concentrated in Kazakhstan. So, the world's largest Shu-Sarysu uranium province is situated on southern Kazakhstan, with resources exceeding 1 billion tonnes of uranium. No less, than 3 unique deposits with resources exceeding 100,000 tonnes are situated here. From the economic point of view the most important thing is that these deposits are suitable for in-situ leaching, which is the cheapest, environmentally friendly and most efficient method available for uranium extracting. In 1997 the Kazatomprom National Joint-Stock Company united all Kazakhstan's uranium enterprises (3 mine and concentrating plants, Volkovgeologiya Joint-Stock Company and the Ulbinskij Metallurgical plant). In 1998 uranium production came to 1,500 tonnes (860 kg in 1997). In 1999 investment to the industry were about $ 30 million. Plans for development of Kazakhstan's uranium industry provide a significant role for foreign partners. At present, 2 large companies (Comeco (Canada), Cogema (France) working in Kazakhstan. Kazakatomprom continues to attract foreign investors. The company's administration announced that in that in next year they have plan to make a radical step: to sell 67 % of stocks to strategic investors (at present 100 % of stocks belongs to state). Authors of the article regard, that the Kazakhstan's uranium industry still has significant reserves to develop. Even if the scenario for the uranium industry could be unfavorable, uranium production in Kazakhstan may triple within the next three to four years. The processing of uranium by the Ulbinskij Metallurgical Plant and the production of some by-products, such as rhenium, vanadium and rare-earth elements, may provide more profits. Obviously, the sale of uranium (as well as of any other reserves) cannot make Kazakhstan a prosperous country. However, country's uranium industry has a god chance to become one of the most important and advanced sectors of national economy

  3. Bioremediation of uranium contamination with enzymatic uranium reduction

    Science.gov (United States)

    Lovley, D.R.; Phillips, E.J.P.

    1992-01-01

    Enzymatic uranium reduction by Desulfovibrio desulfuricans readily removed uranium from solution in a batch system or when D. desulfuricans was separated from the bulk of the uranium-containing water by a semipermeable membrane. Uranium reduction continued at concentrations as high as 24 mM. Of a variety of potentially inhibiting anions and metals evaluated, only high concentrations of copper inhibited uranium reduction. Freeze-dried cells, stored aerobically, reduced uranium as fast as fresh cells. D. desulfuricans reduced uranium in pH 4 and pH 7.4 mine drainage waters and in uraniumcontaining groundwaters from a contaminated Department of Energy site. Enzymatic uranium reduction has several potential advantages over other bioprocessing techniques for uranium removal, the most important of which are as follows: the ability to precipitate uranium that is in the form of a uranyl carbonate complex; high capacity for uranium removal per cell; the formation of a compact, relatively pure, uranium precipitate.

  4. The quest for safe and reliable fuel cladding materials

    Energy Technology Data Exchange (ETDEWEB)

    Pino, Eddy S.; Abe, Alfredo Y., E-mail: eddypino132@hotmail.com [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Giovedi, Claudia, E-mail: claudia.giovedi@labrisco.usp.br [Universidade de Sao Paulo (POLI/USP), Sao Paulo, SP (Brazil). Lab. de Analise, Avaliacao e Gerenciamento de Risco

    2015-07-01

    The tragic Fukushima Daiichi Nuclear Plant accident of March, 2011, has brought great unrest and challenge to the nuclear industry, which, in collaboration with universities and nuclear research institutes, is making great efforts to improve the safety in nuclear reactors developing accident tolerant fuels (ATF). This involves the study of different materials to be applied as cladding and, also, the improvement in the fuel properties in order to enhance the fuel performance and safety, specifically under accident conditions. Related to the cladding, iron based alloys and silicon carbide (SiC) materials have been studied as a good alternative. In the case of austenitic stainless steel, there is the advantage that the austenitic stainless steel 304 was used as cladding material in the first PWR (Pressurized Water Reactor) registering a good performance. Then, alternated cladding materials such as iron based alloys (304, 310, 316, 347) should be used to replace the zirconium-based alloys in order to improve safety. In this paper, these cladding materials are evaluated in terms of their physical and chemical properties; among them, strength and creep resistance, thermal conductivity, thermal stability and corrosion resistance. Additionally, these properties are compared with those of conventional zirconium-based alloys, the most used material in actual PWR, to assess the advantages and disadvantages of each material concerning to fuel performance and safety contribution. (author)

  5. POST CRITICAL HEAT TRANSFER AND FUEL CLADDING OXIDATION

    Directory of Open Access Journals (Sweden)

    Vojtěch Caha

    2016-12-01

    Full Text Available The knowledge of heat transfer coefficient in the post critical heat flux region in nuclear reactor safety is very important. Although the nuclear reactors normally operate at conditions where critical heat flux (CHF is not reached, accidents where dryout occur are possible. Most serious postulated accidents are a loss of coolant accident or reactivity initiated accident which can lead to CHF or post CHF conditions and possible disruption of core integrity. Moreover, this is also influenced by an oxide layer on the cladding surface. The paper deals with the study of mathematical models and correlations used for heat transfer calculation, especially in post dryout region, and fuel cladding oxidation kinetics of currently operated nuclear reactors. The study is focused on increasing of accuracy and reliability of safety limit calculations (e.g. DNBR or fuel cladding temperature. The paper presents coupled code which was developed for the solution of forced convection flow in heated channel and oxidation of fuel cladding. The code is capable of calculating temperature distribution in the coolant, cladding and fuel and also the thickness of an oxide layer.

  6. Fuel-cladding chemical interaction in mixed-oxide fuels

    International Nuclear Information System (INIS)

    Lawrence, L.A.; Weber, J.W.; Devary, J.L.

    1978-10-01

    The character and extent of fuel-cladding chemical interaction (FCCI) was established for UO 2 -25 wt% PuO 2 clad with 20% cold worked Type 316 stainless steel irradiated at high cladding temperatures to peak burnups greater than 8 atom %. The data base consists of 153 data sets from fuel pins irradiated in EBR-II with peak burnups to 9.5 atom %, local cladding inner surface temperatures to 725 0 C, and exposure times to 415 equivalent full power days. As-fabricated oxygen-to-metal ratios (O/M) ranged from 1.938 to 1.984 with the bulk of the data in the range 1.96 to 1.98. HEDL P-15 pins provided data at low heat rates, approx. 200 W/cm, and P-23 series pins provided data at higher heat rates, approx. 400 W/cm. A design practice for breeder reactors is to consider an initial reduction of 50 microns in cladding thickness to compensate for possible FCCI. This approach was considered to be a conservative approximation in the absence of a comprehensive design correlation for extent of interaction. This work provides to the designer a statistically based correlation for depth of FCCI which reflects the influences of the major fuel and operating parameters on FCCI

  7. Cladding Alloys for Fluoride Salt Compatibility Final Report

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-05-01

    This interim 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 coating 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 coating inaccessible surfaces such as the interior surfaces of heat exchangers. The final project report will feature an experimental evaluation of the performance of the two selected cladding techniques.

  8. The quest for safe and reliable fuel cladding materials

    International Nuclear Information System (INIS)

    Pino, Eddy S.; Abe, Alfredo Y.; Giovedi, Claudia

    2015-01-01

    The tragic Fukushima Daiichi Nuclear Plant accident of March, 2011, has brought great unrest and challenge to the nuclear industry, which, in collaboration with universities and nuclear research institutes, is making great efforts to improve the safety in nuclear reactors developing accident tolerant fuels (ATF). This involves the study of different materials to be applied as cladding and, also, the improvement in the fuel properties in order to enhance the fuel performance and safety, specifically under accident conditions. Related to the cladding, iron based alloys and silicon carbide (SiC) materials have been studied as a good alternative. In the case of austenitic stainless steel, there is the advantage that the austenitic stainless steel 304 was used as cladding material in the first PWR (Pressurized Water Reactor) registering a good performance. Then, alternated cladding materials such as iron based alloys (304, 310, 316, 347) should be used to replace the zirconium-based alloys in order to improve safety. In this paper, these cladding materials are evaluated in terms of their physical and chemical properties; among them, strength and creep resistance, thermal conductivity, thermal stability and corrosion resistance. Additionally, these properties are compared with those of conventional zirconium-based alloys, the most used material in actual PWR, to assess the advantages and disadvantages of each material concerning to fuel performance and safety contribution. (author)

  9. Manufacturing and supply of a mercury cooled rotary uranium target for the GELINA accelerator facility of IRMM

    International Nuclear Information System (INIS)

    Febvre, M.

    2009-01-01

    AREVA CERCA is a manufacturer of Uranium fuel elements for Research Reactors, Mechanical Components for nuclear industry and Radioactive Reference Sources for nuclear industry and medicine. The GELINA high-resolution neutron time-of-flight measurements facility of IRMM (Institute for Reference Materials and Measurements) operates with a very special Uranium target. There is no other neutron production target in the world that has the same or even similar characteristics. Over more than two decades, AREVA CERCA developed and manufactured the rotary Uranium targets for the GELINA facility of IRMM. The use of Uranium and the induced activation during operation in GELINA ask for extremely reliable equipment, responding to very high safety demands. For the fabrication of the target, the very high quality control standards applicable for the fabrication of nuclear equipment must be applied. The material of the Uranium-molybdenum disk must fulfill very restrictive conditions. The target core of the target is cooled by a stream of Mercury, pumped by an electromagnetic pump manufactured by AREVA CERCA. Temperatures in the Uranium disk must be measured with thermocouples distributed inside the Uranium core, at the Uranium/cladding interface and at the inlet and outlet of the Mercury. The target has to rotate between an upper and a lower neutron moderator with very severe geometrical tolerances. The very small gap between the Uranium and the cladding must be filled with Helium before sealing. Preparation and installation of the Target at the GELINA linear Accelerator is presented. Performances and reliability of the targets are reported combined with major experiences and results gained by IRMM. (author)

  10. Measurement of thermal diffusivity of depleted uranium metal microspheres

    Energy Technology Data Exchange (ETDEWEB)

    Humrickhouse-Helmreich, Carissa J., E-mail: carissahelmreich@tamu.edu [Texas A and M University, Department of Nuclear Engineering, 337 Zachry Engineering Center, 3133 TAMU, College Station, TX 77843 (United States); Corbin, Rob, E-mail: rcorbin@terrapower.com [TerraPower, LLC, 330 120th Ave NE, Suite 100, Bellevue, WA 98005 (United States); McDeavitt, Sean M., E-mail: mcdeavitt@tamu.edu [Texas A and M University, Department of Nuclear Engineering, 337 Zachry Engineering Center, 3133 TAMU, College Station, TX 77843 (United States)

    2014-03-15

    The high void space of nuclear fuels composed of homogeneous uranium metal microspheres may allow them to achieve ultra-high burnup by accommodating fuel swelling and reducing fuel/cladding interactions; however, the relatively low thermal conductivity of microsphere nuclear fuels may limit their application. To support the development of microsphere nuclear fuels, an apparatus was designed in a glovebox and used to measure the apparent thermal diffusivity of a packed bed of depleted uranium (DU) microspheres with argon fill in the void spaces. The developed Crucible Heater Test Assembly (CHTA) recorded radial temperature changes due to an initial heat pulse from a central thin-diameter cartridge heater. Using thermocouple positions and time–temperature data, the apparent thermal diffusivity was calculated. The thermal conductivity of the DU microspheres was calculated based on the thermal diffusivity from the CHTA, known material densities and specific heat capacities, and an assumed 70% packing density based on prior measurements. Results indicate that DU metal microspheres have very low thermal conductivity, relative to solid uranium metal, and rapidly form an oxidation layer even in a low oxygen environment. At 500 °C, the thermal conductivity of the DU metal microsphere bed was 0.431 ± 0.0560 W/m-K compared to the literature value of approximately 32 W/m-K for solid uranium metal.

  11. Measurement of thermal diffusivity of depleted uranium metal microspheres

    Science.gov (United States)

    Humrickhouse-Helmreich, Carissa J.; Corbin, Rob; McDeavitt, Sean M.

    2014-03-01

    The high void space of nuclear fuels composed of homogeneous uranium metal microspheres may allow them to achieve ultra-high burnup by accommodating fuel swelling and reducing fuel/cladding interactions; however, the relatively low thermal conductivity of microsphere nuclear fuels may limit their application. To support the development of microsphere nuclear fuels, an apparatus was designed in a glovebox and used to measure the apparent thermal diffusivity of a packed bed of depleted uranium (DU) microspheres with argon fill in the void spaces. The developed Crucible Heater Test Assembly (CHTA) recorded radial temperature changes due to an initial heat pulse from a central thin-diameter cartridge heater. Using thermocouple positions and time-temperature data, the apparent thermal diffusivity was calculated. The thermal conductivity of the DU microspheres was calculated based on the thermal diffusivity from the CHTA, known material densities and specific heat capacities, and an assumed 70% packing density based on prior measurements. Results indicate that DU metal microspheres have very low thermal conductivity, relative to solid uranium metal, and rapidly form an oxidation layer even in a low oxygen environment. At 500 °C, the thermal conductivity of the DU metal microsphere bed was 0.431 ± 0.0560 W/m-K compared to the literature value of approximately 32 W/m-K for solid uranium metal.

  12. Measurement of thermal diffusivity of depleted uranium metal microspheres

    International Nuclear Information System (INIS)

    Humrickhouse-Helmreich, Carissa J.; Corbin, Rob; McDeavitt, Sean M.

    2014-01-01

    The high void space of nuclear fuels composed of homogeneous uranium metal microspheres may allow them to achieve ultra-high burnup by accommodating fuel swelling and reducing fuel/cladding interactions; however, the relatively low thermal conductivity of microsphere nuclear fuels may limit their application. To support the development of microsphere nuclear fuels, an apparatus was designed in a glovebox and used to measure the apparent thermal diffusivity of a packed bed of depleted uranium (DU) microspheres with argon fill in the void spaces. The developed Crucible Heater Test Assembly (CHTA) recorded radial temperature changes due to an initial heat pulse from a central thin-diameter cartridge heater. Using thermocouple positions and time–temperature data, the apparent thermal diffusivity was calculated. The thermal conductivity of the DU microspheres was calculated based on the thermal diffusivity from the CHTA, known material densities and specific heat capacities, and an assumed 70% packing density based on prior measurements. Results indicate that DU metal microspheres have very low thermal conductivity, relative to solid uranium metal, and rapidly form an oxidation layer even in a low oxygen environment. At 500 °C, the thermal conductivity of the DU metal microsphere bed was 0.431 ± 0.0560 W/m-K compared to the literature value of approximately 32 W/m-K for solid uranium metal

  13. Geochemical exploration for uranium

    International Nuclear Information System (INIS)

    1988-01-01

    This Technical Report is designed mainly to introduce the methods and techniques of uranium geochemical exploration to exploration geologists who may not have had experience with geochemical exploration methods in their uranium programmes. The methods presented have been widely used in the uranium exploration industry for more than two decades. The intention has not been to produce an exhaustive, detailed manual, although detailed instructions are given for a field and laboratory data recording scheme and a satisfactory analytical method for the geochemical determination of uranium. Rather, the intention has been to introduce the concepts and methods of uranium exploration geochemistry in sufficient detail to guide the user in their effective use. Readers are advised to consult general references on geochemical exploration to increase their understanding of geochemical techniques for uranium

  14. Uranium purchases report 1992

    International Nuclear Information System (INIS)

    1993-01-01

    Data reported by domestic nuclear utility companies in their responses to the 1991 and 1992 ''Uranium Industry Annual Survey,'' Form EIA-858, Schedule B ''Uranium Marketing Activities,are provided in response to the requirements in the Energy Policy Act 1992. Data on utility uranium purchases and imports are shown on Table 1. Utility enrichment feed deliveries and secondary market acquisitions of uranium equivalent of US DOE separative work units are shown on Table 2. Appendix A contains a listing of firms that sold uranium to US utilities during 1992 under new domestic purchase contracts. Appendix B contains a similar listing of firms that sold uranium to US utilities during 1992 under new import purchase contracts. Appendix C contains an explanation of Form EIA-858 survey methodologies with emphasis on the processing of Schedule B data

  15. Politics of Uranium

    International Nuclear Information System (INIS)

    Anon.

    1982-01-01

    Uranium is the most political of all the elements, the material for the production of both the large amounts of electricity and the most destructive weapons in the world. The problems that its dual potential creates are only now beginning to become evident. Author Norman Moss looks at this situation and sheds light on many of the questions that emerge. The nuclear issue always comes back to how much uranium there is, what can be done with it, and which countries have it. Starting with a concise history of uranium and explaining its technology in terms the nonspecialist can understand, The Politics of Uranium considers the political issues that technical arguments obscure. It tells the little-known story of the international uranium cartel, explains the entanglements of governments with the uranium trade, and describes the consequences of wrong decisions and blunders-especially the problems of nuclear waste. It also examines the intellectual and emotional roots of the anti-nuclear movement

  16. Titrimetric determination of uranium

    International Nuclear Information System (INIS)

    Florence, T.M.

    1989-01-01

    Titrimetric methods are almost invariably used for the high precision assay of uranium compounds, because gravimetric methods are nonselective, and not as reliable. Although precipitation titrations have been used, for example with cupferron and ferrocyanide, and chelate titrations with EDTA and oxine give reasonable results, in practice only redox titrations find routine use. With all redox titration methods for uranium a precision of 01 to 02 percent can be achieved, and precisions as high as 0.003 percent have been claimed for the more refined techniques. There are two types of redox titrations for uranium in common use. The first involves the direct titration of uranium (VI) to uranium (IV) with a standard solution of a strong reductant, such as chromous chloride or titanous chloride, and the second requires a preliminary reduction of uranium to the (IV) or (III) state, followed by titration back to the (VI) state with a standard oxidant. Both types of redox titrations are discussed. 4 figs

  17. Uranium Newsletter. No. 1

    International Nuclear Information System (INIS)

    1987-03-01

    The new Uranium Newsletter is presented as an IAEA annual newsletter. The organization of the IAEA and its involvement with uranium since its founding in 1957 is described. The ''Red Book'' (Uranium Resources, Production and Demand) is mentioned. The Technical Assistance Programme of the IAEA in this field is also briefly mentioned. The contents also include information on the following meetings: The Technical Committee Meeting on Uranium Deposits in Magmatic and Metamorphic Rocks, Advisory Group Meeting on the Use of Airborne Radiometric Data, and the Technical Committee Meeting on Metallogenesis. Recent publications are listed. Current research contracts in uranium exploration are mentioned. IAEA publications on uranium (in press) are listed also. Country reports from the following countries are included: Australia, Brazil, Canada, China (People's Republic of), Denmark, Finland, Germany (Federal Republic of), Malaysia, Philippines, Portugal, South Africa (Republic of), Spain, Syrian Arab Republic, United Kingdom, United States of America, Zambia, and Greece. There is also a report from the Commission of European Communities

  18. Process for continuous production of metallic uranium and uranium alloys

    Science.gov (United States)

    Hayden, Jr., Howard W.; Horton, James A.; Elliott, Guy R. B.

    1995-01-01

    A method is described for forming metallic uranium, or a uranium alloy, from uranium oxide in a manner which substantially eliminates the formation of uranium-containing wastes. A source of uranium dioxide is first provided, for example, by reducing uranium trioxide (UO.sub.3), or any other substantially stable uranium oxide, to form the uranium dioxide (UO.sub.2). This uranium dioxide is then chlorinated to form uranium tetrachloride (UCl.sub.4), and the uranium tetrachloride is then reduced to metallic uranium by reacting the uranium chloride with a metal which will form the chloride of the metal. This last step may be carried out in the presence of another metal capable of forming one or more alloys with metallic uranium to thereby lower the melting point of the reduced uranium product. The metal chloride formed during the uranium tetrachloride reduction step may then be reduced in an electrolysis cell to recover and recycle the metal back to the uranium tetrachloride reduction operation and the chlorine gas back to the uranium dioxide chlorination operation.

  19. Process for continuous production of metallic uranium and uranium alloys

    Science.gov (United States)

    Hayden, H.W. Jr.; Horton, J.A.; Elliott, G.R.B.

    1995-06-06

    A method is described for forming metallic uranium, or a uranium alloy, from uranium oxide in a manner which substantially eliminates the formation of uranium-containing wastes. A source of uranium dioxide is first provided, for example, by reducing uranium trioxide (UO{sub 3}), or any other substantially stable uranium oxide, to form the uranium dioxide (UO{sub 2}). This uranium dioxide is then chlorinated to form uranium tetrachloride (UCl{sub 4}), and the uranium tetrachloride is then reduced to metallic uranium by reacting the uranium chloride with a metal which will form the chloride of the metal. This last step may be carried out in the presence of another metal capable of forming one or more alloys with metallic uranium to thereby lower the melting point of the reduced uranium product. The metal chloride formed during the uranium tetrachloride reduction step may then be reduced in an electrolysis cell to recover and recycle the metal back to the uranium tetrachloride reduction operation and the chlorine gas back to the uranium dioxide chlorination operation. 4 figs.

  20. New french uranium mineral species

    International Nuclear Information System (INIS)

    Branche, G.; Chervet, J.; Guillemin, C.

    1952-01-01

    In this work, the authors study the french new uranium minerals: parsonsite and renardite, hydrated phosphates of lead and uranium; kasolite: silicate hydrated of uranium and lead uranopilite: sulphate of uranium hydrated; bayleyite: carbonate of uranium and of hydrated magnesium; β uranolite: silicate of uranium and of calcium hydrated. For all these minerals, the authors give the crystallographic, optic characters, and the quantitative chemical analyses. On the other hand, the following species, very rare in the french lodgings, didn't permit to do quantitative analyses. These are: the lanthinite: hydrated uranate oxide; the α uranotile: silicate of uranium and of calcium hydrated; the bassetite: uranium phosphate and of hydrated iron; the hosphuranylite: hydrated uranium phosphate; the becquerelite: hydrated uranium oxide; the curite: oxide of uranium and lead hydrated. Finally, the authors present at the end of this survey a primary mineral: the brannerite, complex of uranium titanate. (author) [fr

  1. Uranium industry annual, 1988

    International Nuclear Information System (INIS)

    1989-01-01

    This report presents data on US uranium raw materials and marketing activities of the domestic uranium industry. It contains aggregated data reported by US companies on the ''Uranium Industry Annual Survey'' (1988), Form EIA-858, and historical data from prior data collections and other pertinent sources. The report was prepared by the Energy Information Administration (EIA), the independent agency for data collection and analysis with the US Department of Energy

  2. Pine Creek uranium province

    International Nuclear Information System (INIS)

    Bower, M.B.; Needham, R.S.; Page, R.W.; Stuart-Smith, P.G.; Wyborn, L.A.I.

    1985-01-01

    The objective of this project is to help establish a sound geological framework of the Pine Creek region through regional geological, geochemical and geophysical studies. Uranium ore at the Coronation Hill U-Au mine is confined to a wedge of conglomerate in faulted contact with altered volcanics. The uranium, which is classified as epigenetic sandstone type, is derived from a uranium-enriched felsic volcanic source

  3. Uranium in Canada

    International Nuclear Information System (INIS)

    1985-09-01

    In 1974 the Minister of Energy, Mines and Resources (EMR) established a Uranium Resource Appraisal Group (URAG) within EMR to audit annually Canada's uranium resources for the purpose of implementing the federal government's uranium export policy. A major objective of this policy was to ensure that Canadian uranium supplies would be sufficient to meet the needs of Canada's nuclear power program. As projections of installed nuclear power growth in Canada over the long term have been successively revised downwards (the concern about domestic security of supply is less relevant now than it was 10 years ago) and as Canadian uranium supply capabilities have expanded significantly. Canada has maintained its status as the western world's leading exporter of uranium and has become the world's leading producer. Domestic uranium resource estimates have increased to 551 000 tonnes U recoverable from mineable ore since URAG completed its last formal assessment (1982). In 1984, Canada's five primary uranium producers employed some 5800 people at their mining and milling operations, and produced concentrates containing some 11 170 tU. It is evident from URAG's 1984 assessment that Canada's known uranium resources, recoverable at uranium prices of $150/kg U or less, are more than sufficient to meet the 30-year fuelling requirements of those reactors that are either in opertaion now or committed or expected to be in-service by 1995. A substantial portion of Canada's identified uranium resources, recoverable within the same price range, is thus surplus to Canadian needs and available for export. Sales worth close to $1 billion annually are assured. Uranium exploration expenditures in Canada in 1983 and 1984 were an estimated $41 million and $35 million, respectively, down markedly from the $128 million reported for 1980. Exploration drilling and surface development drilling in 1983 and 1984 were reported to be 153 000 m and 197 000 m, respectively, some 85% of which was in

  4. Gold and uranium extraction

    International Nuclear Information System (INIS)

    James, G.S.; Davidson, R.J.

    1977-01-01

    A process for extracting gold and uranium from an ore containing them both comprising the steps of pulping the finely comminuted ore with a suitable cyanide solution at an alkaline pH, acidifying the pulp for uranium dissolution, adding carbon activated for gold recovery to the pulp at a suitable stage, separating the loaded activated carbon from the pulp, and recovering gold from the activated carbon and uranium from solution

  5. Performance of HT9 clad metallic fuel at high temperature

    International Nuclear Information System (INIS)

    Pahl, R.G.; Lahm, C.E.; Hayes, S.L.

    1992-01-01

    Steady-state testing of HT9 clad metallic fuel at high temperatures was initiated in EBR-II in November of 1987. At that time U-10 wt. % Zr fuel clad with the low-swelling ferritic/martensitic alloy HT9 was being considered as driver fuel options for both EBR-II and FFTF. The objective of the X447 test described here was to determine the lifetime of HT9 cladding when operated with metallic fuel at beginning of life inside wall temperatures approaching ∼660 degree C. Though stress-temperature design limits for HT9 preclude its use for high burnup applications under these conditions due to excessive thermal creep, the X447 test was carried out to obtain data on high temperature breach phenomena involving metallic fuel since little data existed in that area

  6. Manufacturing process for the metal ceramic hybrid fuel cladding tube

    International Nuclear Information System (INIS)

    Jung, Yang Il; Kim, Sun Han; Park, Jeong Yong

    2012-01-01

    For application in LWRs with suppressed hydrogen release, a metal-ceramic hybrid cladding tube has been proposed. The cladding consists of an inner zirconium tube and outer SiC fiber matrix SiC ceramic composite. The inner zirconium allows the matrix to remain fully sealed even if the ceramic matrix cracks through. The outer SiC composite can increase the safety margin by taking the merits of the SiC itself. However, it is a challenging task to fabricate the metal-ceramic hybrid tube. Processes such as filament winding, matrix impregnation, and surface costing are additionally required for the existing Zr based fuel cladding tubes. In the current paper, the development of the manufacturing process will be introduced

  7. Compatibility of niobium, titanium, and vanadium metals with LMFBR cladding

    International Nuclear Information System (INIS)

    Wilson, C.N.

    1975-10-01

    A series of laboratory capsule annealing experiments were conducted to assess the compatibility of niobium, vanadium, and titanium with 316 stainless steel cladding in the temperature range of 700 to 800 0 C. Niobium, vanadium, and titanium are cantidate oxygen absorber materials for control of oxygen chemistry in LMFBR fuel pins. Capsule examination indicated good compatibility between niobium and 316 stainless steel at 800 0 C. Potential compatibility problems between cladding and vanadium or titanium were indicated at 800 0 C under reducing conditions. In the presence of Pu/sub 0.25/U/sub 0.75/O/sub 1.98/ fuel (Δanti G 02 congruent to -160 kcal/mole) no reaction was observed between vanadium or titanium and cladding at 800 0 C

  8. Chemical Dissolution of Simulant FCA Cladding and Plates

    Energy Technology Data Exchange (ETDEWEB)

    Daniel, G. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Pierce, R. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); O' Rourke, P. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-11-08

    The Savannah River Site (SRS) has received some fast critical assembly (FCA) fuel from the Japan Atomic Energy Agency (JAEA) for disposition. Among the JAEA FCA fuel are approximately 7090 rectangular Stainless Steel clad fuel elements. Each element has an internal Pu-10.6Al alloy metal wafer. The thickness of each element is either 1/16 inch or 1/32 inch. The dimensions of each element ranges from 2 inches x 1 inch to 2 inches x 4 inches. This report discusses the potential chemical dissolution of the FCA clad material or stainless steel. This technology uses nitric acid-potassium fluoride (HNO3-KF) flowsheets of H-Canyon to dissolve the FCA elements from a rack of materials. Historically, dissolution flowsheets have aimed to maximize Pu dissolution rates while minimizing stainless steel dissolution (corrosion) rates. Because the FCA cladding is made of stainless steel, this work sought to accelerate stainless steel dissolution.

  9. Characterization Of Cladding Hull Wastes From Used Nuclear Fuels

    Directory of Open Access Journals (Sweden)

    Kang K.H.

    2015-06-01

    Full Text Available Used cladding hulls from pressurized water reactor (PWR are characterized to provide useful information for the treatment and disposal of cladding hull wastes. The radioactivity and the mass of gamma emitting nuclides increases with an increase in the fuel burn-up and their removal ratios are found to be more than 99 wt.% except Co-60 and Cs-137. In the result of measuring the concentrations of U and Pu included in the cladding hull wastes, most of the residues are remained on the surface and the removal ratio of U and Pu are revealed to be over 99.98 wt.% for the fuel burn-up of 35,000 MWd/tU. An electron probe micro-analyzer (EPMA line scanning shows that radioactive fission products are penetrated into the Zr oxide layer, which is proportional to the fuel burn-up. The oxidative decladding process exhibits more efficient removal ratio of radionuclides.

  10. Clad photon sieve for generating localized hollow beams

    Science.gov (United States)

    Cheng, Yiguang; Tong, Junmin; Zhu, Jiangping; Liu, Junbo; Hu, Song; He, Yu

    2016-02-01

    A novel photon sieve structure called clad photon sieve is proposed to generate localized hollow beams and its design principle and focusing properties are studied. The clad photon sieve is composed of the internal zone and external zone with pinholes being positioned on the dark zones. Pinholes in the internal zone and in the external zone give destructive interference to the focus, leading to localized hollow beams being generated on the focal plane. Focusing properties of clad photon sieve with different focal lengths, zone numbers and modulation factors are also studied by theoretical calculations, numerical simulations and experiments, showing that the central dark spot size can be controlled by the focal length and rings number, and the intensity of the central dark spot varies with different modulation factors related with the internal zone and the external zone. This photon sieve can be useful for trapping and manipulating of particles and cooling of atoms.

  11. Compatibility of niobium, titanium, and vanadium metals with LMFBR cladding

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, C.N.

    1975-10-01

    A series of laboratory capsule annealing experiments were conducted to assess the compatibility of niobium, vanadium, and titanium with 316 stainless steel cladding in the temperature range of 700 to 800/sup 0/C. Niobium, vanadium, and titanium are cantidate oxygen absorber materials for control of oxygen chemistry in LMFBR fuel pins. Capsule examination indicated good compatibility between niobium and 316 stainless steel at 800/sup 0/C. Potential compatibility problems between cladding and vanadium or titanium were indicated at 800/sup 0/C under reducing conditions. In the presence of Pu/sub 0.25/U/sub 0.75/O/sub 1.98/ fuel (..delta..anti G/sub 02/ congruent to -160 kcal/mole) no reaction was observed between vanadium or titanium and cladding at 800/sup 0/C.

  12. Absorptivity Measurements and Heat Source Modeling to Simulate Laser Cladding

    Science.gov (United States)

    Wirth, Florian; Eisenbarth, Daniel; Wegener, Konrad

    The laser cladding process gains importance, as it does not only allow the application of surface coatings, but also additive manufacturing of three-dimensional parts. In both cases, process simulation can contribute to process optimization. Heat source modeling is one of the main issues for an accurate model and simulation of the laser cladding process. While the laser beam intensity distribution is readily known, the other two main effects on the process' heat input are non-trivial. Namely the measurement of the absorptivity of the applied materials as well as the powder attenuation. Therefore, calorimetry measurements were carried out. The measurement method and the measurement results for laser cladding of Stellite 6 on structural steel S 235 and for the processing of Inconel 625 are presented both using a CO2 laser as well as a high power diode laser (HPDL). Additionally, a heat source model is deduced.

  13. International trade in uranium

    International Nuclear Information System (INIS)

    Two reports are presented; one has been prepared by the Uranium Institute and is submitted by the United Kingdom delegation, the other by the United States delegation. The report of the Uranium Institute deals with the influence of the government on international trade in uranium. This influence becomes apparent predominantly by export and import restrictions, as well as by price controls. The contribution submitted by the United States is a uranium market trend analysis, with pricing methods and contracting modes as well as the effect of government policies being investigated in the light of recent developments

  14. Uranium concentration in fossils

    International Nuclear Information System (INIS)

    Okano, J.; Uyeda, C.

    1988-01-01

    Recently it is known that fossil bones tend to accumulate uranium. The uranium concentration, C u in fossils has been measured so far by γ ray spectroscopy or by fission track method. The authors applied secondary ion mass spectrometry, SIMS, to detect the uranium in fossil samples. The purpose of this work is to investigate the possibility of semi-quantitative analyses of uranium in fossils, and to study the correlation between C u and the age of fossil bones. The further purpose of this work is to apply SIMS to measure the distribution of C u in fossil teeth

  15. analysis methods of uranium

    International Nuclear Information System (INIS)

    Bekdemir, N.; Acarkan, S.

    1997-01-01

    There are various methods for the determination of uranium. The most often used methods are spectrophotometric (PAR, DBM and Arsenazo III) and potentiometric titration methods. For uranium contents between 1-300 g/LU potentiometric titration method based on oxidation-reduction reactions gives reliable results. PAR (1-pyridiyl-2-azo resorcinol) is a sensitive reagent for uranium, forming complexes in aqueous solutions. It is a suitable method for determination of uranium at concentrations between 2-400microgram U. In this study, the spectrophotometric and potentiometric analysis methods, used in the Nuclear Fuel Department will be discussed in detail and other methods and their principles will be briefly mentioned

  16. Uranium production from phosphates

    International Nuclear Information System (INIS)

    Ketzinel, Z.; Folkman, Y.

    1979-05-01

    According to estimates of the world's uranium consumption, exploitation of most rich sources is expected by the 1980's. Forecasts show that the rate of uranium consumption will increase towards the end of the century. It is therefore desirable to exploit poor sources not yet in use. In the near future, the most reasonable source for developing uranium is phosphate rock. Uranium reserves in phosphates are estimated at a few million tons. Production of uranium from phosphates is as a by-product of phosphate rock processing and phosphoric acid production; it will then be possible to save the costs incurred in crushing and dissolving the rock when calculating uranium production costs. Estimates show that the U.S. wastes about 3,000 tons of uranium per annum in phosphoric acid based fertilisers. Studies have also been carried out in France, Yugoslavia and India. In Israel, during the 1950's, a small plant was operated in Haifa by 'Chemical and Phosphates'. Uranium processes have also been developed by linking with the extraction processes at Arad. Currently there is almost no activity on this subject because there are no large phosphoric acid plants which would enable production to take place on a reasonable scale. Discussions are taking place about the installation of a plant for phosphoric acid production utilising the 'wet process', producing 200 to 250,000 tons P 2 O 5 per annum. It is necessary to combine these facilities with uranium production plant. (author)

  17. METHOD OF ROLLING URANIUM

    Science.gov (United States)

    Smith, C.S.

    1959-08-01

    A method is described for rolling uranium metal at relatively low temperatures and under non-oxidizing conditions. The method involves the steps of heating the uranium to 200 deg C in an oil bath, withdrawing the uranium and permitting the oil to drain so that only a thin protective coating remains and rolling the oil coated uranium at a temperature of 200 deg C to give about a 15% reduction in thickness at each pass. The operation may be repeated to accomplish about a 90% reduction without edge cracking, checking or any appreciable increase in brittleness.

  18. Influence Of The Laser Cladding Strategies On The Mechanical Properties Of Inconel 718

    International Nuclear Information System (INIS)

    Lamikiz, A.; Tabernero, I.; Ukar, E.; Lopez de Lacalle, L. N.; Delgado, J.

    2011-01-01

    This work presents different experimental results of the mechanical properties of Inconel registered 718 test parts built-up by laser cladding. Recently, turbine manufacturers for aeronautical sector have presented high interest on laser cladding processes. This process allows building fully functional structures on superalloys, such as Inconel registered 718, with high flexibility on complex shapes. However, there is limited data on mechanical properties of the laser cladding structures. Moreover, the available data do not include the influence of process parameters and laser cladding strategies. Therefore, a complete study of the influence of the laser cladding parameters and mainly, the variation of the tensile strength with the laser cladding strategy is presented. The results show that there is a high directionality of mechanical properties, depending on the strategies of laser cladding process. In other words, the test parts show a fiber -like structure that should be considered on the laser cladding strategy selection.

  19. Low-Stress Silicon Cladding for Surface Finishing Large UVOIR Mirrors Project

    Data.gov (United States)

    National Aeronautics and Space Administration — In this Phase I research, ZeCoat Corporation demonstrated a low-stress silicon cladding process for surface finishing large UVOIR mirrors. A polishable cladding is...

  20. Gap conductance in Zircaloy-clad LWR fuel rods

    International Nuclear Information System (INIS)

    Ainscough, J.B.

    1982-04-01

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

  1. Comparison of models discribing cladding deformations during LOCA

    International Nuclear Information System (INIS)

    Chakraborty, A.K.; Zipper, R.

    1981-05-01

    This report compares the important models for the determination of cladding deformations during LOCA. In addition to the comparisons of underlying assumptions of different models the same is done for the coefficients applied for the models. In order to assess the predictive capability of the models the calculated results are compared with the experimental results of the individual claddings. It was found out that the results of temperature ramp tests could be calculated better than that of the pressure ramp tests. The calculations revealed that even with the simplified assumption of the model used in TESPA the agreement of the calculated results with those of model NORA was relatively good. (orig.) [de

  2. Cladding dimensional changes in mixed oxide fuel pins

    International Nuclear Information System (INIS)

    Makenas, B.J.; Jost, J.W.; Hales, J.W.

    1980-01-01

    Several types of stainless steel (304, 316, 316-20% Cold Worked, and 316 Titanium stabilized 20% CW) have been used as cladding for mixed oxide (U, Pu)O 2 fuel pins irradiated in EBR-II. All of the materials have performed satisfactorily in their respective experimental subassemblies but significant differences in swelling and inelastic strain behavior have been found at high fast fluences among the different materials and among different heats of the same material. Cladding diameter increases were measured for 622 developmental and FFTF reference design fuel pins which were irradiated in 19 experimental subassemblies. Fuel pin diameters were determined from multiangle axial trace profilometry measurements

  3. Technical committee meeting on fuel and cladding interaction. Summary report

    International Nuclear Information System (INIS)

    1977-04-01

    Experiments and experiences concerning fuel-cladding interaction in thermal and fast neutron flux burnup are dealt with. A number of results from in-pile and out-of pile experiments with different fuel pins with cladding made of different stainless steels showed the importance of corrosion process, dependent on the burnup, core temperature, metal-oxide ratio, and other steady state parameters in the core of fast reactors (most frequently LMFBRs). This is of importance for fuel pins design and fabrication. Mixed oxide fuel is treated in many cases

  4. Synthesis of clad motion experiments interpretation: codes and validation

    International Nuclear Information System (INIS)

    Papin, J.; Fortunato, M.; Seiler, J.M.

    1983-04-01

    This communication deals with clad melting and relocation phenomena related to LMFBR safety analysis of loss of flow accidents. We present: - the physical models developed at DSN/CEN Cadarache in single channel and bundle geometry. The interpretation with these models of experiments performed by the STT (CEN Grenoble). It comes out that we have now obtained a good understanding of the involved phenomena in single channel geometry. On the other hand, further studies are necessary for a better knowledge of clad motion phenomena in bundle cases with conditions close to reactor ones

  5. Irradiation experience with HT9-clad metallic fuel

    International Nuclear Information System (INIS)

    Pahl, R.G.; Lahm, C.E.; Tsai, H.; Billone, M.C.

    1991-01-01

    The safe and reliable performance of metallic fuel is currently under study and demonstration in the Integral Fast Reactor program. In-reactor tests of HT9-clad metallic fuel have now reached maturity and have all shown good performance characteristics to burnups exceeding 17.5 at. % in the lead assembly. Because this low-swelling tempered martensitic alloy is the cladding of choice for high fluence applications, the experimental observations and performance modelling efforts reported in this paper play an important role in demonstrating reliability

  6. Modelling of pellet-cladding interaction in PWR's

    International Nuclear Information System (INIS)

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

    1992-01-01

    The pellet-cladding interaction that can occur in a PWR fuel rod design is modelled with the computer codes FRAPCON-1 and ANSYS. The fuel performance code FRAPCON-1 analyses the fuel rod irradiation behavior and generates the initial conditions for the localized fuel rod thermal and mechanical modelling in two and three-dimensional finite elements with ANSYS. In the mechanical modelling, a pellet fragment is placed in the fuel rod gap. Two types of fuel rod cladding materials are considered: Zircaloy and austenitic stainless steel. (author)

  7. Engineering of metal-clad optical nanocavity to optimize coupling with integrated waveguides

    OpenAIRE

    Kim, Myung-Ki; Li, Zheng; Huang, Kun; Going, Ryan; Wu, Ming C.; Choo, Hyuck

    2013-01-01

    We propose a cladding engineering method that flexibly modifies the radiation patterns and rates of metal-clad nanoscale optical cavity. Optimally adjusting the cladding symmetry of the metal-clad nanoscale optical cavity modifies the modal symmetry and produces highly directional radiation that leads to 90% coupling efficiency into an integrated waveguide. In addition, the radiation rate of the cavity mode can be matched to its absorption rate by adjusting the thickness of the bottom-claddin...

  8. Report on Reactor Physics Assessment of Candidate Accident Tolerant Fuel Cladding Materials in LWRs

    Energy Technology Data Exchange (ETDEWEB)

    Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); George, Nathan [Univ. of Tennessee, Knoxville, TN (United States); Maldonado, G. Ivan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Worrall, Andrew [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-08-28

    This work focuses on ATF concepts being researched at Oak Ridge National Laboratory (ORNL), expanding on previous studies of using alternate cladding materials in pressurized water reactors (PWRs). The neutronic performance of two leading alternate cladding materials were assessed in boiling water reactors (BWRs): iron-chromium-aluminum (FeCrAl) cladding, and silicon carbide (SiC)-based composite cladding. This report fulfills ORNL Milestone M3FT-15OR0202332 within the fiscal year 2015 (FY15)

  9. Ceramic Coatings for Clad (The C3 Project): Advanced Accident-Tolerant Ceramic Coatings for Zr-Alloy Cladding

    Energy Technology Data Exchange (ETDEWEB)

    Sickafus, Kurt E. [Univ. of Tennessee, Knoxville, TN (United States); Wirth, Brian [Univ. of Tennessee, Knoxville, TN (United States); Miller, Larry [Univ. of Tennessee, Knoxville, TN (United States); Weber, Bill [Univ. of Tennessee, Knoxville, TN (United States); Zhang, Yanwen [Univ. of Tennessee, Knoxville, TN (United States); Patel, Maulik [Univ. of Tennessee, Knoxville, TN (United States); Motta, Arthur [Pennsylvania State Univ., University Park, PA (United States); Wolfe, Doug [Pennsylvania State Univ., University Park, PA (United States); Fratoni, Max [Univ. of California, Berkeley, CA (United States); Raj, Rishi [Univ. of Colorado, Boulder, CO (United States); Plunkett, Kenneth [Univ. of Colorado, Boulder, CO (United States); Was, Gary [Univ. of Michigan, Ann Arbor, MI (United States); Hollis, Kendall [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Nelson, Andy [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Stanek, Chris [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Comstock, Robert [Westinghouse Electric Corporation, Pittsburgh, PA (United States); Partezana, Jonna [Westinghouse Electric Corporation, Pittsburgh, PA (United States); Whittle, Karl [Univ. of Sheffield (United Kingdom); Preuss, Michael [Univ. of Manchester (United Kingdom); Withers, Philip [Univ. of Manchester (United Kingdom); Wilkinson, Angus [Univ. of Oxford (United Kingdom); Donnelly, Stephen [Univ. of Huddersfield (United Kingdom); Riley, Daniel [Australian Nuclear Science and Technology Organisation, Syndney (Australia)

    2017-02-14

    The goal of this NEUP-IRP project is to develop a fuel concept based on an advanced ceramic coating for Zr-alloy cladding. The coated cladding must exhibit demonstrably improved performance compared to conventional Zr-alloy clad in the following respects: During normal service, the ceramic coating should decrease cladding oxidation and hydrogen pickup (the latter leads to hydriding and embrittlement). During a reactor transient (e.g., a loss of coolant accident), the ceramic coating must minimize or at least significantly delay oxidation of the Zr-alloy cladding, thus reducing the amount of hydrogen generated and the oxygen ingress into the cladding. The specific objectives of this project are as follows: To produce durable ceramic coatings on Zr-alloy clad using two possible routes: (i) MAX phase ceramic coatings or similar nitride or carbide coatings; and (ii) graded interface architecture (multilayer) ceramic coatings, using, for instance, an oxide such as yttria-stabilized zirconia (YSZ) as the outer protective layer. To characterize the structural and physical properties of the coated clad samples produced in 1. above, especially the corrosion properties under simulated normal and transient reactor operating conditions. To perform computational analyses to assess the effects of such coatings on fuel performance and reactor neutronics, and to perform fuel cycle analyses to assess the economic viability of modifying conventional Zr-alloy cladding with ceramic coatings. This project meets a number of the goals outlined in the NEUP-IRP call for proposals, including: Improve the fuel/cladding system through innovative designs (e.g. coatings/liners for zirconium-based cladding) Reduce or eliminate hydrogen generation Increase resistance to bulk steam oxidation Achievement of our goals and objectives, as defined above, will lead to safer light-water reactor (LWR) nuclear fuel assemblies, due to improved cladding properties and built-in accident resistance, as well as

  10. Trends in uranium supply

    International Nuclear Information System (INIS)

    Hansen, M.

    1976-01-01

    Prior to the development of nuclear power, uranium ores were used to a very limited extent as a ceramic colouring agent, as a source of radium and in some places as a source of vanadium. Perhaps before that, because of the bright orange and yellow colours of its secondary ores, it was probably used as ceremonial paint by primitive man. After the discovery of nuclear fission a whole new industry emerged, complete with its problems of demand, resources and supply. Spurred by special incentives in the early years of this new nuclear industry, prospectors discovered over 20 000 occurrences of uranium in North America alone, and by 1959 total world production reached a peak of 34 000 tonnes uranium from mines in South Africa, Canada and United States. This rapid growth also led to new problems. As purchases for military purposes ended, government procurement contracts were not renewed, and the large reserves developed as a result of government purchase incentives, in combination with lack of substantial commercial market, resulted in an over-supply of uranium. Typically, an over-supply of uranium together with national stockpiling at low prices resulted in depression of prices to less than $5 per pound by 1971. Although forecasts made in the early 1970's increased confidence in the future of nuclear power, and consequently the demand for uranium, prices remained low until the end of 1973 when OPEC announced a very large increase in oil prices and quite naturally, prices for coal also rose substantially. The economics of nuclear fuel immediately improved and prices for uranium began to climb in 1974. But the world-wide impact of the OPEC decision also produced negative effects on the uranium industry. Uranium production costs rose dramatically, as did capital costs, and money for investment in new uranium ventures became more scarce and more expensive. However, the uranium supply picture today offers hope of satisfactory development in spite of the many problems to be

  11. Uranium industry annual 1993

    International Nuclear Information System (INIS)

    1994-09-01

    Uranium production in the United States has declined dramatically from a peak of 43.7 million pounds U 3 O 8 (16.8 thousand metric tons uranium (U)) in 1980 to 3.1 million pounds U 3 O 8 (1.2 thousand metric tons U) in 1993. This decline is attributed to the world uranium market experiencing oversupply and intense competition. Large inventories of uranium accumulated when optimistic forecasts for growth in nuclear power generation were not realized. The other factor which is affecting U.S. uranium production is that some other countries, notably Australia and Canada, possess higher quality uranium reserves that can be mined at lower costs than those of the United States. Realizing its competitive advantage, Canada was the world's largest producer in 1993 with an output of 23.9 million pounds U 3 O 8 (9.2 thousand metric tons U). The U.S. uranium industry, responding to over a decade of declining market prices, has downsized and adopted less costly and more efficient production methods. The main result has been a suspension of production from conventional mines and mills. Since mid-1992, only nonconventional production facilities, chiefly in situ leach (ISL) mining and byproduct recovery, have operated in the United States. In contrast, nonconventional sources provided only 13 percent of the uranium produced in 1980. ISL mining has developed into the most cost efficient and environmentally acceptable method for producing uranium in the United States. The process, also known as solution mining, differs from conventional mining in that solutions are used to recover uranium from the ground without excavating the ore and generating associated solid waste. This article describes the current ISL Yang technology and its regulatory approval process, and provides an analysis of the factors favoring ISL mining over conventional methods in a declining uranium market

  12. Uranium industry annual 1993

    Energy Technology Data Exchange (ETDEWEB)

    1994-09-01

    Uranium production in the United States has declined dramatically from a peak of 43.7 million pounds U{sub 3}O{sub 8} (16.8 thousand metric tons uranium (U)) in 1980 to 3.1 million pounds U{sub 3}O{sub 8} (1.2 thousand metric tons U) in 1993. This decline is attributed to the world uranium market experiencing oversupply and intense competition. Large inventories of uranium accumulated when optimistic forecasts for growth in nuclear power generation were not realized. The other factor which is affecting U.S. uranium production is that some other countries, notably Australia and Canada, possess higher quality uranium reserves that can be mined at lower costs than those of the United States. Realizing its competitive advantage, Canada was the world`s largest producer in 1993 with an output of 23.9 million pounds U{sub 3}O{sub 8} (9.2 thousand metric tons U). The U.S. uranium industry, responding to over a decade of declining market prices, has downsized and adopted less costly and more efficient production methods. The main result has been a suspension of production from conventional mines and mills. Since mid-1992, only nonconventional production facilities, chiefly in situ leach (ISL) mining and byproduct recovery, have operated in the United States. In contrast, nonconventional sources provided only 13 percent of the uranium produced in 1980. ISL mining has developed into the most cost efficient and environmentally acceptable method for producing uranium in the United States. The process, also known as solution mining, differs from conventional mining in that solutions are used to recover uranium from the ground without excavating the ore and generating associated solid waste. This article describes the current ISL Yang technology and its regulatory approval process, and provides an analysis of the factors favoring ISL mining over conventional methods in a declining uranium market.

  13. Analysis of pellet cladding interaction and creep of U 3SIi2 fuel for use in light water reactors

    Science.gov (United States)

    Metzger, Kathryn E.

    Following the accident at the Fukushima plant, enhancing the accident tolerance of the light water reactor (LWR) fleet became a topic of serious discussion. Under the direction of congress, the DOE office of Nuclear Energy added accident tolerant fuel development as a primary component to the existing Advanced Fuels Program. The DOE defines accident tolerant fuels as fuels that "in comparison with the standard UO2- Zircaloy system currently used by the nuclear industry, can tolerate loss of active cooling in the reactor core for a considerably longer time period (depending on the LWR system and accident scenario) while maintaining or improving the fuel performance during normal operations, operational transients, as well as design-basis and beyond design-basis events." To be economically viable, proposed accident tolerant fuels and claddings should be backward compatible with LWR designs, provide significant operating cost improvements such as power uprates, increased fuel burnup, or increased cycle length. In terms of safety, an alternative fuel pellet must have resistance to water corrosion comparable to UO2, thermal conductivity equal to or larger than that of UO2, and a melting temperature that allows the material to remain solid under power reactor conditions. Among the candidates, U3Si2 has a number of advantageous thermophysical properties, including; high density, high thermal conductivity at room temperature, and a high melting temperature. These properties support its use as an accident tolerant fuel while its high uranium density is capable of supporting uprates to the LWR fleet. This research characterizes U3Si2 pellets and analyzes U3Si2 under light water reactor conditions using the fuel performance code BISON. While some thermophysical properties for U3Si2 have been found in the literature, the irradiation behavior is sparse and limited to experience with dispersion fuels. Accordingly, the creep behavior for U3Si2 has been unknown, making it

  14. Gradient microstructure in laser clad TiC-reinforced Ni-alloy composite coating

    NARCIS (Netherlands)

    Pei, Y.T.; Zuo, T.C.

    1998-01-01

    A gradient TiC–(Ni alloy) composite coating was produced by one step laser cladding with pre-placed mixture powder on a 1045 steel substrate. The clad layers consisted of TiC particles, γ-Ni primary dendrites and interdendritic eutectics. From the bottom to the top of the clad layer produced at 2000

  15. Uranium geochemistry, mineralogy, geology, exploration and resources

    International Nuclear Information System (INIS)

    De Vivo, B.

    1984-01-01

    This book comprises papers on the following topics: history of radioactivity; uranium in mantle processes; transport and deposition of uranium in hydrothermal systems at temperatures up to 300 0 C: Geological implications; geochemical behaviour of uranium in the supergene environment; uranium exploration techniques; uranium mineralogy; time, crustal evolution and generation of uranium deposits; uranium exploration; geochemistry of uranium in the hydrographic network; uranium deposits of the world, excluding Europe; uranium deposits in Europe; uranium in the economics of energy; role of high heat production granites in uranium province formation; and uranium deposits

  16. Aspects and application of etch-pit technique in the distribution detection of uranium in solids

    International Nuclear Information System (INIS)

    Sachett, I.A.

    1973-01-01

    Some aspects related to the use of standard glass microscope slide as solid state track detector are presented, as well as studies of measurement of distribution of uranium in various solids determined by means of this technique. The behavior of the various parameters related to the properties of the detector and the results of their measurements are treated. Results obtained on the determination of distribution of uranium in samples of fuel elements of the 'ARGONAUTA' research reactor of Instituto de Engenharia Nuclear - Rio de Janeiro, are included. The samples were taken from the fuel-cladding interface. Finally, some data on levels of contamination of the laboratory equipment in use by uranium during the course of this work are presented [pt

  17. Australia and uranium

    International Nuclear Information System (INIS)

    Anon.

    1977-01-01

    A brief justification of the Australian Government's decision to mine and export Australian Uranium is presented along with a description of the Alligator River Region in the Northern Territory where the major mines are to be located. Aboriginal interests and welfare in the region, the proposed Kakadu National Park and the economic benefits resulting from uranium development are also briefly covered. (J.R.)

  18. Preparation of uranium tetrafluoride

    International Nuclear Information System (INIS)

    Wirths, G.

    1981-01-01

    Uranium dioxide is converted to uranium tetrafluoride under stoichiometric excess of hydrogen fluoride. The water formed in the process and the unreacted hydrogen fluoride are cooled and the condensate fractionally distilled into water and approx. 40% hydrofluoric acid. The hydrofluoric acid and water-free hydrogen fluoride are fed back into the process. (WI) [de

  19. Uranium: biokinetics and toxicity

    International Nuclear Information System (INIS)

    Menetrier, F.; Renaud-Salis, V.; Flury-Herard, A.

    2000-01-01

    This report was achieved as a part of a collaboration with the Fuel Cycle Direction. Its aim was to give the state of the art about: the behaviour of uranium in the human organism (biokinetics) after ingestion, its toxicity (mainly renal) and the current regulation about its incorporation. Both in the upstream and in the downstream of the fuel cycle, uranium remains, quantitatively, the first element in the cycle which is, at the present time, temporarily disposed or recycled. Such a considerable quantity of uranium sets the problem of its risk on the health. In the long term, the biosphere may be affected and consequently the public may ingest water or food contaminated with uranium. In this way, radiological and chemical toxicity risk may be activated. This report emphasizes: the necessity of confirming some experimental and epidemiological biokinetic data used or not in the ICRP models. Unsolved questions remain about the gastrointestinal absorption according to chemical form (valency state, mixtures...), mass and individual variations (age, disease) further a chronic ingestion of uranium. It is well established that uranium is mainly deposited in the skeleton and the kidney. But the skeleton kinetics following a chronic ingestion and especially in some diseases has to be more elucidated; the necessity of taking into account uranium at first as a chemical toxic, essentially in the kidney and determining the threshold of functional lesion. In this way, it is important to look for some specific markers; the problem of not considering chemical toxicity of uranium in the texts regulating its incorporation

  20. Ranger uranium project

    International Nuclear Information System (INIS)

    1979-01-01

    Details are given of an agreement between Peko-Wallsend Operations Ltd., Electrolytic Zinc Company of Australasia Limited, the Australian Atomic Energy Commission and Ranger Uranium Mines Pty. Ltd. regarding the management of the joint Ranger uranium project in the Northern Territory of Australia

  1. Rheinbraun's Australian uranium business

    International Nuclear Information System (INIS)

    Kirschbaum, S.

    1989-01-01

    The leaflet argues against the mining activities of the Rheinische Braunkohlenwerke AG in Germany and especially against uranium mining in Australia. The ethno-ecological impact on flora and fauna, aborigines and miners are pointed out. Uranium mining and lignite mining are compared. (HSCH) [de

  2. Nuclear and uranium policies

    International Nuclear Information System (INIS)

    MacNabb, G.M.; Uranium Canada Ltd., Ottawa, Ontario)

    The background of the uranium industry in Canada is described. Government policies with respect to ownership of the uranium mining industry, price stabilization, and especially reservation of sufficient supplies of nuclear fuels for domestic utilities, are explained. Canadian policy re nuclear exports and safeguards is outlined. (E.C.B.)

  3. Uranium dioxide pellets

    International Nuclear Information System (INIS)

    Zawidzki, T.W.

    1979-01-01

    Sintered uranium dioxide pellets composed of particles of size > 50 microns suitable for power reactor use are made by incorporating a small amount of sulphur into the uranium dioxide before sintering. The increase in grain size achieved results in an improvement in overall efficiency when such pellets are used in a power reactor. (author)

  4. Uranium: A Dentist's perspective

    Science.gov (United States)

    Toor, R. S. S.; Brar, G. S.

    2012-01-01

    Uranium is a naturally occurring radionuclide found in granite and other mineral deposits. In its natural state, it consists of three isotopes (U-234, U-235 and U-238). On an average, 1% – 2% of ingested uranium is absorbed in the gastrointestinal tract in adults. The absorbed uranium rapidly enters the bloodstream and forms a diffusible ionic uranyl hydrogen carbonate complex (UO2HCO3+) which is in equilibrium with a nondiffusible uranyl albumin complex. In the skeleton, the uranyl ion replaces calcium in the hydroxyapatite complex of the bone crystal. Although in North India, there is a risk of radiological toxicity from orally ingested natural uranium, the principal health effects are chemical toxicity. The skeleton and kidney are the primary sites of uranium accumulation. Acute high dose of uranyl nitrate delays tooth eruption, and mandibular growth and development, probably due to its effect on target cells. Based on all previous research and recommendations, the role of a dentist is to educate the masses about the adverse effects of uranium on the overall as well as the dental health. The authors recommended that apart from the discontinuation of the addition of uranium to porcelain, the Public community water supplies must also comply with the Environmental Protection Agency (EPA) standards of uranium levels being not more than 30 ppb (parts per billion). PMID:24478959

  5. Uranium and transuranium analysis

    International Nuclear Information System (INIS)

    Regnaud, F.

    1989-01-01

    Analytical chemistry of uranium, neptunium, plutonium, americium and curium is reviewed. Uranium and neptunium are mainly treated and curium is only briefly evoked. Analysis methods include coulometry, titration, mass spectrometry, absorption spectrometry, spectrofluorometry, X-ray spectrometry, nuclear methods and radiation spectrometry [fr

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

    International Nuclear Information System (INIS)

    Venter, S.

    1978-05-01

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

  7. Uranium dioxide electrolysis

    Science.gov (United States)

    Willit, James L [Batavia, IL; Ackerman, John P [Prescott, AZ; Williamson, Mark A [Naperville, IL

    2009-12-29

    This is a single stage process for treating spent nuclear fuel from light water reactors. The spent nuclear fuel, uranium oxide, UO.sub.2, is added to a solution of UCl.sub.4 dissolved in molten LiCl. A carbon anode and a metallic cathode is positioned in the molten salt bath. A power source is connected to the electrodes and a voltage greater than or equal to 1.3 volts is applied to the bath. At the anode, the carbon is oxidized to form carbon dioxide and uranium chloride. At the cathode, uranium is electroplated. The uranium chloride at the cathode reacts with more uranium oxide to continue the reaction. The process may also be used with other transuranic oxides and rare earth metal oxides.

  8. EPR of uranium ions

    International Nuclear Information System (INIS)

    Ursu, I.; Lupei, V.

    1984-02-01

    A review of the electron paramagnetic resonance data on the uranium ions is given. After a general account of the electronic structure of the uranium free atoms and ions, the influence of the external fields (magnetic field, crystal fields) is discussed. The main information obtained from EPR studies on the uranium ions in crystals are emphasized: identification of the valence and of the ground electronic state, determination of the structure of the centers, crystal field effects, role of the intermediate coupling and of the J-mixing, role of the covalency, determination of the nuclear spin, maqnetic dipole moment and electric quadrupole moment of the odd isotopes of uranium. These data emphasize the fact that the actinide group has its own identity and this is accutely manifested at the beginning of the 5fsup(n) series encompassed by the uranium ions. (authors)

  9. Jabiluka uranium project

    International Nuclear Information System (INIS)

    Anon.

    1991-01-01

    The Jabiluka uranium and gold deposit located in the Northern Territory of Australia is the world's largest known primary uranium deposits and as such has the potential to become one of the most important uranium projects in the world. Despite the financial and structural challenges facing the major owner Pancontinental Mining Limited and the changing political policies in Australia, Jabiluka is well situated for development during the 1990's. With the availability of numerous financial and development alternatives, Jabiluka could, by the turn of the century, take its rightful place among the first rank of world uranium producers. The paper discusses ownership, location, property rights, licensing, environmental concerns, marketing and development, capital costs, royalties, uranium policy considerations, geologic exploration history, regional and site geology, and mining and milling operations

  10. Uranium deposit research, 1983

    International Nuclear Information System (INIS)

    Ruzicka, V.; LeCheminant, G.M.

    1984-01-01

    Research on uranium deposits in Canada, conducted as a prerequisite for assessment of the Estimated Additional Resources of uranium, revealed that (a) the uranium-gold association in rudites of the Huronian Supergroup preferably occurs in the carbon layers; (b) chloritized ore at the Panel mine, Elliot Lake, Ontario, occurs locally in tectonically disturbed areas in the vicinity of diabase dykes; (c) mineralization in the Black Sturgeon Lake area, Ontario, formed from solutions in structural and lithological traps; (d) the Cigar Lake deposit, Saskatchewan, has two phases of mineralization: monomineralic and polymetallic; (e) mineralization of the JEB (Canoxy Ltd.) deposit is similar to that at McClean Lake; (f) the uranium-carbon assemblage was identified in the Claude deposit, Carswell Structure; and (g) the Otish Mountains area, Quebec, should be considered as a significant uranium-polymetallic metallogenic province

  11. Fundamental metallurgical aspects of axial splitting in zircaloy cladding

    International Nuclear Information System (INIS)

    Chung, H. M.

    2000-01-01

    Fundamental metallurgical aspects of axial splitting in irradiated Zircaloy cladding have been investigated by microstructural characterization and analytical modeling, with emphasis on application of the results to understand high-burnup fuel failure under RIA situations. Optical microscopy, SEM, and TEM were conducted on BWR and PWR fuel cladding tubes that were irradiated to fluence levels of 3.3 x 10 21 n cm -2 to 5.9 x 10 21 n cm -2 (E > 1 MeV) and tested in hot cell at 292--325 C in Ar. The morphology, distribution, and habit planes of macroscopic and microscopic hydrides in as-irradiated and posttest cladding were determined by stereo-TEM. The type and magnitude of the residual stress produced in association with oxide-layer growth and dense hydride precipitation, and several synergistic factors that strongly influence axial-splitting behavior were analyzed. The results of the microstructural characterization and stress analyses were then correlated with axial-splitting behavior of high-burnup PWR cladding reported for simulated-RIA conditions. The effects of key test procedures and their implications for the interpretation of RIA test results are discussed

  12. Advances in appendage joining techniques for PHWR fuel cladding

    International Nuclear Information System (INIS)

    Desai, P.B.; Ray, T.K.; Date, V.G.; Purushotham, D.S.C.

    1995-01-01

    This paper describes work carried out at the BARC on the development of a technique to join tiny appendages (spacers and bearing pads) to thin cladding (before loading of UO 2 pellets) by resistance welding for PHWR fuel assemblies. The work includes qualifying the process for production environment, designing prototype equipment for regular production and quality monitoring. In the first phase of development, welding of appendages on UO 2 loaded elements was successfully developed, and is being used in production. Welding of appendages on to empty clad tubes is a superior technique for several reasons. Many problems associated with development of welding on empty tubes were resolved. work was initiated, in the second phase of the development task, to select a suitable technique to join appendages on empty clad tubes without any collapse of thin clad. Several alternatives were reviewed and assessed such as laser, full face welding, shim welding and shrink fitting ring spacers. Selection of a method using a mandrel and a modified electrode geometry was fully developed. Results were optimized and process development successfully completed. Appropriate weld monitoring techniques were also reviewed for their adaptation. This technique is useful for 19, 22 as well as 37 element assemblies. (author)

  13. Prediction of cladding life in waste package environments

    International Nuclear Information System (INIS)

    McCoy, J.K.; Doering, T.W.

    1994-01-01

    Fuel cladding can potentially provide longer containment or slower release of radionuclides from spent fuel after geologic disposal. To predict the amount of benefit that cladding can provide, we surveyed degradation modes and developed a model for creep rupture by diffusion-controlled cavity growth, the mechanism that several authors have concluded is the most important. In this mechanism, voids nucleate on the grain boundaries and grow by diffusion of vacancies along the grain boundaries to the voids. When a certain fraction of the grain boundary area is covered with voids, the material fails. An analytic expression for cladding lifetime is developed. Besides materials constants, the predicted lifetime depends on the temperature history, the hoop stress in the cladding, the spacing between void nuclei, and the micro-structure. The inclusion of microstructure is a significant new feature of the model; this feature is used to help avoid excessive conservatism. The model is applied in a sample calculation for disposal of spent fuel, and the practice of using temperature limits to evaluate repository designs is examined

  14. Production and quality control of fuel cladding tubes for LWRs

    International Nuclear Information System (INIS)

    Matsuda, Katsuhiko; Hagi, Shigeki; Anada, Hiroyuki; Abe, Hideaki; Hyodo, Shigetoshi

    1994-01-01

    This paper reviews the recent fabrication technology and corrosion resistance study of fuel cladding tubes for LWRs conducted by Sumitomo Metal Industries Ltd. started the research on zircaloy in 1957. In 1980, the factory exclusively for the production of cladding tubes was founded, and the mass production system on full scale was established. Thereafter, the various improvement of the production technology, the development of new products, and the heightening of the performance mainly on the corrosion resistance have been tested and studied. Recently, the works in the production processes were almost automated, and the installation of the production lines advanced, and the stabilization of product quality and the rationalization of costs are promoted. Moreover, the development of the zircaloy cladding tubes having high corrosion resistance has been advanced to cope with the long term cycle operation of LWRs hereafter. The features of zircaloy cladding tubes, the manufacturing processes, the improvement of the manufacturing technology, the improvement of the corrosion resistance and so on are reported. (K.I.)

  15. Clad failure detection in G 3 - operational feedback

    International Nuclear Information System (INIS)

    Plisson, J.

    1964-01-01

    After briefly reviewing the role and the principles of clad failure detection, the author describes the working conditions and the conclusions reached after 4 years operation of this installation on the reactor G 3. He mentions also the modifications made to the original installation as well as the tests carried out and the experiments under way. (author) [fr

  16. Radioluminescent clad optical fibre X-ray sensor

    NARCIS (Netherlands)

    Fitzpatrick, C.; O'Donoghue, C.; Lewis, E.; Schöbel, J.; Bastaiens, B.; Bastiaens, Hubertus M.J.; van der Slot, Petrus J.M.

    2003-01-01

    An optical fibre X-ray sensor using cladding to core coupling of radioluminescent emissions from X-ray phosphors is presented. The sensor is tested using a line source normally used in the pre-ionisation of an excimer laser and is calibrated against a pendosimeter and the emission intensity of a

  17. Microstrain Determination in Individual Grains of Laser Deposited Cladding Layers

    NARCIS (Netherlands)

    de Oliveira, Uazir O. B.; Ocelik, Vaclav; De Hosson, Jeff T. M.; Chandra, T; Tsuzaki, K; Militzer, M; Ravindran, C

    2007-01-01

    The laser cladding technique makes the deposition of thick metallic, wear and corrosion resistant coatings feasible on weaker substrates. During the process, localized high thermal gradients generate internal stresses that may cause cracking when these overcome the fracture stress. To explain the

  18. Statistical mechanical analysis of LMFBR fuel cladding tubes

    International Nuclear Information System (INIS)

    Poncelet, J.-P.; Pay, A.

    1977-01-01

    The most important design requirement on fuel pin cladding for LMFBR's is its mechanical integrity. Disruptive factors include internal pressure from mixed oxide fuel fission gas release, thermal stresses and high temperature creep, neutron-induced differential void-swelling as a source of stress in the cladding and irradiation creep of stainless steel material, corrosion by fission products. Under irradiation these load-restraining mechanisms are accentuated by stainless steel embrittlement and strength alterations. To account for the numerous uncertainties involved in the analysis by theoretical models and computer codes statistical tools are unavoidably requested, i.e. Monte Carlo simulation methods. Thanks to these techniques, uncertainties in nominal characteristics, material properties and environmental conditions can be linked up in a correct way and used for a more accurate conceptual design. First, a thermal creep damage index is set up through a sufficiently sophisticated clad physical analysis including arbitrary time dependence of power and neutron flux as well as effects of sodium temperature, burnup and steel mechanical behavior. Although this strain limit approach implies a more general but time consuming model., on the counterpart the net output is improved and e.g. clad temperature, stress and strain maxima may be easily assessed. A full spectrum of variables are statistically treated to account for their probability distributions. Creep damage probability may be obtained and can contribute to a quantitative fuel probability estimation

  19. Foam coating on aluminum alloy with laser cladding

    NARCIS (Netherlands)

    Ocelik, V.; van Heeswijk, V.; de Hosson, J.T.M.; Csach, K.

    dThis article concentrates on the creation of a foam layer on an Al-Si substrate with laser technology. The cladding of At-Si powder in the front of a laser track has been separated from the side injection of mixture of Al-Si/TiH2 powder (foaming agent), which allows for fine tuning of the main

  20. Achilles tests finally nail PWR fuel clad ballooning fears

    International Nuclear Information System (INIS)

    Dore, P.; McMinn, K.

    1992-01-01

    A conclusive series of experiments carried out by AEA Reactor Services at its Achilles rig in the UK has finally allayed fears that fuel clad ballooning is a major safety problem for Sizewell B, Britain's first Pressurized Water Reactor. The experiments are described in this article. (author)

  1. Viscoelastic modelling of Zircaloy cladding in-pile transient creep

    Energy Technology Data Exchange (ETDEWEB)

    Tulkki, Ville, E-mail: ville.tulkki@vtt.fi; Ikonen, Timo

    2015-02-15

    In fuel behaviour modelling accurate description of the cladding stress response is important for both operational and safety considerations. The cladding creep determines in part the width of the gas gap, the duration to pellet-cladding contact and the stresses to the cladding due to the pellet expansion. Conventionally the strain hardening rule has been used to describe the creep response to transient loads in engineering applications. However, it has been well documented that the strain hardening rule does not describe well results of tests with load drops or reversals. In our earlier work we have developed a model for primary creep which can be used to simulate the in- and out-of-pile creep tests. Since then several creep experiments have entered into public domain. In this paper we develop the model formulation based on the theory of viscoelasticity, and show that this model can reproduce the new experimental results. We also show that the creep strain recovery encountered in experimental measurements can be explained by viscoelastic behaviour.

  2. Delayed hydride cracking of Zircaloy-4 fuel cladding

    International Nuclear Information System (INIS)

    Pizarro, Luis M.; Fernandez, Silvia; Lafont, Claudio; Mizrahi, Rafael; Haddad, Roberto

    2007-01-01

    Crack propagation rates, grown by the delayed hydride cracking mechanism, were measured in Zircaloy-4 fuel cladding, according to a Coordinated Research Project (CRP) sponsored by the International Atomic Energy Agency (IAEA). During the first stage of the program a Round Robin Testing was performed on fuel cladding samples provided by Studsvik (Sweden), of the type used in PWR reactors. Crack growth in the axial direction is obtained through the specially developed 'pin load testing' (PLT) device. In these tests, crack propagation rates were determined at 250 C degrees on several samples of the material described above, obtaining a mean value of about 2.5 x 10 -8 m s -1 . The results were analyzed and compared satisfactorily with those obtained by the other laboratories participating in the CRP. At the present moment, similar tests on CANDU and Atucha I type fuel cladding are being performed. It is thought that the obtained results will give valuable information concerning the analysis of possible failures affecting fuel cladding under reactor operation. (author) [es

  3. Uranium in Canada

    International Nuclear Information System (INIS)

    1987-09-01

    Canadian uranium exploration and development efforts in 1985 and 1986 resulted in a significant increase in estimates of measured uranium resources. New discoveries have more than made up for production during 1985 and 1986, and for the elimination of some resources from the overall estimates, due to the sustained upward pressure on production costs and the stagnation of uranium prices in real terms. Canada possesses a large portion of the world's uranium resources that are of current economic interest and remains the major focus of inter-national uranium exploration activity. Expenditures for uranium exploration in Canada in 1985 and 1986 were $32 million and $33 million, respectively. Although much lower than the $130 million total reported for 1979, expenditures for 1987 are forecast to increase. Exploration and surface development drilling in 1985 and 1986 were reported to be 183 000 m and 165σ2 000 m, respectively, 85 per cent of which was in Saskatchewan. Canada has maintained its position as the world's leading producer and exporter of uranium. By the year 2000, Canada's annual uranium requirements will be about 2 100 tU. Canada's known uranium resources are more than sufficient to meet the 30-year fuel requirements of those reactors in Canada that are either in operation now or expected to be in service by the late 1990s. A substantial portion of Canada's identified uranium resources is thus surplus to Canadian needs and available for export. Annual sales currently approach $1 billion, of which exports account for 85 per cent. Forward domestic and export contract commitments totalled 73 000 tU and 62 000 tU, respectively, as of early 1987

  4. An Innovative Ceramic Corrosion Protection System for Zircaloy Cladding

    Energy Technology Data Exchange (ETDEWEB)

    Ronald H. Baney, Dr. D. Butt, Dr. P. Demkowicz, Dr. G. Fuchs Department of Materials Science; James S. Tulenko, Department of Nuclear and Radiological Engineering; University of Florida.

    2003-02-19

    Light Water reactor (LWR) fuel performance is currently limited by thermal, chemical and mechanical constraints associated with the design, fabrication, and operation of the fuel in incore operation. Corrosion of the zirconium based (Zircaloy-4) alloy cladding of the fuel is a primary limiting factor. Recent success at the University of Florida in developing thin ceramic films with great adhesive properties for metal substrates offers an innovative breakthrough for eliminating a major weakness of the Zircaloy clad. ?The University of Florida proposes to coat the existing Zircaloy clad tubes with a ceramic coating for corrosion protection. An added bonus of this approach would be the implementation of a boron-containing burnable poison outer layer will also be demonstrated as part of the ceramic coating development. In this proposed effort, emphasis will be on the ceramic coating with only demonstration of feasibility on the burnable outer coating approach. This proposed program i s expected to give a step change (approximately a doubling) in clad lifetime before failure due to corrosion. In the development of ceramic coatings for Zircaloy-4 clad, silicon carbide and zirconium carbide coatings will first be applied to Zircaloy-4 coupons and cladding samples by thermal assisted chemical vapor deposition, plasma assisted chemical vapor deposition or by laser ablation deposition. All of these processes are in use at the University of Florida and have shown great potential. The questions of adhesion and thermal expansion mismatch of the ceramic coating to the Zircaloy substrate will be addressed. Several solutions to these conditions will be examined, if needed. These solutions include the use of a zirconium oxide compliant layer, employment of a laser roughened surface and the use of a gradient composition interlayer. These solutions have already been shown to be effective for other high modulus coatings on metal substrates. Mechanical properties and adhesion of the

  5. Hydrogenation and high temperature oxidation of Zirconium claddings

    International Nuclear Information System (INIS)

    Novotny, T.; Perez-Feró, E.; Horváth, M.

    2015-01-01

    In the last few years a new series of experiments started for supporting the new LOCA criteria, considering the proposals of US NRC. The effects which can cause the embrittlement of VVER fuel claddings were reviewed and evaluated in the framework of the project. The purpose of the work was to determine how the fuel cladding’s hydrogen uptake under normal operating conditions, effect the behavior of the cladding under LOCA conditions. As a first step a gas system equipment with gas valves and pressure gauge was built, in which the zirconium alloy can absorb hydrogen under controlled conditions. In this apparatus E110 (produced by electrolytic method, currently used at Paks NPP) and E110G (produced by a new technology) alloys were hydrogenated to predetermined hydrogen contents. According the results of ring compression tests the E110G alloys lose their ductility above 3200 ppm hydrogen content. This limit can be applied to determine the ductile-brittle transition of the nuclear fuel claddings. After the hydrogenation, high temperature oxidation experiments were carried out on the E110G and E110 samples at 1000 °C and 1200 °C. 16 pieces of E110G and 8 samples of E110 with 300 ppm and 600 ppm hydrogen content were tested. The oxidation of the specimens was performed in steam, under isothermal conditions. Based on the ring compression tests load-displacement curves were recorded. The main objective of the compression tests was to determine the ductile-brittle transition. These results were compared to the results of our previous experiments where the samples did not contain hydrogen. The original claddings showed more ductile behavior than the samples with hydrogen content. The higher hydrogen content resulted in a more brittle mechanical behavior. However no significant difference was observed in the oxidation kinetics of the same cladding types with different hydrogen content. The experiments showed that the normal operating hydrogen uptake of the fuel claddings

  6. Steam attack studies in SiC clad and advanced steel clad

    International Nuclear Information System (INIS)

    Terrani, Kurt; Snead, Lance; Pint, Bruce

    2013-01-01

    The most recent generation of zirconium alloys used as nuclear fuel cladding in LWRs offer exceptional performance under normal operating conditions with failure rates below 1 ppm. This level of performance is due to six decades of active research and development by nuclear operators, vendors, government laboratories, and academia. During this same period, iron alloys, which were successfully used in the early days of nuclear power but supplanted by zirconium alloys, have undergone broad-based development including specialty specialised steels for high-temperature corrosion and steam resistance. This coupled with the recent trends in fuel handling, reactor operation and the extreme control of coolant chemistry begs for re-examination of the application of advanced iron-alloys as LWR fuel cladding, specifically, oxidation resistant iron alloys that offer large margins of safety under severe accident conditions. Recently, steam oxidation tests have been performed on a wide range of austenitic and ferritic iron alloys at ORNL where the dependence of the oxidation kinetics on alloy chemistry has been determined. As the temperature increases the minimum Cr content in the ferritic and austenitic alloys needs to be increased for oxidation resistance. At temperatures ≥1200 deg. C ∼25% Cr content is necessary for adequate oxidation resistance. At the fixed Cr content, an increase in Ni content in the austenitic alloys enhances the oxidation resistance. Note that the 25% minimum Cr content disqualifies the conventional austenitic (18Cr-8Ni) alloys such as 304L as oxidation resistant materials in steam. The most promising alloys are ferritic alloys that contain some Al in addition to Cr. FeCrAl alloys with 20% Cr and 5% Al exhibit exceptionally low oxidation in high-temperature steam up to 1300 deg. C due to formation of a protective alumina film on the surface of the material. High-pressure steam oxidation tests (up to 2 MPa) were also performed that showed the effect of

  7. Thermal gradient effects on the oxidation of Zircaloy fuel cladding

    International Nuclear Information System (INIS)

    Klein, A.C.; Reyes, J.N. Jr.; Maguire, M.A.

    1990-01-01

    A Thermal Gradient Test Facility (TGTF) has been designed and constructed to measure the thermal gradient effect on pressurized water reactor (PWR) fuel rod cladding. The TGTF includes a heat flux simulator assembly capable of producing a wide range of PWR operating conditions including water flow velocities and temperatures, water chemistry conditions, cladding temperatures, and heat fluxes ranging to 160 W/cm 2 . It is fully instrumented including a large number of thermocouples both inside the water flow channel and inside the cladding. Two test programs are in progress. First, cladding specimens are pre-oxidized in air at 500 deg. C and in 400 deg. C steam for various lengths of time to develop a range of uniform oxide thicknesses from 1 to 60 micrometers. The pre-oxidized specimens are placed in the TGTF to characterize the oxide thermal conductivity under a variety of water flow and heat flux conditions. Second, to overcome the long exposure times required under typical PWR conditions a series of tests with the addition of high concentrations of lithium hydroxide to the water are being considered. Static autoclave tests have been conducted with lithium hydroxide concentrations ranging from 0 to 2 moles per liter at 300, 330, and 360 deg. C for up to 36 hours. Results for zircaloy-4 show a considerable increase in the weight gain for the exposed samples with oxidation rate enhancement factors as high as 70 times that of pure water. Operation of the TGTF with elevated lithium hydroxide levels will yield real-time information concerning the effects of a heat flux on the oxidation kinetics of zircaloy fuel rod cladding. (author). 5 refs, 5 figs, 2 tabs

  8. Uranium producers foresee new boom

    International Nuclear Information System (INIS)

    McIntyre, H.

    1979-01-01

    The status of uranium production in Canada is reviewed. Uranium resources in Saskatchewan and Ontario are described and the role of the Cluff Lake inquiry in securing a government decision in favour of further uranium development is mentioned. There have been other uranium strikes near Kelowna, British Columbia and in the Northwest Territories. Increasing uranium demand and favourable prices are making the development of northern resources economically attractive. In fact, all uranium currently produced has been committed to domestic and export contracts so that there is considerable room for expanding the production of uranium in Canada. (T.I.)

  9. Irradiation effects of the zirconium oxidation and the uranium diffusion in zirconia

    International Nuclear Information System (INIS)

    Bererd, N.

    2003-01-01

    The context of this study is the direct storage of spent fuel assemblies after operation in reactor. In order to obtain data on the capacities of the can as the uranium diffusion barrier, a fundamental study has been carried out for modelling the internal cladding surface under and without irradiation. The behaviour of zirconium in reactor conditions has at first been studied. A thin uranium target enriched with fissile isotope has been put on a zirconium sample, the set being irradiated by a thermal neutrons flux leading to the fission of the deposited uranium. The energetic history of the formed fission products has revealed two steps: 1)the zirconium oxidation and 2)the diffusion of uranium in the zirconia formed at 480 degrees C. A diffusion coefficient under irradiation has been measured. Its value is 10 -15 cm 2 .s -1 . In order to be able to reveal clearly the effect of the irradiation by the fission products on the zirconium oxidation, measurements of thermal oxidation and under 129 Xe irradiation have been carried out. They have shown that the oxidation is strongly accelerated by the irradiation and that the temperature is negligible until 480 degrees C. On the other hand, the thermal diffusion of the uranium in zirconium and in zirconia has been studied by coupling ion implantation and Rutherford backscattering spectroscopy. This study shows that the uranium diffuses in zirconium and is trapped in zirconia in a UO 3 shape. (O.M.)

  10. Development of high uranium-density fuels for use in research reactors

    International Nuclear Information System (INIS)

    Ugajin, Mitsuhiro; Akabori, Mitsuo; Itoh, Akinori

    1996-01-01

    The uranium silicide U 3 Si 2 possesses uranium density 11.3 gU/cm 3 with a congruent melting point of 1665degC, and is now successfully in use as a research reactor fuel. Another uranium silicide U 3 Si and U 6 Me-type uranium alloys (Me=Fe,Mn,Ni) have been chosen as new fuel materials because of the higher uranium densities 14.9 and 17.0 gU/cm 3 , respectively. Experiments were carried out to fabricate miniature aluminum-dispersion plate-type and aluminum-clad disk-type fuels by using the conventional picture-frame method and a hot-pressing technique, respectively. These included the above-mentioned new fuel materials as well as U 3 Si 2 . Totally 14 miniplates with uranium densities from 4.0 to 6.3 gU/cm 3 of fuel meat were prepared together with 28 disk-type fuel containing structurally-modified U 3 Si, and subjected to the neutron irradiation in JMTR (Japan Materials Testing Reactor). Some results of postirradiation examinations are presented. (author)

  11. An allowable cladding peak temperature for spent nuclear fuels in interim dry storage

    Science.gov (United States)

    Cha, Hyun-Jin; Jang, Ki-Nam; Kim, Kyu-Tae

    2018-01-01

    Allowable cladding peak temperatures for spent fuel cladding integrity in interim dry storage were investigated, considering hydride reorientation and mechanical property degradation behaviors of unirradiated and neutron irradiated Zr-Nb cladding tubes. Cladding tube specimens were heated up to various temperatures and then cooled down under tensile hoop stresses. Cool-down specimens indicate that higher heat-up temperature and larger tensile hoop stress generated larger radial hydride precipitation and smaller tensile strength and plastic hoop strain. Unirradiated specimens generated relatively larger radial hydride precipitation and plastic strain than did neutron irradiated specimens. Assuming a minimum plastic strain requirement of 5% for cladding integrity maintenance in interim dry storage, it is proposed that a cladding peak temperature during the interim dry storage is to keep below 250 °C if cladding tubes are cooled down to room temperature.

  12. Effect of Co - based Alloy on Properties of Laser Cladding Layer

    Science.gov (United States)

    Yang, Y.; Jiang, Z. P.; Li, H. Z.

    2017-11-01

    A large number of laser cladding experiments have been carried out using 20CrMnTi steel as substrate and Co-based alloy as cladding material. The influence of Co-based alloy on the laser cladding properties of 20CrMnTi steel was studied by analyzing the macroscopic and microscopic characteristics of cladding crack susceptibility, dilution rate, microstructure and friction and wear properties. The results show that the high-power laser cladding of Co-based material can obtain a flat defect-free cladding layer with compact structure and low crack susceptibility. A multi-layer cladding strategy with variable power can be used to fabricate thin wall structures without collapse Parts, the surface smooth without pores.

  13. Uranium tipped ammunition

    International Nuclear Information System (INIS)

    Roche, P.

    1993-01-01

    During the uranium enrichment process required to make nuclear weapons or fuel, the concentration of the 'fissile' U-235 isotope has to be increased. What is left, depleted uranium, is about half as radioactive as natural uranium, but very dense and extremely hard. It is used in armour piercing shells. External radiation levels from depleted uranium (DU) are low. However DU is about as toxic as lead and could be harmful to the kidneys if eaten or inhaled. It is estimated that between 40 and 300 tonnes of depleted uranium were left behind by the Allied armies after the Gulf war. The biggest hazard would be from depleted uranium shells which have hit Iraqui armoured vehicles and the resulting dust inhaled. There is a possible link between depleted uranium shells and an illness known as 'Desert Storm Syndrome' occurring in some Gulf war veterans. As these shells are a toxic and radioactive hazard to health and the environment their use and testing should be stopped because of the risks to troops and those living near test firing ranges. (UK)

  14. Purification of uranium metal

    International Nuclear Information System (INIS)

    Suzuki, Kenji; Shikama, Tatsuo; Ochiai, Akira.

    1993-01-01

    We developed the system for purifying uranium metal and its metallic compounds and for growing highly pure uranium compounds to study their intrinsic physical properties. Uranium metal was zone refined under low contamination conditions as far as possible. The degree of the purity of uranium metal was examined by the conventional electrical resistivity measurement and by the chemical analysis using the inductive coupled plasma emission spectrometry (ICP). The results show that some metallic impurities evaporated by the r.f. heating and other usual metallic impurities moved to the end of a rod with a molten zone. Therefore, we conclude that the zone refining technique is much effective to the removal of metallic impurities and we obtained high purified uranium metal of 99.99% up with regarding to metallic impurities. The maximum residual resistivity ratio, the r.r.r., so far obtained was about 17-20. Using the purified uranium, we are attempting to grow a highly pure uranium-titanium single crystals. (author)

  15. Uranium deposits in Africa

    International Nuclear Information System (INIS)

    Wilpolt, R.H.; Simov, S.D.

    1979-01-01

    Africa is not only known for its spectacular diamond, gold, copper, chromium, platinum and phosphorus deposits but also for its uranium deposits. At least two uranium provinces can be distinguished - the southern, with the equatorial sub-province; and the south Saharan province. Uranium deposits are distributed either in cratons or in mobile belts, the first of sandstone and quartz-pebble conglomerate type, while those located in mobile belts are predominantly of vein and similar (disseminated) type. Uranium deposits occur within Precambrian rocks or in younger platform sediments, but close to the exposed Precambrian basement. The Proterozoic host rocks consist of sediments, metamorphics or granitoids. In contrast to Phanerozoic continental uranium-bearing sediments, those in the Precambrian are in marginal marine facies but they do contain organic material. The geology of Africa is briefly reviewed with the emphasis on those features which might control the distribution of uranium. The evolution of the African Platform is considered as a progressive reduction of its craton area which has been affected by three major Precambrian tectonic events. A short survey on the geology of known uranium deposits is made. However, some deposits and occurrences for which little published material is available are treated in more detail. (author)

  16. The uranium in the environment

    International Nuclear Information System (INIS)

    2001-01-01

    The uranium is a natural element omnipresent in the environment, with a complex chemistry more and more understood. Many studies are always today devoted to this element to better improve the uranium behavior in the environment. To illustrate this knowledge and for the public information the CEA published this paper. It gathers in four chapters: historical aspects and properties of the uranium, the uranium in the environment and the impacts, the metrology of the uranium and its migration. (A.L.B.)

  17. Uranium resource assessments

    International Nuclear Information System (INIS)

    1981-01-01

    The objective of this investigation is to examine what is generally known about uranium resources, what is subject to conjecture, how well do the explorers themselves understand the occurrence of uranium, and who are the various participants in the exploration process. From this we hope to reach a better understanding of the quality of uranium resource estimates as well as the nature of the exploration process. The underlying questions will remain unanswered. But given an inability to estimate precisely our uranium resources, how much do we really need to know. To answer this latter question, the various Department of Energy needs for uranium resource estimates are examined. This allows consideration of whether or not given the absence of more complete long-term supply data and the associated problems of uranium deliverability for the electric utility industry, we are now threatened with nuclear power plants eventually standing idle due to an unanticipated lack of fuel for their reactors. Obviously this is of some consequence to the government and energy consuming public. The report is organized into four parts. Section I evaluates the uranium resource data base and the various methodologies of resource assessment. Part II describes the manner in which a private company goes about exploring for uranium and the nature of its internal need for resource information. Part III examines the structure of the industry for the purpose of determining the character of the industry with respect to resource development. Part IV arrives at conclusions about the emerging pattern of industrial behavior with respect to uranium supply and the implications this has for coping with national energy issues

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

    International Nuclear Information System (INIS)

    2009-12-01

    A large variety of research reactor spent fuel with different fuel meats, different geometries and different enrichments in 235 U are presently stored underwater in basins located around the world. More than 90% of these fuels are clad in aluminium or aluminium based alloys that are notoriously susceptible to corrosion in water of less than optimum quality. Some fuel is stored in the reactor pools themselves, some in auxiliary pools (or basins) close to the reactor and some stored at away-from-reactor pools. Since the early 1990s, when corrosion induced degradation of the fuel cladding was observed in many of the pools, corrosion of research reactor aluminium clad spent nuclear fuel stored in light water filled basins has become a major concern, and programmes were implemented at the sites to improve fuel storage conditions. The IAEA has since then established a number of programmatic activities to address corrosion of research reactor aluminium clad spent nuclear fuel in water. Of special relevance was the Coordinated Research Project (CRP) on Corrosion of Research Reactor Aluminium Clad Spent Fuel in Water (Phase I) initiated in 1996, whose results were published in IAEA Technical Reports Series No. 418. At the end of this CRP it was considered necessary that a continuation of the CRP should concentrate on fuel storage basins that had demonstrated significant corrosion problems and would therefore provide additional insight into the fundamentals of localized corrosion of aluminium. As a consequence, the IAEA started a new CRP entitled Corrosion of Research Reactor Aluminium Clad Spent Fuel in Water (Phase II), to carry out more comprehensive research in some specific areas of corrosion of aluminium clad spent nuclear fuel in water. In addition to this CRP, one of the activities under IAEA's Technical Cooperation Regional Project for Latin America Management of Spent Fuel from Research Reactors (2001-2006) was corrosion monitoring and surveillance of research

  19. Depleted uranium management alternatives

    Energy Technology Data Exchange (ETDEWEB)

    Hertzler, T.J.; Nishimoto, D.D.

    1994-08-01

    This report evaluates two management alternatives for Department of Energy depleted uranium: continued storage as uranium hexafluoride, and conversion to uranium metal and fabrication to shielding for spent nuclear fuel containers. The results will be used to compare the costs with other alternatives, such as disposal. Cost estimates for the continued storage alternative are based on a life-cycle of 27 years through the year 2020. Cost estimates for the recycle alternative are based on existing conversion process costs and Capital costs for fabricating the containers. Additionally, the recycle alternative accounts for costs associated with intermediate product resale and secondary waste disposal for materials generated during the conversion process.

  20. Vacuum fusion of uranium

    International Nuclear Information System (INIS)

    Stohr, J.A.

    1957-01-01

    After having outlined that vacuum fusion and moulding of uranium and of its alloys have some technical and economic benefits (vacuum operations avoid uranium oxidation and result in some purification; precision moulding avoids machining, chip production and chemical reprocessing of these chips; direct production of the desired shape is possible by precision moulding), this report presents the uranium fusion unit (its low pressure enclosure and pumping device, the crucible-mould assembly, and the MF supply device). The author describes the different steps of cast production, and briefly comments the obtained results

  1. Uranium absorption study pile

    International Nuclear Information System (INIS)

    Raievski, V.; Sautiez, B.

    1959-01-01

    The report describes a pile designed to measure the absorption of fuel slugs. The pile is of graphite and comprises a central section composed of uranium rods in a regular lattice. RaBe sources and BF 3 counters are situated on either side of the center. A given uranium charge is compared with a specimen charge of about 560 kg, and the difference in absorption between the two noted. The sensitivity of the equipment will detect absorption variations of about a few ppm boron (10 -6 boron per gr. of uranium) or better. (author) [fr

  2. Uranium mining and milling

    International Nuclear Information System (INIS)

    Floeter, W.

    1976-01-01

    In this report uranium mining and milling are reviewed. The fuel cycle, different types of uranium geological deposits, blending of ores, open cast and underground mining, the mining cost and radiation protection in mines are treated in the first part of this report. In the second part, the milling of uranium ores is treated, including process technology, acid and alkaline leaching, process design for physical and chemical treatment of the ores, and the cost. Each chapter is clarified by added figures, diagrams, tables, and flowsheets. (HK) [de

  3. Uranium in coal

    International Nuclear Information System (INIS)

    Facer, J.F. Jr.

    1979-05-01

    United States production of coal in 1977 was 695 million short tons of which 477 million tons were burned in power plants. The ash from these power plants was about 67 million tons containing an estimated 900 tons U 3 O 8 , assuming 14 percent ash from the type of coal used by utilities and 12 ppM U contained in ash. Perhaps 1 to 3 percent of the domestic uranium requirement could be met from coal ash, provided processing technology could be developed for uranium recovery at acceptable costs. However, the environmental problems for disposal of the slimy leached ash would be enormous. The average uranium grade of coal in the United States is less than half of that of the Earth's crust. Burning the coal concentrates the contained uranium in the ash from 2 to 20 times. However, even at 20 times concentration, the percent uranium in coal ash is less than 1/100 of the grade of the uranium ore being processed today from conventional deposits. Although it is conceivable that some coal ash might contain enough uranium to make the ash an economic source of uranium, this is not now the case for ash from any major coal-fired power plant in the United States. During 1963 to 67, about 180,000 tons of uranium-bearing carbonaceous rock from the southwestern part of the Williston Basin were mined and processed to recover about 1 million pounds of U 3 O 8 . None of this material has been mined since 1967. The uranium reserves of the area are small, and the environmental problems with calcining the lignitic material may be prohibitive. Some other uraniferous coal and lignite could be mined and processed as a uranium ore, but less than half of one percent of the domestic $30 reserves are in coal. A few samples of mid-continent coal have been reported to contain about 100 ppM U but little is known about the size of such deposits or the likelihood that they will be mined and used for power plant fuel to produce a high-uranium ash

  4. Possible uranium sources of Streltsovsky uranium ore field

    International Nuclear Information System (INIS)

    Zhang Lisheng

    2005-01-01

    The uranium deposit of the Late Jurassic Streltsovaky caldera in Transbaikalia of Russia is the largest uranium field associated with volcanics in the world, its uranium reserves are 280 000 t U, and it is the largest uranium resources in Russia. About one third of the caldera stratigraphic pile consists of strongly-altered rhyolites. Uranium resources of the Streltsovsky caldera are much larger than any other volcanic-related uranium districts in the world. Besides, the efficiency of hydrothermal alteration, uranium resources appear to result from the juxtaposition of two major uranium sources; highly fractionated peralkaline rhyolites of Jurassic age in the caldera, and U-rich subalkaline granites of Variscan age in the basement in which the major uranium-bearing accessory minerals were metamict at the time of the hydrothermal ore formation. (authors)

  5. Uranium in alkaline rocks

    International Nuclear Information System (INIS)

    Murphy, M.; Wollenberg, H.; Strisower, B.; Bowman, H.; Flexser, S.; Carmichael, I.

    1978-04-01

    Geologic and geochemical criteria were developed for the occurrence of economic uranium deposits in alkaline igneous rocks. A literature search, a limited chemical analytical program, and visits to three prominent alkaline-rock localities (Ilimaussaq, Greenland; Pocos de Caldas, Brazil; and Powderhorn, Colorado) were made to establish criteria to determine if a site had some uranium resource potential. From the literature, four alkaline-intrusive occurrences of differing character were identified as type-localities for uranium mineralization, and the important aspects of these localities were described. These characteristics were used to categorize and evaluate U.S. occurrences. The literature search disclosed 69 U.S. sites, encompassing nepheline syenite, alkaline granite, and carbonatite. It was possible to compare two-thirds of these sites to the type localities. A ranking system identified ten of the sites as most likely to have uranium resource potential

  6. PURIFICATION OF URANIUM FUELS

    Science.gov (United States)

    Niedrach, L.W.; Glamm, A.C.

    1959-09-01

    An electrolytic process of refining or decontaminating uranium is presented. The impure uranium is made the anode of an electrolytic cell. The molten salt electrolyte of this cell comprises a uranium halide such as UF/sub 4/ or UCl/sub 3/ and an alkaline earth metal halide such as CaCl/sub 2/, BaF/sub 2/, or BaCl/sub 2/. The cathode of the cell is a metal such as Mn, Cr, Co, Fe, or Ni which forms a low melting eutectic with U. The cell is operated at a temperature below the melting point of U. In operation the electrodeposited uranium becomes alloyed with the metal of the cathode, and the low melting alloy thus formed drips from the cathode.

  7. Uranium market activities

    International Nuclear Information System (INIS)

    Patterson, J.A.

    1975-01-01

    Results are summarized from the 1974 ERDA annual survey of buyers and sellers and from a survey of uranium price data which provided information on additional domestic buying activity during the first half of 1975 through 1982

  8. Uranium industry seminar: proceedings

    International Nuclear Information System (INIS)

    1981-01-01

    The eleventh annual Uranium Industry Seminar, sponsored by the Grand Junction Area Office of the US Department of Energy (DOE), was held in Grand Junction, Colorado, on October 21 and 22, 1981. There were 491 registered attendees as compared to 700 attending the previous year. The attendees were largely from uranium and other energy resource companies, electric utility firms, energy consultants and service companies, and governmental agencies. In addition, there were representatives present from Indian tribes, universities, the media, DOE laboratories, and foreign countries and organizations. Papers presented at the seminar dealt with uranium policies, exploration, resources, supply, enrichment, and market conditions. There also were papers on the National Uranium Resource Evaluation Program and international activities. Thirteen papers included in this report have been abstracted and indexed

  9. Uranium extraction from seawater

    International Nuclear Information System (INIS)

    Bals, H.G.

    1978-01-01

    Problems and work of other institutes in the field of uranium extraction from sea water are presented. UEB developments, UEB processes, and work done to realize the UEB concept are discussed. (HK) [de

  10. Uranium dioxide pellets

    International Nuclear Information System (INIS)

    Zawidzki, T.W.

    1982-01-01

    A process for the preparation of a sintered, high density, large crystal grain size uranium dioxide pellet is described which involves: (i) reacting a uranyl nitrate of formula UO 2 (NO 3 ) 2 .6H 2 O with a sulphur source, at a temperature of from about 300 deg. C to provide a sulphur-containing uranium trioxide; (ii) reacting the thus-obtained modified uranium trioxide with ammonium nitrate to form an insoluble sulphur-containing ammonium uranate; (iii) neutralizing the thus-formed slurry with ammonium hydroxide to precipitate out as an insoluble ammonium uranate the remaining dissolved uranium; (iv) recovering the thus-formed precipitates in a dry state; (v) reducing the dry precipitate to UO 2 , and forming it into 'green' pellets; and (vi) sintering the pellets in a hydrogen atmosphere at an elevated temperature

  11. Uranium in alkaline rocks

    Energy Technology Data Exchange (ETDEWEB)

    Murphy, M.; Wollenberg, H.; Strisower, B.; Bowman, H.; Flexser, S.; Carmichael, I.

    1978-04-01

    Geologic and geochemical criteria were developed for the occurrence of economic uranium deposits in alkaline igneous rocks. A literature search, a limited chemical analytical program, and visits to three prominent alkaline-rock localities (Ilimaussaq, Greenland; Pocos de Caldas, Brazil; and Powderhorn, Colorado) were made to establish criteria to determine if a site had some uranium resource potential. From the literature, four alkaline-intrusive occurrences of differing character were identified as type-localities for uranium mineralization, and the important aspects of these localities were described. These characteristics were used to categorize and evaluate U.S. occurrences. The literature search disclosed 69 U.S. sites, encompassing nepheline syenite, alkaline granite, and carbonatite. It was possible to compare two-thirds of these sites to the type localities. A ranking system identified ten of the sites as most likely to have uranium resource potential.

  12. Ontario's uranium mining industry

    International Nuclear Information System (INIS)

    Runnalls, O.J.C.

    1981-01-01

    This report traces the Ontario uranium mining industry from the first discovery of uranium north of Sault Ste. Marie through the uranium boom of the 1950's when Elliot Lake and Bancroft were developed, the cutbacks of the 1960s, the renewed enthusiasm in exploration and development of the 1970s to the current position when continued production for the domestic market is assured. Ontario, with developed mines and operational expertise, will be in a position to compete for export markets as they reopen. The low level of expenditures for uranium exploration and the lack of new discoveries are noted. The report also reviews and places in perspective the development of policies and regulations governing the industry and the jurisdictional relationships of the Federal and Provincial governments

  13. Uranium Research in Senegal

    International Nuclear Information System (INIS)

    Kanouté, Mamadou

    2015-01-01

    The work of mining companies have so far not proved economic uranium resources, but they have nevertheless contributed greatly to a better understanding of the geology, particularly in Eastern Senegal, on the upper Precambrian basin including which equivalents exist throughout West Africa (the uranium belt of Zaire) prospected by CEA-COGEMA teams. The researches carried out in Senegal, but also in Guinea and Mali helped establish a detailed map and understand the course of geological history. With new exploration techniques and data of airborne geophysical (radiometric) provided by the Mining Sector Support Programme (PASMI 9th EDF 9 ACP SE 09), AREVA, at the end of the first period validity of the exploration permit increased significantly, the resources. Prospects are favorable to a doubling of resources; objective of a uranium mine in Senegal. Synergies are possible and desirable with joint exploitation of uranium deposits located in Mali, near the border with Senegal.

  14. Uranium industry seminar

    International Nuclear Information System (INIS)

    1980-01-01

    The tenth annual Uranium Industry Seminar, sponsored by the US Department of Energy's (DOE) Grand Junction Office, was held in Grand Junction, Colorado, on October 22 and 23, 1980. There were 700 registered attendees as compared to 833 attending the previous year. The attendees were drawn largely from uranium and other energy resource companies, electric utility firms, energy consultants and service companies, and governmental agencies. In addition, there were representatives present from Indian tribes, universities, the media, DOE laboratories, and foreign countries and organizations. There were 14 papers presented at the seminar by speakers from the Department of Energy, US Geological Survey, and Bendix Field Engineering Corporation which is the on-site prime contractor for DOE's Grand Junction Office. The topics the papers dealt with were uranium policies, exploration, respources, supply, enrichment, and market conditions. There also were papers describing the National Uranium Resource Evaluation program and international activities. All 14 papers in this Proceedings have been abstracted and indexed

  15. The uranium market prospects

    International Nuclear Information System (INIS)

    Lloyd, R.

    1981-01-01

    A historical analysis of the uranium market points out the cyclical nature of the market and suggests that the spot price, exploration levels, and mill capacity utilization rate are dependent on economic factors. An examination of the current uranium market suggests that the effects of the forecasted surplus supply, the diminishing returns in exploration and the long lead times and high costs of development may mean that future production levels are uncertain. The general prospects for the uranium industry are also uncertain because of barriers to trade, environmental regulations and public opinion. The paper concludes that by the use of long term contracts, appropriate inventory policy and greater discussion between producers and consumers the prospects for the uranium market can be made more certain and further imbalances in demand and supply can be avoided. (author)

  16. Uranium in granites

    International Nuclear Information System (INIS)

    Maurice, Y.T.

    1982-01-01

    Recent research activities of the Canadian Uranium in Granites Study are presented in 18 papers and 3 abstracts. 'Granites' is used as a generic term for granitoids, granitic rocks, and plutonic rocks

  17. Uranium determination in zirconium

    International Nuclear Information System (INIS)

    Serdeiro, N.H.; Bianco de Salas, G.H.; Correia, R.J.

    1987-01-01

    The method used for the spectrometric uranium determination with 2-(2-thiolase)-5-diethylaminophenol was modified for its application in the zirconium samples analysis with an uranium content of the 0.1% order. The samples, previously dissolved in nitric acid, were submitted to a separative stage of liquid-liquid extraction, with a trioctylphosphine (TOPO) oxide diluted in cyclohexane. A sodium fluoride aqueous solution was necessary to be aggregated in the spectrometric determination so as to complex the zirconium vestiges, which could be present, originated by the Zr/U high relation of the initial sample. Under the established working conditions, different spectrometric assays, dyes absorption spectra and its uranium complex, complex stability, PH influence determination of the dyes-uranium relation, calculation of the complex's apparent formation constant and its molar absorption, were performed. (Author)

  18. Uranium purchases report 1993

    International Nuclear Information System (INIS)

    1994-01-01

    Data reported by domestic nuclear utility companies in their responses to the 1991 through 1993 ''Uranium Industry Annual Survey,'' Form EIA-858, Schedule B,'' Uranium Marketing Activities,'' are provided in response to the requirements in the Energy Policy Act 1992. Appendix A contains an explanation of Form EIA-858 survey methodologies with emphasis on the processing of Schedule B data. Additional information published in this report not included in Uranium Purchases Report 1992, includes a new data table. Presented in Table 1 are US utility purchases of uranium and enrichment services by origin country. Also, this report contains additional purchase information covering average price and contract duration. Table 2 is an update of Table 1 and Table 3 is an update of Table 2 from the previous year's report. The report contains a glossary of terms

  19. Uranium hexafluoride purification

    International Nuclear Information System (INIS)

    Araujo, Eneas F. de

    1986-01-01

    Uranium hexafluoride might contain a large amount of impurities after manufacturing or handling. Three usual methods of purification of uranium hexafluoride were presented: selective sorption, sublimation, and distillation. Since uranium hexafluoride usually is contaminated with hydrogen fluoride, a theoretical study of the phase equilibrium properties was performed for the binary system UF 6 -HF. A large deviation from the ideal solution behaviour was observed. A purification unity based on a constant reflux batch distillation process was developed. A procedure was established in order to design the re boiler, condenser and packed columns for the UF 6 -HF mixture separation. A bench scale facility for fractional distillation of uranium hexafluoride was described. Basic operations for that facility and results extracted from several batches were discussed. (author)

  20. Hydrogen in uranium

    International Nuclear Information System (INIS)

    Darras, R.; Caillat, R.

    1959-01-01

    Hydrogen impairs the properties of metals, in particular this can assume considerable importance with uranium when used as a fuel material. Hydrogen can be associated with uranium in two distinct ways: either dissolved in the metal, or in the form of a hydride. The solubility of hydrogen in uranium depends on the allotropic form of the metal; it is in addition greater in molten than in solid uranium. Two types of hydride, both given by the formula UH 3 , have been identified. The first, the α-type. is stable at low temperature. It is generally accompanied by small amounts of the β-type, into which it is wholly transformed above 140 deg.C. Reprint of a paper published in Progress in Nuclear Energy, Series V, Vol. 2, 'Metallurgy and Fuels', p. 19-27

  1. Uranium Location Database

    Data.gov (United States)

    U.S. Environmental Protection Agency — A GIS compiled locational database in Microsoft Access of ~15,000 mines with uranium occurrence or production, primarily in the western United States. The metadata...

  2. Uranium in Canada

    International Nuclear Information System (INIS)

    1989-01-01

    In 1988 Canada's five uranium producers reported output of concentrate containing a record 12,470 metric tons of uranium (tU), or about one third of total Western world production. Shipments exceeded 13,200 tU, valued at $Cdn 1.1 billion. Most of Canada's uranium output is available for export for peaceful purposes, as domestic requirements represent about 15 percent of production. The six uranium marketers signed new sales contracts for over 11,000 tU, mostly destined for the United States. Annual exports peaked in 1987 at 12,790 tU, falling back to 10,430 tU in 1988. Forward domestic and export contract commitments were more than 70,000 tU and 60,000 tU, respectively, as of early 1989. The uranium industry in Canada was restructured and consolidated by merger and acquisition, including the formation of Cameco. Three uranium projects were also advanced. The Athabasca Basin is the primary target for the discovery of high-grade low-cost uranium deposits. Discovery of new reserves in 1987 and 1988 did not fully replace the record output over the two-year period. The estimate of overall resources as of January 1989 was down by 4 percent from January 1987 to a total (measured, indicated and inferred) of 544,000 tU. Exploration expenditures reached $Cdn 37 million in 1987 and $59 million in 1988, due largely to the test mining programs at the Cigar Lake and Midwest projects in Saskatchewan. Spot market prices fell to all-time lows from 1987 to mid-1989, and there is little sign of relief. Canadian uranium production capability could fall below 12,000 tU before the late 1990s; however, should market conditions warrant output could be increased beyond 15,000 tU. Canada's known uranium resources are more than sufficient to meet the 30-year fuel requirements of those reactors in Canada that are now or are expected to be in service by the late 1990s. There is significant potential for discovering additional uranium resources. Canada's uranium production is equivalent, in

  3. US uranium market developments

    International Nuclear Information System (INIS)

    Krusiewski, S.V.; Thomas, D.C.

    1981-01-01

    Domestic uranium delivery commitments for the 1981 to 1990 period reached a peak in the July 1980 survey and then declined in the January 1981 survey and again in the July 1981 survey. However, there are sizable sales contracts through the mid-1980s. In the latter part of this decade, unfilled requirements increase which can provide a needed market for domestic producers. Older contracts are helping to keep the average contract prices, including market price settlements, rather stable. However, average market price settlements decreased from data reported in January 1981, but some of these deliveries represent settlement of litigation. Foreign uranium procurement is scheduled to exceed deliveries of US uranium to foreign buyers in the 1981 to 1990 period. However, the actual use of foreign uranium has been quite low as US enrichment services customers have preferred to buy US uranium. Based on over four and one-half years of data, only about 7% foreign uranium has been brought to the Department of Energy for enrichment. Inventories of natural and enriched uranium in buyers' hands continue to increase. This is a concern to the uranium-producing industry. However, the industry should not be concerned about DOE-owned inventories, which are needed to supply Government requirements. There is absolutely no plan to dispose of DOE inventories on the commercial market. Capital expenditures reached a peak of $800 million in 1979. This decreased to $780 million in 1980, although higher expenditures were planned for the year. A very sharp reduction in plans for 1981, from $830 to $450 million, has been reported. A further reduction to $350 million is planned for 1982. However, it is interesting to note that the planned expenditures for 1982 are above the expenditures for 1975, a period of industury expansion

  4. Recovery of uranium values

    International Nuclear Information System (INIS)

    Rowden, G.A.

    1982-01-01

    A process is provided for the recovery of uranium from an organic extractant phase containing an amine. The extractant phase is contacted in a number of mixing stages with an acidic aqueous stripping phase containing sulphate ions, and the phases are passed together through a series of mixing stages while maintaining a dispersion of droplets of one phase in the other. Uranium is precipitated from the final stage by raising the pH. An apparatus having several mixing chambers is described

  5. Uranium determination in water

    International Nuclear Information System (INIS)

    Prudenzo, E.J.; Puga, Maria J.; Cerchietti, Maria L.R.; Arguelles, Maria G.

    2005-01-01

    In our laboratory, a procedure has been assessed to determine uranium content of water in normal situations. The method proposed without sample pre-treatment, is simple and rapid. Uranium mass is measured by fluorimetry. For calculation of detection limit (Ld) and quantification level (Lq) we used blank samples and the results were analyzed for different statistical test. The calculation of total propagated uncertainty and sources contribution on real samples are presented. (author)

  6. Uranium project. Geochemistry prospection

    International Nuclear Information System (INIS)

    Lambert, J.

    1983-01-01

    Geochemistry studies the distribution of the chemicals elements in the terrestrial crust and its ways to migrate. The terminology used in this report is the following one: 1) Principles of the prospection geochemistry 2) Stages of the prospection geochemistry 3)utility of the prospection geochemistry 4) geochemistry of uranium 5) procedures used within the framework of uranium project 6) Average available 7) Selection of the zones of prospection geochemistry 8) Stages of the prospection, Sample preparation and analisis 9) Presentation of the results

  7. Worldwide developments in uranium

    International Nuclear Information System (INIS)

    Hoellen, E.E.

    1987-01-01

    World uranium production will continue to change in most major producing nations. Canadian production will increase and will be increasingly dominated by western producers as eastern Canadian high-cost production declines. Australian production will increase as major projects come into operation before 2000. US production will stabilize through the end of the century. South African production will be dependent upon the worldwide support for economic sanctions. China's entry into the world market injects yet another variable into the already cloudy supply picture. Many risks and uncertainties will face uranium producers through the 1980s. Recognizing that the uranium industry is not a fast-growing market, many existing and potential producers are seeking alternate investment courses, causing a restructuring of the world uranium production industry in ways not anticipated even a few years ago. During the restructuring process, world uranium production will most likely continue to exceed uranium consumption, resulting in a further buildup of world uranium inventories. Inventory sales will continue to redistribute this material. As inventory selling runs its course, users will turn to normal sources of supply, stimulating additional production to meet needs. Stimulation in the form of higher prices will be determined by how fast producers are willing and able to return to the market. Production costs are expected to have an increasing impact as it has become apparent that uranium resources are large in comparison to projected consumption. Conversely, security-of-supply issues have seemed to be of decreasing magnitude as Canada, Australia, and other non-US producers continue to meet delivery commitments

  8. U for uranium

    International Nuclear Information System (INIS)

    Anon.

    1982-01-01

    The Beisa Mine is unique in South Africa - it is the only underground mine with uranium as its main product and gold as a by-product. At the rate of 1,2 Mt/a, the life of Beisa is estimated on 26 years. Beisa's metallurgical plant is designed to handle initially a monthly throughput of 100 000t of ore, from which uranium, gold and silver will be extracted

  9. Uranium leads political stakes

    International Nuclear Information System (INIS)

    James, D.

    2009-01-01

    Until the announcement by the federal Environment Minister Peter Garrett that the government would permit uranium mining at Beverly Four Mile, South Australia, there had been little news flow from the sector over the past year. Uranium was the first to turn down, even before the United States sub-prime mortgage crisis began to cause shock waves through the global economy, a report by BGF Equities analyst Warwick Grigor shows.

  10. Uranium purchases report 1994

    International Nuclear Information System (INIS)

    1995-07-01

    US utilities are required to report to the Secretary of Energy annually the country of origin and the seller of any uranium or enriched uranium purchased or imported into the US, as well as the country of origin and seller of any enrichment services purchased by the utility. This report compiles these data and also contains a glossary of terms and additional purchase information covering average price and contract duration. 3 tabs

  11. Ranger uranium environmental enquiry

    International Nuclear Information System (INIS)

    1976-07-01

    The submission is divided into three sections. Section A considers the international implications of the development of uranium resources including economic and resource aspects and environmental and social aspects. Section B outlines the government's position on export controls over uranium and its effect on the introduction of nuclear power in Australia. Section C describes the licensing and regulatory functions that would be needed to monitor the environmental and health aspects of the Ranger project. (R.L.)

  12. Uranium - the nuclear fuel

    International Nuclear Information System (INIS)

    Smith, E.E.N.

    1976-01-01

    A brief history is presented of Canadian uranium exploration, production, and sales. Statistics show that Canada is a good customer for its own uranium due to a rapidly expanding nuclear power program. Due to an average 10 year lag between commencement of exploration and production, and with current producers sold out through 1985, it is imperative that exploration efforts be increased. (E.C.B.)

  13. Uranium tailings bibliography

    International Nuclear Information System (INIS)

    Holoway, C.F.; Goldsmith, W.A.; Eldridge, V.M.

    1975-12-01

    A bibliography containing 1,212 references is presented with its focus on the general problem of reducing human exposure to the radionuclides contained in the tailings from the milling of uranium ore. The references are divided into seven broad categories: uranium tailings pile (problems and perspectives), standards and philosophy, etiology of radiation effects, internal dosimetry and metabolism, environmental transport, background sources of tailings radionuclides, and large-area decontamination

  14. Canada's uranium industry

    International Nuclear Information System (INIS)

    Runnalls, O.J.C.

    The history of uranium mining, exploration, and trade in Canada is outlined. Past complex variations in market conditions are analyzed and future prospects for production and demand are forecast. Canadian and world resources of uranium are compared. Export policies and their rationales are explained. The Canadian nuclear fuel industry is described and fuel performance in-reactor is shown to be excellent. (E.C.B.)

  15. URANIUM SEPARATION PROCESS

    Science.gov (United States)

    Lyon, W.L.

    1962-04-17

    A method of separating uranium oxides from PuO/sub 2/, ThO/sub 2/, and other actinide oxides is described. The oxide mixture is suspended in a fused salt melt and a chlorinating agent such as chlorine gas or phosgene is sparged through the suspension. Uranium oxides are selectively chlorinated and dissolve in the melt, which may then be filtered to remove the unchlorinated oxides of the other actinides. (AEC)

  16. Fabrication of a tantalum-clad tungsten target for KENS

    International Nuclear Information System (INIS)

    Kawai, Masayoshi; Kikuchi, Kenji; Kurishita, Hiroaki; Li, J.-F.; Furusaka, Michihiro

    2001-01-01

    Since the cold neutron source intensity of KENS (the spallation neutron source at High Energy Accelerator Research Organization) was decreased into about a third of the designed value because a cadmium liner at the cold neutron source deformed and obstructed the neutron beam line, the target-moderator-and-reflector assembly (TMRA) has been replaced by a new one aimed at improving the neutron performance and recovering the cold neutron source. The tantalum target has also been replaced by a tantalum-clad tungsten one. In order to bond the tantalum-clad with the tungsten block, a hot isostatic press (HIP) process was applied and optimized. It was found that gaseous interstitial impurity elements severely attacked tantalum and embrittled, and that the getter materials such as zirconium and tantalum were effective to reduce the embrittlement

  17. Development of performance improvement of the cladding by ion beam

    International Nuclear Information System (INIS)

    Jung, Ki Sok; Kim, W.; Lee, J. H.; Choi, B. H.; Han, J. K.; Yoon, Y. H.; Lee, J. S.; Kwon, H. S.

    1997-07-01

    Ion implantation for the enhancement of the mechanical properties and the corrosion properties of the nuclear fuel cladding material has been conducted. Nitrogen implantations at 120 kV, > 5x10 17 ions/cm 2 , and >500 deg C resulted in the ZrN formation at the substrate surfaces. Nitrogen implantation at the elevated temperature in the oxygen atmosphere resulted in the great enhancement of the mechanical properties such as wear hardnesses. The hardness enhancement was especially significant up to 3 times at or above 500 deg C and 3x10 17 ions/cm 2 . The fretting wear resistance could be enhanced by 2 times at a dose of 8x10 17 ions/cm 2 . For the dedicated implanter of the cladding, main parts of the system were fabricated, e.g. ion sources, analyzing magnet, acceleration tube, vacuum lines, and the control programs. (author). 96 refs., 88 figs

  18. Modeling of realistic cladding structures for photonic bandgap fibers

    DEFF Research Database (Denmark)

    Mortensen, Niels Asger; Nielsen, Martin Dybendal

    2004-01-01

    Cladding structures of photonic bandgap fibers often have airholes of noncircular shape, and, typically, close-to-hexagonal airholes with curved corners are observed. We study photonic bandgaps in such structures by aid of a two-parameter representation of the size and curvature. For the fundamen......Cladding structures of photonic bandgap fibers often have airholes of noncircular shape, and, typically, close-to-hexagonal airholes with curved corners are observed. We study photonic bandgaps in such structures by aid of a two-parameter representation of the size and curvature....... For the fundamental bandgap we find that the bandgap edges (the intersections with the air line) shift toward shorter wavelengths when the air-filling fraction f is increased. The bandgap also broadens, and the relative bandwidth increases exponentially with f2. Compared with recent experiments [Nature 424, 657 (2003...

  19. Fracture resistance of irradiated stainless steel clad vessels

    International Nuclear Information System (INIS)

    McCabe, D.E.

    1988-01-01

    The surface crack embedded in the clad layer of a reactor vessel has been identified as a critical safety assessment condition relative to the pressurized thermal shock accident scenario. This project was initiated to determine the severity of such cracks experimentally, using irradiated material, and to identify the material property and stress conditions in the local region of the crack that are significant to the analysis. Bend bar tests provided the experimental simulation of the subject RPV surface crack. This report covers analysis techniques used and presents the findings indicated by the experimental results for irradiated and unirradiated materials. The irradiated clad specimens are the first of this type aimed at representing actual conditions in the wall of an irradiated pressure vessel. (author)

  20. Solution-mediated cladding doping of commercial polymer optical fibers

    Science.gov (United States)

    Stajanca, Pavol; Topolniak, Ievgeniia; Pötschke, Samuel; Krebber, Katerina

    2018-03-01

    Solution doping of commercial polymethyl methacrylate (PMMA) polymer optical fibers (POFs) is presented as a novel approach for preparation of custom cladding-doped POFs (CD-POFs). The presented method is based on a solution-mediated diffusion of dopant molecules into the fiber cladding upon soaking of POFs in a methanol-dopant solution. The method was tested on three different commercial POFs using Rhodamine B as a fluorescent dopant. The dynamics of the diffusion process was studied in order to optimize the doping procedure in terms of selection of the most suitable POF, doping time and conditions. Using the optimized procedure, longer segment of fluorescent CD-POF was prepared and its performance was characterized. Fiber's potential for sensing and illumination applications was demonstrated and discussed. The proposed method represents a simple and cheap way for fabrication of custom, short to medium length CD-POFs with various dopants.

  1. Pulsed Magnetic Welding for Advanced Core and Cladding Steel

    Energy Technology Data Exchange (ETDEWEB)

    Cao, Guoping [Univ. of Wisconsin, Madison, WI (United States); Yang, Yong [Univ. of Florida, Gainesville, FL (United States)

    2013-12-19

    To investigate a solid-state joining method, pulsed magnetic welding (PMW), for welding the advanced core and cladding steels to be used in Generation IV systems, with a specific application for fuel pin end-plug welding. As another alternative solid state welding technique, pulsed magnetic welding (PMW) has not been extensively explored on the advanced steels. The resultant weld can be free from microstructure defects (pores, non-metallic inclusions, segregation of alloying elements). More specifically, the following objectives are to be achieved: 1. To design a suitable welding apparatus fixture, and optimize welding parameters for repeatable and acceptable joining of the fuel pin end-plug. The welding will be evaluated using tensile tests for lap joint weldments and helium leak tests for the fuel pin end-plug; 2 Investigate the microstructural and mechanical properties changes in PMW weldments of proposed advanced core and cladding alloys; 3. Simulate the irradiation effects on the PWM weldments using ion irradiation.

  2. Modeling of Zircaloy cladding degradation under repository conditions

    International Nuclear Information System (INIS)

    Santanam, L.; Raghavan, S.; Chin, B.A.

    1989-07-01

    Two potential degradation mechanisms, creep and stress corrosion cracking, of Zircaloy cladding during repository storage of spent nuclear fuel have 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. A stress analysis of fuel rods has been performed. Stresses in the outer zirconium oxide layer and the inner Zircaloy tube have been predicted for typical internal pressurization, oxide layer thickness, volume expansion from formation of the oxide layer and thermal expansion coefficients of the cladding and oxide. Stress relaxation occurring in-reactor has also been taken into account. The calculations indicate that for the anticipated storage conditions investigated, the outer zirconium oxide layer is in a state of compression thus making it unlikely that stress corrosion cracking of the exterior surface will occur. 20 refs., 6 figs., 9 tabs

  3. Modeling of MOX Fuel Pellet-Clad Interaction Using ABAQUS

    International Nuclear Information System (INIS)

    Ambrosek, Richard G.; Pedersen, Robert C.; Maple, Amanda

    2002-01-01

    Post-irradiation examination (PIE) has indicated an increase in the outer diameter of fuel pins being irradiated in the Advanced Test Reactor (ATR) for the MOX irradiation program. The diameter increase is the largest in the region between fuel pellets. The fuel pellet was modeled using PATRAN and the model was evaluated using ABAQUS, version 6.2. The results from the analysis indicate the non-uniform clad diameter is caused by interaction between the fuel pellet and the clad. The results also demonstrate that the interaction is not uniform over the pellet axial length, with the largest interaction occurring in the region of the pellet-pellet interface. Results were obtained for an axisymmetric model and for a 1/8 pie shaped segment, using the coupled temperature-displacement solution technique. (authors)

  4. The characterization of activities associated with irradiated fuel element claddings

    International Nuclear Information System (INIS)

    Jenkins, I.L.; Bolus, D.J.; Glover, K.M.; Haynes, J.W.; Mapper, D.; Marwick, A.D.; Waterman, M.J.

    1982-01-01

    The object of the present work was to characterise the natures and amounts of the various α and βγ activities associated with cladding hulls. The claddings studied were stainless steel from a Fast Reactor and from an Advanced Gas Reactor and Zircaloy from a Boiling Water Reactor, from a Pressurized Water Reactor and from a Steam Generating Heavy Water Reactor. The hulls were examined by the following methods: alpha spectrometry to identify and quantify the α emitters and to estimate their depths of penetration, partial and complete dissolution of hulls followed by gross α counting, α spectrometry and γ spectrometry, fission track autoradiography to determine the distribution of fissile material associated with hulls, neutron activation to determine the total fissile content of the hulls, chemical separations followed by β counting and chemical treatment with various reagents to examine the ease of decontamination

  5. Chemical interaction at the FBR cladding fuel interfaces

    International Nuclear Information System (INIS)

    Delbrassine, A.; Retels, J.; Dirven, P.

    1978-01-01

    Pins containing UO 2 -30 wt.%PuO 2 and/or Caesium and/or Telluriom as doping elements have been irradiated for about 40 days in the BR2 reactor. The effects of two Cs/Te ratios, namely 1.3 and 4 and a wide range of O/M ratios on the inner corrosion of the clad have been investigated. The influence of Tellurium on the attack of the cladding has been pointed out. It may be responsible for the Chromium NS Nickel depletion in the grain boundaries of the steel. It is necessary to measure the effective Ts/Te ratio associated with the local corrosion layers. This local Cs/Te ratio should be more useful than the initial mean Cs/Te ratio in a pin for understanding the corrosion phenomene. (author)

  6. Pulsed Magnetic Welding for Advanced Core and Cladding Steel

    International Nuclear Information System (INIS)

    Cao, Guoping; Yang, Yong

    2013-01-01

    To investigate a solid-state joining method, pulsed magnetic welding (PMW), for welding the advanced core and cladding steels to be used in Generation IV systems, with a specific application for fuel pin end-plug welding. As another alternative solid state welding technique, pulsed magnetic welding (PMW) has not been extensively explored on the advanced steels. The resultant weld can be free from microstructure defects (pores, non-metallic inclusions, segregation of alloying elements). More specifically, the following objectives are to be achieved: 1. To design a suitable welding apparatus fixture, and optimize welding parameters for repeatable and acceptable joining of the fuel pin end-plug. The welding will be evaluated using tensile tests for lap joint weldments and helium leak tests for the fuel pin end-plug; 2 Investigate the microstructural and mechanical properties changes in PMW weldments of proposed advanced core and cladding alloys; 3. Simulate the irradiation effects on the PWM weldments using ion irradiation.

  7. The Streltsovskoye uranium district

    International Nuclear Information System (INIS)

    Ischukova, L.P.

    1997-01-01

    This paper describes the geology of the Streltsovskoye uranium district located in south-eastern Zabaikalie region, Chita Province, Siberia, Russia. This district hosts Russia's only currently active uranium production centre. The uranium ore was discovered from 1963 to 1967 by drilling below fluorite veins which had minor associated uranium mineralization and radioactive anomalies. The uranium occurs as large scale vein stockwork deposits of hydrothermal origin within a volcano-tectonic caldera formed by continental volcanism of Late Mesozoic age. Rocks occurring in the caldera include basalt and trachydacite, overlain by rhyolite, and with associated interbedded sediments. The ore bodies occur in steeply dipping faults, with the greatest concentrations located where faults along the margins of the caldera intersect steeply dipping, cross cutting, northeasterly and northwesterly striking faults. The Streltsovskoye caldera extends over an area of 150 km 2 and is underlain by a large batholith. The 19 identified uranium deposits occurred in structural features that cut through the caldera sequence and extend into the basement rocks. The caldera has a maximum thickness of 1400 metres. Details of several deposits are given, including descriptions of mineralization and associated alteration. (author). 10 figs

  8. Uranium deposits through time

    International Nuclear Information System (INIS)

    Derry, D.R.

    1980-01-01

    The distribution of different types of uranium deposits through geological time has been noted by various authors, especially as affecting deposits formed in the Precambrian era. While this is significant, it may have been oversimplified by grouping deposits within geological time limits that are very extensive when compared with Phanerozoic groupings. As more detailed information becomes available, we find that some of the higher-grade deposits have complicated histories involving several stages of concentration separated by long periods of time. A comparison of the proportion of uranium production and reserves contributed by individual classes (and geological ages) of deposits shows a changing trend in favour of the unconformity class. Possibly more attention should be paid to metallographic uranium areas and their localizations. Some areas, e.g. Gabon, West Africa and Baker Lake, N.W.T., have more than one type of uranium deposit. A further consideration is that much of the known uranium reserves in the non-Communist world lies within a dozen areas, each ranging between 3000 and 130 000 km 2 . It is possible that such uranium metallographic areas were localized by very ancient fracture systems, now not easily recognized, in the original Earth's crust

  9. Production of uranium peroxide

    International Nuclear Information System (INIS)

    Caropreso, F.E.; Kreuz, D.F.

    1977-01-01

    A process is claimed of recovering uranium values as uranium peroxide from an aqueous uranyl solution containing dissolved vanadium and sodium impurities by treating the uranyl solution with hydrogen peroxide in an amount sufficient to have an excess of at least 0.5 parts H 2 O 2 per part of vanadium (V 2 O 5 ) above the stoichiometric amount required to form the uranium peroxide, the hydrogen peroxide treatment is carried out in three sequential phases consisting of I, a precipitation phase in which the hydrogen peroxide is added to the uranyl solution to precipitate the uranium peroxide and the pH of the reaction medium maintained in the range of 2.5 to 5.5 for a period of from about 1 to 60 minutes after the hydrogen peroxide addition; II, a digestion phase in which the pH of the reaction medium is maintained in the range of 3.0 to 7.0 for a period of about 5 to 180 minutes and III, a final phase in which the pH of the reaction medium is maintained in the range of 4.0 to 7.0 for a period of about 1 to 60 minutes during which time the uranium peroxide is separated from the reaction solution containing the dissolved vanadium and sodium impurities. The excess hydrogen peroxide is maintained during the entire treatment up until the uranium peroxide is separated from the reaction medium

  10. Analytical method of uranium (IV) and uranium (VI) in uranium ores and uranium-bearing rocks

    International Nuclear Information System (INIS)

    Shen Zhuqin; Zheng Yongfeng; Li Qingzhen; Zhong Miaolan; Gu Dingxiang

    1995-11-01

    The best conditions for keeping the original valences of uranium during the dissolution and separation procedure of geological samples (especially those micro uranium-bearing rock) were studied. With the exist of high concentration protectants, the sample was decomposed with concentration HF at 40 +- 5 degree C. The U(VI) was dissolved completely and formed stable complex UO 2 F 2 , the U(IV) was precipitated rapidly and carried by carrier. Quantitative separation was carried out immediately with suction. The decomposition of sample and separation of solid/liquid phases was completed within two minutes. After separation, the U(IV) and U(VI) were determined quantitatively with laser fluorescence or voltametry respectively according to the uranium content. The limit of detection for this method is 0.7 μg/g, RSD is 10.5%, the determinate range of uranium is 2 x 10 -6 ∼10 -1 g/g. The uranium contents and their valence state ratio were measured for more than one hundred samples of sand stone and granite, the accuracy and precision of these results are satisfactory for uranium geological research. (12 tabs.; 11 refs.)

  11. Automated uranium titration system

    International Nuclear Information System (INIS)

    Takahashi, M.; Kato, Y.

    1983-01-01

    An automated titration system based on the Davies-Gray method has been developed for accurate determination of uranium. The system consists of a potentiometric titrator with precise burettes, a sample changer, an electronic balance and a desk-top computer with a printer. Fifty-five titration vessels are loaded in the sample changer. The first three contain the standard solution for standardizing potassium dichromate titrant, and the next two and the last two contain the control samples for data quality assurance. The other forty-eight measurements are carried out for sixteen unknown samples. Sample solution containing about 100 mg uranium is taken in a titration vessel. At the pretreatment position, uranium (VI) is reduced to uranium (IV) by iron (II). After the valency adjustment, the vessel is transferred to the titration position. The rate of titrant addition is automatically controlled to be slower near the end-point. The last figure (0.01 mL) of the equivalent titrant volume for uranium is calculated from the potential change. The results obtained with this system on 100 mg uranium gave a precision of 0.2% (RSD,n=3) and an accuracy of better than 0.1%. Fifty-five titrations are accomplished in 10 hours. (author)

  12. Uranium-scintillator device

    International Nuclear Information System (INIS)

    Smith, S.D.

    1979-01-01

    The calorimeter subgroup of the 1977 ISABELLE Summer Workshop strongly recommended investigation of the uranium-scintillator device because of its several attractive features: (1) increased resolution for hadronic energy, (2) fast time response, (3) high density (i.e., 16 cm of calorimeter per interaction length), and, in comparison with uranium--liquid argon detectors, (4) ease of construction, (5) simple electronics, and (6) lower cost. The AFM group at the CERN ISR became interested in such a calorimeter for substantially the same reasons, and in the fall of 1977 carried out tests on a uranium-scintillator (U-Sc) calorimeter with the same uranium plates used in their 1974 studies of the uranium--liquid argon (U-LA) calorimeter. The chief disadvantage of the scintillator test was that the uranium plates were too small to fully contain the hadronic showers. However, since the scintillator and liquid argon tests were made with the plates, direct comparison of the two types of devices could be made

  13. Uranium control in phosphogypsum

    International Nuclear Information System (INIS)

    Hurst, F.J.; Arnold, W.D.

    1980-01-01

    In wet-process phosphoric acid plants, both previous and recent test results show that uranium dissolution from phosphate rock is significantly higher when the rock is acidulated under oxidizing conditions than under reducing conditions. Excess sulfate and excess fluoride further enhance the distribution of uranium to the cake. Apparently the U(IV) present in the crystal lattice of the apatite plus that formed by reduction of U(IV) by FE(II) during acidulation is trapped or carried into the crystal lattice of the calcium sulfate crystals as they form and grow. The amount of uranium that distributes to hemihydrate filter cake is up to seven times higher than the amount that distributes to the dihydrate cake. About 60% of the uranium in hemihydrate cakes can be readily leached after hydration of the cake, but the residual uranium (20 to 30%) is very difficult to remove economically. Much additional research is needed to develop methods for minimizing uranium losses to calcium filter cakes

  14. Multiresponse Optimization of Laser Cladding Steel + VC Using Grey Relational Analysis in the Taguchi Method

    Science.gov (United States)

    Zhang, Zhe; Kovacevic, Radovan

    2016-07-01

    Laser cladding of metal matrix composite coatings (MMCs) has become an effective and economic method to improve the wear resistance of mechanical components. The clad quality characteristics such as clad height, carbide fraction, carbide dissolution, and matrix hardness in MMCs determine the wear resistance of the coatings. These clad quality characteristics are influenced greatly by the laser cladding processing parameters. In this study, American Iron and Steel Institute (AISI) 420 + 20% vanadium carbide (VC) was deposited on mild steel with a high powder direct diode laser. The Taguchi-based Grey relational method was used to optimize the laser cladding processing parameters (laser power, scanning speed, and powder feed rate) with the consideration of multiple clad characteristics related to wear resistance (clad height, carbide volume fraction, and Fe-matrix hardness). A Taguchi L9 orthogonal array was designed to study the effects of processing parameters on each response. The contribution and significance of each processing parameter on each clad characteristic were investigated by the analysis of variance (ANOVA). The Grey relational grade acquired from Grey relational analysis was used as the performance characteristic to obtain the optimal combination of processing parameters. Based on the optimal processing parameters, the phases and microstructure of the laser-cladded coating were characterized by using x-ray diffraction (XRD) and scanning electron microscopy (SEM) with energy-dispersive spectroscopy (EDS).

  15. Advances in the manufacture of clad tubes and components for PHWR fuel bundle

    International Nuclear Information System (INIS)

    Saibaba, N.; Jha, S.K.; Chandrasekha, B.; Tonpe, S.; Jayaraj, R.N.

    2010-01-01

    Fuel bundles for Pressurized Heavy Water Reactors (PHWRs) consists of Uranium di-oxide pellets encapsulated into thin wall Zircaloy clad tubes. Other components such as end caps, bearing pads and spacer pads are the integral elements of the fuel bundle. As the fuel assembly is subjected to severe operating conditions of high temperature and pressure in addition to continual irradiation exposure, all the components are manufactured conforming to stringent specifications with respect to chemical composition, mechanical & metallurgical properties and dimensional tolerances. The integrity of each component is ensured by NDE at different stages of manufacture. The manufacturing route for fuel tubes and components comprise of a combination of thermomechanical processing and each process step has marked effect on the final properties. The fuel tubes are manufactured by processing the extruded blanks in four stage cold pilgering with intermediate annealing and final stress relieving operation. The bar material is produced by hot extrusion followed by multi-pass swaging and intermediate annealing. Spacer pads and bearing pads are manufactured by blanking and coining of Zircaloy sheet which is made by a combination of hot and cold rolling operations. Due to the small size and stringent dimensional requirements of these appendages, selection of production route and optimization of process parameters are important. This paper discusses about various measures taken for improving the recoveries and mechanical and corrosion properties of the tube, sheet and bar materials being manufactured at Nuclear Fuel Complex, Hyderabad For the production of clad tubes, modifications at extrusion stage to reduce the wall thickness variation, introduction of ultrasonic testing of extruded blanks, optimization of cold working and heat treatment parameters at various stages of production etc. were done. The finished bar material is subjected to 100% Ultrasonic and eddy current testing to ensure

  16. Advances in the manufacture of clad tubes and components for PHWR fuel bundle

    Energy Technology Data Exchange (ETDEWEB)

    Saibaba, N.; Jha, S.K.; Chandrasekha, B.; Tonpe, S.; Jayaraj, R.N. [Nuclear Fuel Complex, Hyderabad (India)

    2010-07-01

    Fuel bundles for Pressurized Heavy Water Reactors (PHWRs) consists of Uranium di-oxide pellets encapsulated into thin wall Zircaloy clad tubes. Other components such as end caps, bearing pads and spacer pads are the integral elements of the fuel bundle. As the fuel assembly is subjected to severe operating conditions of high temperature and pressure in addition to continual irradiation exposure, all the components are manufactured conforming to stringent specifications with respect to chemical composition, mechanical & metallurgical properties and dimensional tolerances. The integrity of each component is ensured by NDE at different stages of manufacture. The manufacturing route for fuel tubes and components comprise of a combination of thermomechanical processing and each process step has marked effect on the final properties. The fuel tubes are manufactured by processing the extruded blanks in four stage cold pilgering with intermediate annealing and final stress relieving operation. The bar material is produced by hot extrusion followed by multi-pass swaging and intermediate annealing. Spacer pads and bearing pads are manufactured by blanking and coining of Zircaloy sheet which is made by a combination of hot and cold rolling operations. Due to the small size and stringent dimensional requirements of these appendages, selection of production route and optimization of process parameters are important. This paper discusses about various measures taken for improving the recoveries and mechanical and corrosion properties of the tube, sheet and bar materials being manufactured at Nuclear Fuel Complex, Hyderabad For the production of clad tubes, modifications at extrusion stage to reduce the wall thickness variation, introduction of ultrasonic testing of extruded blanks, optimization of cold working and heat treatment parameters at various stages of production etc. were done. The finished bar material is subjected to 100% Ultrasonic and eddy current testing to ensure

  17. Radiation hardness of new Kuraray double cladded optical fibers

    International Nuclear Information System (INIS)

    Bedeschi, F.; Menzione, A.; Budagov, Yu.; Chirikov-Zorin, I.; Solov'ev, A.; Turchanovich, L.; Vasil'chenko, V.

    1996-01-01

    The radiation hardness of the new plastic scintillating and clear fibers irradiated by 137 Cs γ-flux and by pulsed reactor fast neutrons were investigated. All the studied fibers were of S-type (with S=70) and had a double cladding. Optical fibers degradation study after irradiation shows that the level of radiation hardness lower that what is expected from results of previous studies. 9 refs., 6 figs

  18. Interactions of Zircaloy cladding with gallium: 1998 midyear status

    International Nuclear Information System (INIS)

    Wilson, D.F.; DiStefano, J.R.; Strizak, J.P.; King, J.F.; Manneschmidt, E.T.

    1998-06-01

    A program has been implemented to evaluate the effect of gallium in mixed-oxide (MOX) fuel derived from weapons-grade (WG) plutonium on Zircaloy cladding performance. The objective is to demonstrate that low levels of gallium will not compromise the performance of the MOX fuel system in a light-water reactor. The graded, four-phase experimental program was designed to evaluate the performance of prototypic Zircaloy cladding materials against (1) liquid gallium (Phase 1), (2) various concentrations of Ga 2 O 3 (Phase 2), (3) centrally heated surrogate fuel pellets with expected levels of gallium (Phase 3), and (4) centrally heated prototypic MOX fuel pellets (Phase 4). This status report describes the results of a series of tests for Phases 1 and 2. Three types of tests are being performed: (1) corrosion, (2) liquid metal embrittlement, and (3) corrosion-mechanical. These tests will determine corrosion mechanisms, thresholds for temperature and concentration of gallium that may delineate behavioral regimes, and changes in the mechanical properties of Zircaloy. Initial results have generally been favorable for the use of WG-MOX fuel. The MOX fuel cladding, Zircaloy, does react with gallium to form intermetallic compounds at ≥300 C; however, this reaction is limited by the mass of gallium and is therefore not expected to be significant with a low level (parts per million) of gallium in the MOX fuel. Although continued migration of gallium into the initially formed intermetallic compound can result in large stresses that may lead to distortion, this was shown to be extremely unlikely because of the low mass of gallium or gallium oxide present and expected clad temperatures below 400 C. Furthermore, no evidence for grain boundary penetration by gallium has been observed

  19. In-situ crack repair by laser cladding

    CSIR Research Space (South Africa)

    Van Rooyen, C

    2010-09-01

    Full Text Available Hammer peening was introduced in order to mechanically seal of squirting water prior to crack sealing. Leaks were reduced from squirting to oozing. During cladding, water started squirting out again resulting in pore formation in some areas. Hammer... is shown in Figure 6. Three small leaking pores were observed in the first layer, indicated by arrows. Hammer peening was applied to the first layer to mechanically seal the leaking pores prior to deposition of the second sealing layer. Successful...

  20. Recovery of uranium by chlorination

    International Nuclear Information System (INIS)

    Komoto, Shigetoshi; Taki, Tomihiro

    1988-01-01

    The recovery of uranium from uraniferous phosphate by conventional process is generally uneconomic, except that uranium is recovered as a by-product. If an economical process by which uranium is recovered efficiently as a chief product is discovered, uraniferous phosphate will be used effectively as uranium ore. By using chiorination which will be expected to be favorable in comparison with conventional process, the recovery of uranium from uraniferous phosphate has been carried out. The paper describes the reaction machanism and general characteristics of the uranium chiorination, and the research done so for. (author)

  1. Microbial biofilm growth on irradiated, spent nuclear fuel cladding

    Science.gov (United States)

    Bruhn, D. F.; Frank, S. M.; Roberto, F. F.; Pinhero, P. J.; Johnson, S. G.

    2009-02-01

    A fundamental criticism regarding the potential for microbial influenced corrosion in spent nuclear fuel cladding or storage containers concerns whether the required microorganisms can, in fact, survive radiation fields inherent in these materials. This study was performed to unequivocally answer this critique by addressing the potential for biofilm formation, the precursor to microbial-influenced corrosion, in radiation fields representative of spent nuclear fuel storage environments. This study involved the formation of a microbial biofilm on irradiated spent nuclear fuel cladding within a hot cell environment. This was accomplished by introducing 22 species of bacteria, in nutrient-rich media, to test vessels containing irradiated cladding sections and that was then surrounded by radioactive source material. The overall dose rate exceeded 2 Gy/h gamma/beta radiation with the total dose received by some of the bacteria reaching 5 × 10 3 Gy. This study provides evidence for the formation of biofilms on spent-fuel materials, and the implication of microbial influenced corrosion in the storage and permanent deposition of spent nuclear fuel in repository environments.

  2. Processing of irradiated, enriched uranium fuels at the Savannah River Plant

    International Nuclear Information System (INIS)

    Hyder, M.L.; Perkins, W.C.; Thompson, M.C.; Burney, G.A.; Russell, E.R.; Holcomb, H.P.; Landon, L.F.

    1979-04-01

    Uranium fuels containing 235 U at enrichments from 1.1% to 94% are processed and recovered, along with neptunium and plutonium byproducts. The fuels to be processed are dissolved in nitric acid. Aluminum-clad fuels are disssolved using a mercury catalyst to give a solution rich in aluminum. Fuels clad in more resistant materials are dissolved in an electrolytic dissolver. The resulting solutions are subjected to head-end treatment, including clarification and adjustment of acid and uranium concentration before being fed to solvent extraction. Uranium, neptunium, and plutonium are separated from fission products and from one another by multistage countercurrent solvent extraction with dilute tri-n-butyl phosphate in kerosene. Nitric acid is used as the salting agent in addition to aluminum or other metal nitrates present in the feed solution. Nuclear safety is maintained through conservative process design and the use of monitoring devices as secondary controls. The enriched uranium is recovered as a dilute solution and shipped off-site for further processing. Neptunium is concentrated and sent to HB-Line for recovery from solution. The relatively small quantities of plutonium present are normally discarded in aqueous waste, unless the content of 238 Pu is high enough to make its recovery desirable. Most of the 238 Pu can be recovered by batch extraction of the waste solution, purified by counter-current solvent extraction, and converted to oxide in HB-Line. By modifying the flowsheet, 239 Pu can be recovered from low-enriched uranium in the extraction cycle; neptunium is then not recovered. The solvent is subjected to an alkaline wash before reuse to remove degraded solvent and fission products. The aqueous waste is concentrated and partially deacidified by evaporation before being neutralized and sent to the waste tanks; nitric acid from the overheads is recovered for reuse

  3. Processing of irradiated, enriched uranium fuels at the Savannah River Plant

    Energy Technology Data Exchange (ETDEWEB)

    Hyder, M L; Perkins, W C; Thompson, M C; Burney, G A; Russell, E R; Holcomb, H P; Landon, L F

    1979-04-01

    Uranium fuels containing /sup 235/U at enrichments from 1.1% to 94% are processed and recovered, along with neptunium and plutonium byproducts. The fuels to be processed are dissolved in nitric acid. Aluminum-clad fuels are disssolved using a mercury catalyst to give a solution rich in aluminum. Fuels clad in more resistant materials are dissolved in an electrolytic dissolver. The resulting solutions are subjected to head-end treatment, including clarification and adjustment of acid and uranium concentration before being fed to solvent extraction. Uranium, neptunium, and plutonium are separated from fission products and from one another by multistage countercurrent solvent extraction with dilute tri-n-butyl phosphate in kerosene. Nitric acid is used as the salting agent in addition to aluminum or other metal nitrates present in the feed solution. Nuclear safety is maintained through conservative process design and the use of monitoring devices as secondary controls. The enriched uranium is recovered as a dilute solution and shipped off-site for further processing. Neptunium is concentrated and sent to HB-Line for recovery from solution. The relatively small quantities of plutonium present are normally discarded in aqueous waste, unless the content of /sup 238/Pu is high enough to make its recovery desirable. Most of the /sup 238/Pu can be recovered by batch extraction of the waste solution, purified by counter-current solvent extraction, and converted to oxide in HB-Line. By modifying the flowsheet, /sup 239/Pu can be recovered from low-enriched uranium in the extraction cycle; neptunium is then not recovered. The solvent is subjected to an alkaline wash before reuse to remove degraded solvent and fission products. The aqueous waste is concentrated and partially deacidified by evaporation before being neutralized and sent to the waste tanks; nitric acid from the overheads is recovered for reuse.

  4. Final report on accident tolerant fuel performance analysis of APMT-Steel Clad/UO₂ fuel and APMT-Steel Clad/UN-U₃Si₅ fuel concepts

    Energy Technology Data Exchange (ETDEWEB)

    Unal, Cetin [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Galloway, Jack D. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2014-09-12

    In FY2014 our group completed and documented analysis of new Accident Tolerant Fuel (ATF) concepts using BISON. We have modeled the viability of moving from Zircaloy to stainless steel cladding in traditional light water reactors (LWRs). We have explored the reactivity penalty of this change using the MCNP-based burnup code Monteburns, while attempting to minimize this penalty by increasing the fuel pellet radius and decreasing the cladding thickness. Fuel performance simulations using BISON have also been performed to quantify changes to structural integrity resulting from thinner stainless steel claddings. We account for thermal and irradiation creep, fission gas swelling, thermal swelling and fuel relocation in the models for both Zircaloy and stainless steel claddings. Additional models that account for the lower oxidation stainless steel APMT are also invoked where available. Irradiation data for HT9 is used as a fallback in the absence of appropriate models. In this study the isotopic vectors within each natural element are varied to assess potential reactivity gains if advanced enrichment capabilities were levied towards cladding technologies. Recommendations on cladding thicknesses for a robust cladding as well as the constitutive components of a less penalizing composition are provided. In the first section (section 1-3), we present results accepted for publication in the 2014 TOPFUEL conference regarding the APMT/UO₂ ATF concept (J. Galloway & C. Unal, Accident Tolerant and Neutronically Favorable LWR Cladding, Proceedings of WRFPM 2014, Sendai, Japan, Paper No.1000050). Next we discuss our preliminary findings from the thermo-mechanical analysis of UN-U₃Si₅ fuel with APMT clad. In this analysis we used models developed from limited data that need to be updated when the irradiation data from ATF-1 test is available. Initial results indicate a swelling rate less than 1.5% is needed to prevent excessive clad stress.

  5. Process for electrolytically preparing uranium metal

    Science.gov (United States)

    Haas, Paul A.

    1989-01-01

    A process for making uranium metal from uranium oxide by first fluorinating uranium oxide to form uranium tetrafluoride and next electrolytically reducing the uranium tetrafluoride with a carbon anode to form uranium metal and CF.sub.4. The CF.sub.4 is reused in the fluorination reaction rather than being disposed of as a hazardous waste.

  6. Cladding of Cr- Modified NiAl Coating on 310 Stainless Steel Weld Cladding and Evaluation of its Wear Behavior.

    Directory of Open Access Journals (Sweden)

    S. Pourmohamadi

    2017-02-01

    Full Text Available In this study, a Cr-modified NiAl coating was fabricated by weld cladding technique using Gas- Tungsten Arc Welding (GTAW process on 310 steel. Chemical composition and microstructure of the coating was studied by X-Ray Diffraction (XRD, optical microscopy and scanning electron microscopy equipped with an Energy Dispersive Spectroscopy (EDS. The wear behavior of the coated steel was examined through pin-on-disc tests at ambient temperature and 400 °C. The results showed that the hardness of coated steel increased remarkably due to the formation of Cr-modified NiAl on the surface. Furthermore, the wear experiments showed that the presence of Cr-modified NiAl coating caused significant improvement in wear resistance of cladding 310 steel at both ambient temperature and 400 °C. These results were discussed based on the wear mechanism obtained from examination of the worn surfaces using SEM.

  7. Oxidation of uranium and uranium alloys

    International Nuclear Information System (INIS)

    Orman, S.

    1976-01-01

    The corrosion behaviour of uranium in oxygen, water and water + oxygen mixtures is compared and contrasted. A considerable amount of work, much of it conflicting, has been published on the U + H 2 O and U + H 2 O + O 2 systems. An attempt has been made to summarise this data and to explain the reasons for the lack of agreement between the experimental results. The evidence for the mechanism involving OH - ion diffusion as the reacting entity in both the U + H 2 O and U + O 2 + H 2 O reactions is advanced. The more limited corrosion data on some lean uranium alloys and on some higher addition alloys referred to as stainless materials is summarised together with some previously unreported results obtained with these materials at AWRE. The data indicates that in the absence of oxygen the lean alloys behave in a similar manner to uranium and evolve hydrogen in approximately theoretical quantities. But the stainless alloys absorb most of the product hydrogen and assessments of reactivity based on hydrogen evolution would be very inaccurate. The direction that future corrosion work on these materials should take is recommended

  8. Australia's uranium export potential

    International Nuclear Information System (INIS)

    Mosher, D.V.

    1981-01-01

    During the period 1954-71 in Australia approximately 9000 MT of U 3 O 8 was produced from five separate localities. Of this, 7000 MT was exported to the United Kingdom and United States and the balance stockpiled by the Australian Atomic Energy Commission (AAEC). Australia's uranium ore reserves occur in eight deposits in three states and the Northern Territory. However, 83% of Australia's reserves are contained in four deposits in lower Proterozoic rocks in the East Alligator River region of the Northern Territory. The AAEC has calculated Australia's recoverable uranium reserves by eliminating estimated losses during the mining and milling of the ores. AAEC has estimated reasonably assured resources of 289,000 MT of uranium at a recovery cost of less than US$80 per kilogram uranium. The companies have collectively announced a larger ore reserve than the Australian Atomic Energy Commission. This difference is a result of the companies adopting different ore reserve categories. On August 25, 1977, the federal government announced that Australia would develop its uranium resources subject to stringent environmental controls, recognition of Aboriginal Land Rights, and international safeguards. Australian uranium production should gradually increase from 1981 onward, growing to 10,000 to 15,000 MT by 1985-86. Further increases in capacity may emerge during the second half of the 1980s when expansion plans are implemented. Exploration for uranium has not been intensive due to delays in developing the existing deposits. It is likely that present reserves can be substantially upgraded if more exploration is carried out. 6 figures, 3 tables

  9. Fuel-cladding mechanical interaction effects in fast reactor mixed oxide fuel

    International Nuclear Information System (INIS)

    Boltax, A.; Biancheria, A.

    1977-01-01

    Thermal and fast reactor irradiation experiments on mixed oxide fuel pins under steady-state and power change conditions reveal evidence for significant fuel-cladding mechanical interaction (FCMI) effects. Analytical studies with the LIFE-III fuel performance code indicate that high cladding stresses can be produced by general and local FCMI effects. Also, evidence is presented to show that local cladding strains can be caused by the accumulation of cesium at the fuel-cladding interface. Although it is apparent that steady-state FCMI effects have not given rise to cladding breaches in current fast reactors, it is anticipated that FCMI may become more important in the future because of interest in: higher fuel burnups; increased power ramp rates; load follow operation; and low swelling cladding alloys. (author)

  10. Tailoring nonlinearity and dispersion of photonic crystal fibers using hybrid cladding

    International Nuclear Information System (INIS)

    Zhao-lun, Liu; Lan-tian, Hou; Wei, Wang

    2009-01-01

    We present a hybrid cladding photonic crystal fiber for shaping high nonlinear and flattened dispersion in a wide range of wavelengths. The new structure adopts hybrid cladding with different pitches, air-holes diameters and air-holes arrayed fashions. The full-vector finite element method with perfectly matched layer is used to investigate the characteristics of the hybrid cladding photonic crystal fiber such as nonlinearity and dispersion properties. The influence of the cladding structure parameters on the nonlinear coefficient and geometric dispersion is analyzed. High nonlinear coefficient and the dispersion properties of fibers are tailored by adjusting the cladding structure parameters. A novel hybrid cladding photonic crystal fiber with high nonlinear coefficient and dispersion flattened which is suited for super continuum generation is designed. (author)

  11. On LMFBR corrosion. Part II: Consideration of the in-reactor fuel-cladding system

    International Nuclear Information System (INIS)

    Bradbury, M.H.; Pickering, S.; Walker, C.T.; Whitlow, W.H.

    1976-05-01

    The scientific and technological aspects of LMFBR cladding corrosion are discussed in detail. Emphasis is placed on the influence of the irradiation environment and the effect of fuel and filler-gas impurities on the corrosion process. These studies are complemented by a concise review of out-of-pile simulation experiments that endeavour to clarify the role of the aggressive fission products cesium, tellurium and iodine. The principal models for cladding corrosion are presented and critically assessed. Areas of uncertainty are exposed and some pertinent experiments are suggested. Consideration is also given to some new observations regarding the role of stress in fuel-cladding reactions and the formation of ferrite in the corrosion zone of the cladding during irradiation. Finally, two technological solutions to the problem of cladding corrosion are proposed. These are based on the use of an oxygen buffer in the fuel and the application of a protective coating to the inner surface of the cladding

  12. Caramel, uranium oxide fuel plates for water cooled reactors

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  13. Pyrophoric behaviour of uranium hydride and uranium powders

    Science.gov (United States)

    Le Guyadec, F.; Génin, X.; Bayle, J. P.; Dugne, O.; Duhart-Barone, A.; Ablitzer, C.

    2010-01-01

    Thermal stability and spontaneous ignition conditions of uranium hydride and uranium metal fine powders have been studied and observed in an original and dedicated experimental device placed inside a glove box under flowing pure argon. Pure uranium hydride powder with low amount of oxide (Oxidation mechanisms are proposed.

  14. Uranium Supply Strategy of China

    International Nuclear Information System (INIS)

    Gao Shangxiong; Zhang Decun; Zhang Yi

    2014-01-01

    Conclusion: With the rapid development of nuclear power in the next few years, uranium demand will increase accordingly. Overseas uranium development will be the major channel to meet the future requirement of NPP demand in China.

  15. Collect method of uranium hexafluoride

    International Nuclear Information System (INIS)

    Moura, S.C.; Bustillos, O.W.V.

    1991-01-01

    A collect method of uranium hexafluoride was designed, constructed and assembled in Analytical Laboratory from Instituto de Energia Atomica, Sao Paulo, Brazil. This method of collect is main for quality control of uranium hexafluoride. (author)

  16. The case against uranium mining

    International Nuclear Information System (INIS)

    Robotham, F.P.

    1980-01-01

    Australia is a potential uranium supplier. The case against uranium mining is presented. Biological effects of radiation, risks involved in reactor operation and the problems of waste disposal are discussed

  17. Reflood behavior at low initial clad temperature in Slab Core Test Facility Core-II

    International Nuclear Information System (INIS)

    Akimoto, Hajime; Sobajima, Makoto; Abe, Yutaka; Iwamura, Takamichi; Ohnuki, Akira; Okubo, Tsutomu; Murao, Yoshio; Okabe, Kazuharu; Adachi, Hiromichi.

    1990-07-01

    In order to study the reflood behavior with low initial clad temperature, a reflood test was performed using the Slab Core Test Facility (SCTF) with initial clad temperature of 573 K. The test conditions of the test are identical with those of SCTF base case test S2-SH1 (initial clad temperature 1073 K) except the initial clad temperature. Through the comparison of results from these two tests, the following conclusions were obtained. (1) The low initial clad temperature resulted in the low differential pressures through the primary loops due to smaller steam generation in the core. (2) The low initial clad temperature caused the accumulated mass in the core to be increased and the accumulated mass in the downcomer to be decreased in the period of the lower plenum injection with accumulator (before 50s). In the later period of the cold leg injection with LPCI (after 100s), the water accumulation rates in the core and the downcomer were almost the same between both tests. (3) The low initial clad temperature resulted in the increase of the core inlet mass flow rate in the lower plenum injection period. However, the core inlet mass flow rate was almost the same regardless of the initial clad temperature in the later period of the cold leg injection period. (4) The low initial clad temperature resulted in the low turnaround temperature, high temperature rise and fast bottom quench front propagation. (5) In the region apart from the quench front, low initial clad temperature resulted in the lower heat transfer. In the region near the quench front, almost the same heat transfer coefficient was observed between both tests. (6) No flow oscillation with a long period was observed in the SCTF test with low initial clad temperature of 573 K, while it was remarkable in the Cylindrical Core Test Facility (CCTF) test which was performed with the same initial clad temperature. (J.P.N.)

  18. Solid-Core Photonic Bandgap Fibers for Cladding-Pumped Raman Amplification

    Science.gov (United States)

    2011-06-03

    that the cladding is uniformly-pumped at its maximum numerical aperture and that the photonic band gap ( PBG ) structure comprised of a triangular array...published 3 Jun 2011 (C) 2011 OSA 6 June 2011 / Vol. 19, No. 12 / OPTICS EXPRESS 11855 beyond the PBG structure. In this case, the high pump cladding...exhibit lower loss [16]. For high order bandgaps to exist in the cladding, the inclusions forming the PBG structure must support a large number of guided

  19. Development of Preliminary HT9 Cladding Tube for Sodium-cooled Fast Reactor (SFR)

    International Nuclear Information System (INIS)

    Kim, Jun Hwan; Baek, Jong Hyuk; Heo, Hyeong Min; Park, Sang Gyu; Kim, Sung Ho; Lee, Chan Bock

    2013-01-01

    To achieve manufacturing technology of the fuel cladding tube in order to keep pace with the predetermined schedule in developing SFR fuel, KAERI has launched in developing fuel cladding tube in cooperation with a domestic steelmaking company. After fabricating medium-sized 1.1 ton HT9 ingot, followed by the multiple processes of hot and cold working, preliminary samples of HT9 seamless cladding tube having 7.4mm in outer diameter, 0.56mm in thickness, and 3m in length were fabricated. The objective of this study is to summarize the brief development status of the HT9 cladding tubes. Mechanical properties like axial tension, biaxial burst, pressurized creep and sodium compatibility of the cladding tubes were carried out to set up the performance evaluation technology to test the prototype FMS cladding tube which is going to be manufactured in next stage. As a part of developing fuel cladding for the Sodium-cooled Fast Reactor (SFR), preliminary HT9 cladding tube was fabricated in cooperation with a domestic steelmaking company. Microstructure as well as mechanical tests like axial tensile test, biaxial burst test, and pressurized creep test of the fuel cladding were carried out. Performance of the domestic HT9 tube was revealed to be similar in the previously fabricated foreign HT9 tube. Further prototype FMS cladding tube is going to be manufactured in next year based on this experience. Various test items like mechanical test, sodium compatibility test, microstructural analysis, basic property, cladding performance under transient situation, and performance under ion and neutron irradiation are going be performed in the future to set up the relevant technology for the licensing of the SFR cladding tube

  20. The significance of cladding material on the integrity of nuclear pressure vessels with cracks

    International Nuclear Information System (INIS)

    Sattari-Far, Iradj.

    1989-05-01

    The significance of the austenitic cladding layer is reviewed in this literature study. The cladding induced stresses are generally not considered when evaluating the severity of flaws in reactor pressure vessels. It has been shown that this emission may be misleading. The necessity to consider the cladding induced stresses is also emphasized in the latest edition of ASME XI. Contrary to what is commonly assumed, the austenitic cladding displays a charpy V transition region with a low ductility. The interface material (HAZ) is the most influenced region by irradiation, and a transition shift of over 100 degree C may be expected. Because of the significant difference in the thermal expansion coefficients of the cladding and the base metal, cladding induced stresses can be set up. Even after PWHT, residual stresses of yield magnitude remain in the cladding and the HAZ at ambient temperature. The cladding induced stresses are temperature dependent and decrease as the temperature increases. The cladding induced stresses have a significant influence on small defects near the inside surface of a pressure vessel. For semielliptical surface cracks, the maximum CTOD-value along the crack front is not found at the deepest point, but in the cladding/base metal interface, having a magnitude three times higher than the value in the deepest point. It implies that this type of crack would propagate along the clad/base material interface. At some point in time, the crack will reach a geometry which may cause such a severe condition at the deepest point that it will start to grow in the depth direction as well. The initiation and growth behaviour of such cracks need to be investigated to be able to assess the significance of cladding on the integrity of nuclear pressure vessels. (author) (50 figs., 33 refs.)

  1. A method for limitation of probability of accumulation of fuel elements claddings damage in WWER

    OpenAIRE

    Sergey N. Pelykh; Mark V. Nikolsky; S. D. Ryabchikov

    2014-01-01

    The aim is to reduce the probability of accumulation of fuel elements claddings damage by developing a method to control the properties of the fuel elements on stages of design and operation of WWER. An averaged over the fuel assembly WWER-1000 fuel element is considered. The probability of depressurization of fuel elements claddings is found. The ability to predict the reliability of claddings by controlling the factors that determine the properties of the fuel elements is proved. The expedi...

  2. International uranium market

    International Nuclear Information System (INIS)

    Neff, T.L.; Jacoby, H.D.

    1980-12-01

    Discussed in this report are 1) how one might think about uranium demand, resources and supply, 2) how producers and consumers see the market and are likely to behave, including specifics about export and import commitments, and 3) how these actors are brought together in the international market. The general conclusion is that much of current anxiety about future uranium supply results primarily from a brief but difficult period in the mid- to late-1970's; and that current conditions and trends are favorable (at least to consumers) that there is now little basis for concern. Inventories contractual positions and producer commitments--when compared with realistic (or even unrealistic) demand estimates--imply a buyer's market for at least the next decade. The result will be considerable increases in market flexibility and resilience to shock, and real prices that are low relative to those of the past few years. There is a need to reconsider assumptions about desired directions of technological development, for many current programs were planned in an era of pessimism about uranium supply and process. Similar questions must be raided about nonproliferation policies that depend on some level of control of fuel supplies by the industrial nations. With a soft and more diversified uranium market, leverage that may have existed in the past is rapidly being eroded. Finally, as world prices turn soft, there may be significant problems created for U.S. uranium producers, who have relatively high costs in relation to several large-scale foreign suppliers

  3. Uranium hexafluoride handling

    International Nuclear Information System (INIS)

    1991-01-01

    The United States Department of Energy, Oak Ridge Field Office, and Martin Marietta Energy Systems, Inc., are co-sponsoring this Second International Conference on Uranium Hexafluoride Handling. The conference is offered as a forum for the exchange of information and concepts regarding the technical and regulatory issues and the safety aspects which relate to the handling of uranium hexafluoride. Through the papers presented here, we attempt not only to share technological advances and lessons learned, but also to demonstrate that we are concerned about the health and safety of our workers and the public, and are good stewards of the environment in which we all work and live. These proceedings are a compilation of the work of many experts in that phase of world-wide industry which comprises the nuclear fuel cycle. Their experience spans the entire range over which uranium hexafluoride is involved in the fuel cycle, from the production of UF 6 from the naturally-occurring oxide to its re-conversion to oxide for reactor fuels. The papers furnish insights into the chemical, physical, and nuclear properties of uranium hexafluoride as they influence its transport, storage, and the design and operation of plant-scale facilities for production, processing, and conversion to oxide. The papers demonstrate, in an industry often cited for its excellent safety record, continuing efforts to further improve safety in all areas of handling uranium hexafluoride

  4. Uranium tailings sampling manual

    International Nuclear Information System (INIS)

    Feenstra, S.; Reades, D.W.; Cherry, J.A.; Chambers, D.B.; Case, G.G.; Ibbotson, B.G.

    1985-01-01

    The purpose of this manual is to describe the requisite sampling procedures for the application of uniform high-quality standards to detailed geotechnical, hydrogeological, geochemical and air quality measurements at Canadian uranium tailings disposal sites. The selection and implementation of applicable sampling procedures for such measurements at uranium tailings disposal sites are complicated by two primary factors. Firstly, the physical and chemical nature of uranium mine tailings and effluent is considerably different from natural soil materials and natural waters. Consequently, many conventional methods for the collection and analysis of natural soils and waters are not directly applicable to tailings. Secondly, there is a wide range in the physical and chemical nature of uranium tailings. The composition of the ore, the milling process, the nature of tailings depositon, and effluent treatment vary considerably and are highly site-specific. Therefore, the definition and implementation of sampling programs for uranium tailings disposal sites require considerable evaluation, and often innovation, to ensure that appropriate sampling and analysis methods are used which provide the flexibility to take into account site-specific considerations. The following chapters describe the objective and scope of a sampling program, preliminary data collection, and the procedures for sampling of tailings solids, surface water and seepage, tailings pore-water, and wind-blown dust and radon

  5. Oxidants for uranium leaching

    International Nuclear Information System (INIS)

    Ho, E.; Ring, B.

    2007-01-01

    Most uranium ores are leached with sulfuric acid under oxidising conditions. This paper reviews the oxidants that have been traditionally used in uranium leaching and discusses their merits in the context of overall flow sheet considerations. Options for alternative oxidants are also discussed. In acid leaching, ferric ion in solution oxidises insoluble uranium(IV) to soluble uranium (VI). Though ferric ion may be added directly, usually an oxidant is added to the circuit to convert ferrous ion to ferric ion in the liquor so that leaching can continue. The most common oxidants are pyrolusite and sodium chlorate. Pyrolusite is relatively cheap but introduces manganese ions into the liquor and consumes twice as much acid as sodium chlorate. Sodium chlorate is a slow reacting oxidant at low temperatures and acidities, and introduces chloride into the leach liquor. Caro's acid, is a non-polluting reagent that was used successfully at the Nabarlek uranium mine, and provided very good control of oxidising conditions. Other oxidants that are now being considered to overcome the disadvantages of pyrolusite and sodium chlorate are oxygen, hydrogen peroxide and SO 2 /O 2 . Some performance data for these oxidants are presented

  6. Uranium hexafluoride handling. Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    1991-12-31

    The United States Department of Energy, Oak Ridge Field Office, and Martin Marietta Energy Systems, Inc., are co-sponsoring this Second International Conference on Uranium Hexafluoride Handling. The conference is offered as a forum for the exchange of information and concepts regarding the technical and regulatory issues and the safety aspects which relate to the handling of uranium hexafluoride. Through the papers presented here, we attempt not only to share technological advances and lessons learned, but also to demonstrate that we are concerned about the health and safety of our workers and the public, and are good stewards of the environment in which we all work and live. These proceedings are a compilation of the work of many experts in that phase of world-wide industry which comprises the nuclear fuel cycle. Their experience spans the entire range over which uranium hexafluoride is involved in the fuel cycle, from the production of UF{sub 6} from the naturally-occurring oxide to its re-conversion to oxide for reactor fuels. The papers furnish insights into the chemical, physical, and nuclear properties of uranium hexafluoride as they influence its transport, storage, and the design and operation of plant-scale facilities for production, processing, and conversion to oxide. The papers demonstrate, in an industry often cited for its excellent safety record, continuing efforts to further improve safety in all areas of handling uranium hexafluoride. Selected papers were processed separately for inclusion in the Energy Science and Technology Database.

  7. Uranium resources, production and demand

    International Nuclear Information System (INIS)

    1988-01-01

    Nuclear power-generating capacity will continue to expand, albeit at a slower pace than during the past fifteen years. This expansion must be matched by an adequately increasing supply of uranium. This report compares uranium supply and demand data in free market countries with the nuclear industry's natural uranium requirements up to the year 2000. It also reviews the status of uranium exploration, resources and production in 46 countries

  8. The Toxicity of Depleted Uranium

    OpenAIRE

    Briner, Wayne

    2010-01-01

    Depleted uranium (DU) is an emerging environmental pollutant that is introduced into the environment primarily by military activity. While depleted uranium is less radioactive than natural uranium, it still retains all the chemical toxicity associated with the original element. In large doses the kidney is the target organ for the acute chemical toxicity of this metal, producing potentially lethal tubular necrosis. In contrast, chronic low dose exposure to depleted uranium may not produce a c...

  9. Review of experimental data for modelling LWR fuel cladding behaviour under loss of coolant accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Massih, Ali R. [Quantum Technologies AB, Uppsala Science Park (Sweden)

    2007-02-15

    Extensive range of experiments has been conducted in the past to quantitatively identify and understand the behaviour of fuel rod under loss-of-coolant accident (LOCA) conditions in light water reactors (LWRs). The obtained experimental data provide the basis for the current emergency core cooling system acceptance criteria under LOCA conditions for LWRs. The results of recent experiments indicate that the cladding alloy composition and high burnup effects influence LOCA acceptance criteria margins. In this report, we review some past important and recent experimental results. We first discuss the background to acceptance criteria for LOCA, namely, clad embrittlement phenomenology, clad embrittlement criteria (limitations on maximum clad oxidation and peak clad temperature) and the experimental bases for the criteria. Two broad kinds of test have been carried out under LOCA conditions: (i) Separate effect tests to study clad oxidation, clad deformation and rupture, and zirconium alloy allotropic phase transition during LOCA. (ii) Integral LOCA tests, in which the entire LOCA sequence is simulated on a single rod or a multi-rod array in a fuel bundle, in laboratory or in a tests and results are discussed and empirical correlations deduced from these tests and quantitative models are conferred. In particular, the impact of niobium in zirconium base clad and hydrogen content of the clad on allotropic phase transformation during LOCA and also the burst stress are discussed. We review some recent LOCA integral test results with emphasis on thermal shock tests. Finally, suggestions for modelling and further evaluation of certain experimental results are made.

  10. Simulation of a pellet-clad mechanical interaction with ABAQUS and its verification

    International Nuclear Information System (INIS)

    Cheon, J.-S.; Lee, B.-H.; Koo, Y.-H.; Sohn, D.-S.; Oh, J.-Y.

    2003-01-01

    Pellet-clad mechanical interaction (PCMI) during power transients for MOX fuel is modelled by a FE method. The PCMI model predicts well clad elongation during power ramp and relaxation during power hold except the fuel behaviour during a power decrease. Higher fiction factor results in the earlier occurrence of PCMI and more enhanced clad elongation. The relaxation is dependent on the irradiation creep rate of the pellet and axial compressive force. Verification of the PCMI model was done using recent MOX experimental data. Temperature and clad elongation for the fuel rod can be evaluated in a reasonable way

  11. Simulation of high burn-up fuel cladding and its safety assessment under LOCA condition

    International Nuclear Information System (INIS)

    Park, Dong Jun; Won, Sung Bin; Choi, Byoung Kwon; Park, Jeong Yong; Koo, Yang Hyun

    2011-01-01

    Current LOCA safety criteria was established in the beginning of 1970s and based on the results obtained from non-irradiated Zircaloy-4 claddings. Because of major advantages in fuel-cycle costs, reactor operation, and waste management, the increase in fuel discharge burn-up is current worldwide trend in the nuclear industry. As the fuel burn-up increases, various phenomena unexpected have been reported due to changes in the condition of reactor operation and in-core environment. Since, it should be considered whether the current Loss-of-coolant accident (LOCA) criteria is suitable for high burn-up fuel cladding or not. In addition, many fuel vendors have recently developed new cladding alloys superior to Zircaloy-4 cladding. The performance of these advanced cladding alloys under LOCA, especially at high burn-up, is not well understood at this time. To better understand high burn-up effects and commercialize new cladding alloys, study of LOCA-related behavior of various types of high burn-up fuel cladding and their data base is essentially required. In this background, postulated LOCA test has been carried out with prehydrided Zircaloy-4 cladding as a surrogate for high burn-up cladding and the relevant results obtained are discussed

  12. Implications and control of fuel-cladding chemical interaction for LMFBR fuel pin design

    International Nuclear Information System (INIS)

    Roake, W.E.

    1977-01-01

    Fuel-cladding-chemical-interaction (FCCI) is typically incorporated into the design of an LMFBR fuel pin as a wastage allowance. Several interrelated factors are considered during the evolution of an LMFBR fuel pin design. Those which are indirectly affected by FCCI include: allowable pin power, fuel restructuring, fission gas migration and release from the fuel, fuel cracking, fuel swelling, in-reactor cladding creep, cladding swelling, and the cladding mechanical strain. Chemical activity of oxygen is the most readily controlled factor in FCCI. Two methods are being investigated: control of total oxygen inventory by limiting fuel O/M, and control of oxygen activity with buffer metals

  13. Formation quality optimization of laser hot wire cladding for repairing martensite precipitation hardening stainless steel

    Science.gov (United States)

    Wen, Peng; Feng, Zhenhua; Zheng, Shiqing

    2015-01-01

    Laser cladding is an advantaged repairing technology due to its low heat input and high flexibility. With preheating wire by resistance heat, laser hot wire cladding shows better process stability and higher deposition efficiency compared to laser cold wire/powder cladding. Multi-pass layer were cladded on the surface of martensite precipitation hardening stainless steel FV520B by fiber laser with ER410NiMo wire. Wire feed rate and preheat current were optimized to obtain stable wire transfer, which guaranteed good formation quality of single pass cladding. Response surface methodology (RSM) was used to optimize processing parameters and predict formation quality of multi-pass cladding. Laser power P, scanning speed Vs, wire feed rate Vf and overlap ratio η were selected as the input variables, while flatness ratio, dilution and incomplete fusion value as the responses. Optimal clad layer with flat surface, low dilution and no incomplete fusion was obtained by appropriately reducing Vf, and increasing P, Vs and η. No defect like pore or crack was found. The tensile strength and impact toughness of the clad layer is respectively 96% and 86% of those of the substrate. The clad layer showed nonuniform microstructure and was divided into quenched areas with coarse lath martensite and tempered areas with tempered martensite due to different thermal cycles in adjacent areas. The tempered areas showed similar hardness to the substrate.

  14. Raman probes based on optically-poled double-clad fiber and coupler

    DEFF Research Database (Denmark)

    Brunetti, Anna Chiara; Margulis, Walter; Rottwitt, Karsten

    2012-01-01

    Two fiber Raman probes are presented, one based on an optically-poled double-clad fiber and the second based on an optically-poled double-clad fiber coupler respectively. Optical poling of the core of the fiber allows for the generation of enough 532nm light to perform Raman spectroscopy...... of a sample of dimethyl sulfoxide (DMSO), when illuminating the waveguide with 1064nm laser light. The Raman signal is collected in the inner cladding, from which it is retrieved with either a bulk dichroic mirror or a double-clad fiber coupler. The coupler allows for a substantial reduction of the fiber...

  15. Corrosion Resistance of Laser Clads of Inconel 625 and Metco 41C

    Science.gov (United States)

    Němeček, Stanislav; Fidler, Lukáš; Fišerová, Pavla

    The present paper explores the impact of laser cladding parameters on the corrosion behaviour of the resulting surface. Powders of Inconel 625 and austenitic Metco 41C steel were deposited on steel substrate. It was confirmed that the level of dilution has profound impact on the corrosion resistance and that dilution has to be minimized. However, the chemical composition of the cladding is altered even in the course of the cladding process, a fact which is related to the increase in the substrate temperature. The cladding process was optimized to achieve maximum corrosion resistance. The results were verified and validated using microscopic observation, chemical analysis and corrosion testing.

  16. Uranium. Resources, production and demand

    International Nuclear Information System (INIS)

    1997-01-01

    The events characterising the world uranium market in the last several years illustrate the persistent uncertainly faced by uranium producers and consumers worldwide. With world nuclear capacity expanding and uranium production satisfying only about 60 per cent of demand, uranium stockpiles continue to be depleted at a high rate. The uncertainty related to the remaining levels of world uranium stockpiles and to the amount of surplus defence material that will be entering the market makes it difficult to determine when a closer balance between uranium supply and demand will be reached. Information in this report provides insights into changes expected in uranium supply and demand until well into the next century. The 'Red Book', jointly prepared by the OECD Nuclear Energy Agency and the International Atomic Energy Agency, is the foremost reference on uranium. This world report is based on official information from 59 countries and includes compilations of statistics on resources, exploration, production and demand as of 1 January 1997. It provides substantial new information from all of the major uranium producing centres in Africa, Australia, Eastern Europe, North America and the New Independent States, including the first-ever official reports on uranium production in Estonia, Mongolia, the Russian Federation and Uzbekistan. It also contains an international expert analysis of industry statistics and worldwide projections of nuclear energy growth, uranium requirements and uranium supply

  17. Sandstone-type uranium deposits

    International Nuclear Information System (INIS)

    Finch, W.I.; Davis, J.F.

    1985-01-01

    World-class sandstone-type uranium deposits are defined as epigenetic concentrations of uranium minerals occurring as uneven impregnations and minor massive replacements primarily in fluvial, lacustrine, and deltaic sandstone formations. The main purpose of this introductory paper is to define, classify, and introduce to the general geologic setting for sandstone-type uranium deposits

  18. Yeguang granite and uranium metallogeny

    International Nuclear Information System (INIS)

    Gong Wenshu

    1988-03-01

    The geological, petrologic and geochemical characteristics of rock body of yeguang granite are discussed with emphasis on clarification of occurrence of uranium in rock body and uranium deposit characteristics related to rock body, based on extensive field observations and laboratory test. View points are presented that granite refers to genetic series of terrestial crust transformation tye and uranium deposit is hot water deposit

  19. Uranium immobilization and nuclear waste

    International Nuclear Information System (INIS)

    Duffy, C.J.; Ogard, A.E.

    1982-02-01

    Considerable information useful in nuclear waste storage can be gained by studying the conditions of uranium ore deposit formation. Further information can be gained by comparing the chemistry of uranium to nuclear fission products and other radionuclides of concern to nuclear waste disposal. Redox state appears to be the most important variable in controlling uranium solubility, especially at near neutral pH, which is characteristic of most ground water. This is probably also true of neptunium, plutonium, and technetium. Further, redox conditions that immobilize uranium should immobilize these elements. The mechanisms that have produced uranium ore bodies in the Earth's crust are somewhat less clear. At the temperatures of hydrothermal uranium deposits, equilibrium models are probably adequate, aqueous uranium (VI) being reduced and precipitated by interaction with ferrous-iron-bearing oxides and silicates. In lower temperature roll-type uranium deposits, overall equilibrium may not have been achieved. The involvement of sulfate-reducing bacteria in ore-body formation has been postulated, but is uncertain. Reduced sulfur species do, however, appear to be involved in much of the low temperature uranium precipitation. Assessment of the possibility of uranium transport in natural ground water is complicated because the system is generally not in overall equilibrium. For this reason, Eh measurements are of limited value. If a ground water is to be capable of reducing uranium, it must contain ions capable of reducing uranium both thermodynamically and kinetically. At present, the best candidates are reduced sulfur species

  20. Uranium industry in the USSR

    International Nuclear Information System (INIS)

    Nikipelov, B.V.; Chernov, A.G.

    1990-01-01

    A brief historical account of the Soviet production of natural and enriched uranium is given. The geological and geographical location of major uranium deposits are mentioned. The processing of natural ores including in-situ leaching (ISL) is also briefly described. Gas centrifuges play a large part in uranium enrichment. The role of Techsnabexport for the export of nuclear materials is explained

  1. Uranium content of Philippine coals

    International Nuclear Information System (INIS)

    De la Rosa, A.M.; Sombrito, E.Z.; Nuguid, Z.S.; Bulos, A.M.; Bucoy, B.M.; De la Cruz, M.

    1984-01-01

    Uranium content of coal samples from seven areas in the Philippines, i.e. Cebu, Semirara, Bislig, Albay, Samar, Malangas and Polilio Is. was found to contain trace quantities of uranium. The mean value of 0.401 ppm U is lower than reported mean uranium contents for coal from other countries. (ELC)

  2. Uranium extraction from underground deposits

    International Nuclear Information System (INIS)

    Wolfe, C.R.

    1982-01-01

    Uranium is extracted from underground deposits by passing an aqueous oxidizing solution of carbon dioxide over the ore in the presence of calcium ions. Complex uranium carbonate or bicarbonate ions are formed which enter the solution. The solution is forced to the surface and the uranium removed from it

  3. Uranium deposits of Zaire

    International Nuclear Information System (INIS)

    Kitmut, D.; Malu wa Kalenga

    1979-01-01

    Since April 1960, following the closing of the Shinkolobwe mine, the Republic of Zaire has ceased to be a producer of uranium. Nevertheless, Gecamines (Generale des carrieres et mines du Zaire), a wholly state-owned company, is continuing its research on uranium occurrences which have been discovered in its concession in the course of aerial radiometric prospecting. The most recent campaign was the one carried out in 1969 and 1972 by Hunting Company. On-the-ground verification of these shows has not yet resulted in the discovery of a workable deposit. There are other sectors cutting across Zaire which might well contain uranium deposits: this is true of the sedimentary phosphates of the region of Lower Zaire as well as of the frontier region between Zaire and the Central African Empire. However, no detailed exploration work has yet been carried out. (author)

  4. Process for recovering uranium

    Science.gov (United States)

    MacWood, G. E.; Wilder, C. D.; Altman, D.

    1959-03-24

    A process useful in recovering uranium from deposits on stainless steel liner surfaces of calutrons is presented. The deposit is removed from the stainless steel surface by washing with aqueous nitric acid. The solution obtained containing uranium, chromium, nickel, copper, and iron is treated with an excess of ammonium hydroxide to precipitnte the uranium, iron, and chromium and convert the nickel and copper to soluble ammonio complexions. The precipitated material is removed, dried and treated with carbon tetrachloride at an elevated temperature of about 500 to 600 deg C to form a vapor mixture of UCl/ sub 4/, UCl/sub 5/, FeCl/sub 3/, and CrCl/sub 4/. The UCl/sub 4/ is separated from this vapor mixture by selective fractional condensation at a temperature of about 500 to 400 deg C.

  5. The Kintyre uranium project

    International Nuclear Information System (INIS)

    Larson, B.

    1997-01-01

    The Kintyre Uranium Project is being developed by Canning Resources Pty Ltd, a subsidiary of Rio Tinto (formerly CRA). The work on the project includes the planning and management of a number of background environmental studies. The company has also commissioned studies by external consultants into process technologies, mining strategies and techniques for extracting the uranium ore from the waste rock. In addition, Canning Resources has made a detailed assessment of the worldwide market potential for Australian uranium in the late 1990s and into the 21st century. The most significant factor affecting the future of this project is the current product price. This price is insufficient to justify the necessary investment to bring this project into production

  6. Uranium market and resources

    International Nuclear Information System (INIS)

    Capus, G.; Arnold, T.

    2005-01-01

    Under the combined effect of various factors, such as interrogations related to facing the climatic changes, the increasing prices of oil versus announced decrease of its resources, the major geopolitical evolution and the remarkable development of Asia, we live nowadays a revival of nuclear power in the very front of stage. In tis context, the following question is posed: could the nuclear fission be a sustainable source of energy when taking into consideration the availability of uranium resources? The article aims at pinpointing the knowledge we have about the world uranium resources, their limits of uncertainty and the relation between knowledge resources and market evolution. To conclude, some susceptible tracks are proposed to improve the using process of uranium resources particularly in softening the impact of high prices

  7. Australian uranium resources

    International Nuclear Information System (INIS)

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

    1987-01-01

    Australia's uranium resources amount to 29% of the WOCA countries (world outside centrally-planned-economies areas) low-cost Reasonably Assured Resources and 28% of the WOCA countries low-cost Estimated Additional Resources. As at 1 January 1986, the Bureau of Mineral Resources estimated Australia's uranium resources as: (1) Cost range to US$80/kg U -Reasonably Assured Resources, 465 000 t U; Estimated Additional Resources, 256 000 t U; (2) Cost range US$80-130/kg U -Reasonably Assured Resources, 56 000 t U; Estimated Additional Resources, 127 000 t U. Most resources are contained in Proterozoic unconformity-related deposits in the Alligator Rivers uranium field in the Northern Territory (Jabiluka, Ranger, Koongarra, Nabarlek deposits) and the Proterozoic stratabound deposit at Olympic Dam on the Stuart Shelf in South Australia

  8. Uranium exploration in Ecuador

    International Nuclear Information System (INIS)

    Severne, B.; Penaherrera, P.F.; Fiallos, V.S.

    1981-01-01

    The 600-km segment of the Andean Cordillera in Ecuador includes zones that can be correlated, geologically, with uranium districts elsewhere in the Andes. It is believed that these essentially unexplored zones have the potential for economic uranium mineralization. Exploration activity to date has been limited, although it has involved both geochemical and radiometric techniques to evaluate geological concepts. Minor uranium occurrences (with chemical analyses up to 100 ppm) have been encountered, which provide further incentive to commence large-scale systematic exploration. It is recognized that a very large exploration budget and considerable technical expertise will be required to ensure exploration success. Consequently, participation by groups of proven capability from other countries will be sought for Ecuador's national exploration programme. (author)

  9. Uranium Enrichment, an overview

    International Nuclear Information System (INIS)

    Coates, J.H.

    1994-01-01

    This general presentation on uranium enrichment will be followed by lectures on more specific topics including descriptions of enrichment processes and assessments of the prevailing commercial and industrial situations. I shall therefore avoid as much as possible duplications with these other lectures, and rather dwell on: some theoretical aspects of enrichment in general, underlying the differences between statistical and selective processes, a review and comparison between enrichment processes, remarks of general order regarding applications, the proliferation potential of enrichment. It is noteworthy that enrichment: may occur twice in the LWR fuel cycle: first by enriching natural uranium, second by reenriching uranium recovered from reprocessing, must meet LWR requirements, and in particular higher assays required by high burn up fuel elements, bears on the structure of the entire front part of the fuel cycle, namely in the conversion/reconversion steps only involving UF 6 for the moment. (author). tabs., figs., 4 refs

  10. Uranium extraction technology

    International Nuclear Information System (INIS)

    1993-01-01

    In 1983 the Nuclear Energy Agency of the Organisation for Economic Co-operation and Development (OECD/NEA) and the IAEA jointly published a book on Uranium Extraction Technology. A primary objective of this report was to document the significant technological developments that took place during the 1970s. The purpose of this present publication is to update and expand the original book. It includes background information about the principle of the unit operations used in uranium ore processing and summarizes the current state of the art. The publication also seeks to preserve the technology and the operating 'know-how' developed over the past ten years. This publication is one of a series of Technical Reports on uranium ore processing that have been prepared by the Division of Nuclear Fuel Cycle and Waste Management at the IAEA. A complete list of these reports is included as an addendum. Refs, figs and tabs

  11. Uranium ore processing

    International Nuclear Information System (INIS)

    Ritcey, G.M.; Haque, K.E.; Lucas, B.H.; Skeaff, J.M.

    1983-01-01

    The authors have developed a complete method of recovering separately uranium, thorium and radium from impure solids such as ores, concentrates, calcines or tailings containing these metals. The technique involves leaching, in at least one stage. The impure solids in finely divided form with an aqueous leachant containing HCl and/or Cl 2 until acceptable amounts of uranium, thorium and radium are dissolved. Uranium is recovered from the solution by solvent extraction and precipitation. Thorium may also be recovered in the same manner. Radium may be recovered by at least one ion exchange, absorption and precipitation. This amount of iron in the solution must be controlled before the acid solution may be recycled for the leaching process. The calcine leached in the first step is prepared in a two stage roast in the presence of both Cl 2 and a metal sulfide. The first stage is at 350-450 0 and the second at 550-700 0

  12. High-resolution characterization of oxidation mechanism of zirconium nuclear fuel cladding alloys

    International Nuclear Information System (INIS)

    Hu, J.; Lozano-Perez, S.; Grovenor, C.

    2015-01-01

    Full text of publication follows. Zirconium alloys are used extensively as cladding materials in modern light water reactors to separate the uranium dioxide (UO 2 ) fuel rods and the coolant water in order to prevent the escape of radioactive fission products whilst maintaining heat transfer to the coolant. With increasing demand for high burn-up in modern nuclear reactors, environmental degradation of these alloys is now the life limiting factor for fuel assemblies. As part of the MUZIC-2 collaboration studying oxidation and hydrogen pickup in Zr alloys, several high resolution analysis techniques have been used to study the microstructure of a range of commercial and developmental Zr alloys. The sample used for this investigation was prepared from a Westinghouse TM developmental alloy with composition of Zr-0.9Nb-0.01Sn-0.08Fe (wt %) in the recrystallized condition. The sample was oxidised in an autoclave at EDF Energy under simulated PWR water conditions at 360 C. degrees for 360 days. Using Transmission Electron Microscope (TEM), we have studied the development of the equiaxed-columnar-equiaxed grain structure, and observe that the columnar grains are both longer and show a stronger preferred texture in more corrosion-resistant alloys. Fresnel imaging revealed the existence of both parallel interconnected pores and some vertically interconnected pores along the columnar oxide grain boundaries, which become more disconnected near the metal-oxide interface. Electron Energy Loss Spectroscopy (EELS) provided accurate quantitative analysis of the oxygen concentration across the interface, identifying the existence of local regions of stoichiometric ZrO and Zr 3 O 2 with varying thickness. These observations will be discussed in the context of current models for oxidation in zirconium alloys. (authors)

  13. The uranium International trade

    International Nuclear Information System (INIS)

    Gonzalez U, L.A.

    1993-01-01

    The aim of this thesis is the understanding of how the present dynamic of uranium International trade is developed, the variables which fall into, the factors that are affecting and conditioning it, in order to clarify which are going to be the outlook in the future of this important resource in front of the present ecological situation and the energetic panorama of XXI Century. For this purpose, as starting point, the uranium is considered as a strategic material which importance take root in its energetic potential as alternate energy source, and for this reason in Chapter I, the general problem of raw materials, its classification and present situation in the global market is presented. In Chapter II, by means of a historical review, is explain what uranium is, how it was discovered, and how since the end of the past Century and during the last three decades of present, uranium pass of practically unknown element, to the position of a strategic raw material, which by degrees, generate an International market, owing to its utilization as a basic resource in the generation of energy. Chapter III, introduce us in the roll played by uranium, since its warlike applications until its utilization in nuclear reactors for the generation of electricity. Also is explain the reason for this change in the perception at global level. Finally, in Chapter IV we enter upon specifically in the present conditions of the International market of this mineral throughout the trends of supply and demand, the main producers, users, price dynamics, and the correlation among these economical variables and other factors of political, social and ecological nature. All of these with the purpose to found out, if there exist, a meaning of the puzzle that seems to be the uranium International trade

  14. Uranium 2000 : International symposium on the process metallurgy of uranium

    International Nuclear Information System (INIS)

    Ozberk, E.; Oliver, A.J.

    2000-01-01

    The International Symposium on the Process Metallurgy of Uranium has been organized as the thirtieth annual meeting of the Hydrometallurgy Section of the Metallurgical Society of the Canadian Institute of Mining, Metallurgy and Petroleum (CIM). This meeting is jointly organized with the Canadian Mineral Processors Division of CIM. The proceedings are a collection of papers from fifteen countries covering the latest research, development, industrial practices and regulatory issues in uranium processing, providing a concise description of the state of this industry. Topics include: uranium industry overview; current milling operations; in-situ uranium mines and processing plants; uranium recovery and further processing; uranium leaching; uranium operations effluent water treatment; tailings disposal, water treatment and decommissioning; mine decommissioning; and international regulations and decommissioning. (author)

  15. Production of sized particles of uranium oxides and uranium oxyfluorides

    International Nuclear Information System (INIS)

    Knudsen, I.E.; Randall, C.C.

    1976-01-01

    A process is claimed for converting uranium hexafluoride (UF 6 ) to uranium dioxide (UO 2 ) of a relatively large particle size in a fluidized bed reactor by mixing uranium hexafluoride with a mixture of steam and hydrogen and by preliminary reacting in an ejector gaseous uranium hexafluoride with steam and hydrogen to form a mixture of uranium and oxide and uranium oxyfluoride seed particles of varying sizes, separating the larger particles from the smaller particles in a cyclone separator, recycling the smaller seed particles through the ejector to increase their size, and introducing the larger seed particles from the cyclone separator into a fluidized bed reactor where the seed particles serve as nuclei on which coarser particles of uranium dioxide are formed. 9 claims, 2 drawing figures

  16. Corrosion-resistant uranium

    Science.gov (United States)

    Hovis, Jr., Victor M.; Pullen, William C.; Kollie, Thomas G.; Bell, Richard T.

    1983-01-01

    The present invention is directed to the protecting of uranium and uranium alloy articles from corrosion by providing the surfaces of the articles with a layer of an ion-plated metal selected from aluminum and zinc to a thickness of at least 60 microinches and then converting at least the outer surface of the ion-plated layer of aluminum or zinc to aluminum chromate or zinc chromate. This conversion of the aluminum or zinc to the chromate form considerably enhances the corrosion resistance of the ion plating so as to effectively protect the coated article from corrosion.

  17. Uranium Conversion & Enrichment

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-02-06

    The isotopes of uranium that are found in nature, and hence in ‘fresh’ Yellowcake’, are not in relative proportions that are suitable for power or weapons applications. The goal of conversion then is to transform the U3O8 yellowcake into UF6. Conversion and enrichment of uranium is usually required to obtain material with enough 235U to be usable as fuel in a reactor or weapon. The cost, size, and complexity of practical conversion and enrichment facilities aid in nonproliferation by design.

  18. Uranium Transport Modeling

    International Nuclear Information System (INIS)

    Bostick, William D.

    2008-01-01

    Uranium contamination is prevalent at many of the U.S. DOE facilities and at several civilian sites that have supported the nuclear fuel cycle. The potential off-site mobility of uranium depends on the partitioning of uranium between aqueous and solid (soil and sediment) phases. Hexavalent U (as uranyl, UO 2 2+ ) is relatively mobile, forming strong complexes with ubiquitous carbonate ion which renders it appreciably soluble even under mild reducing conditions. In the presence of carbonate, partition of uranyl to ferri-hydrate and select other mineral phases is usually maximum in the near-neutral pH range ∼ 5-8. The surface complexation reaction of uranyl with iron-containing minerals has been used as one means to model subsurface migration, used in conjunction with information on the site water chemistry and hydrology. Partitioning of uranium is often studied by short-term batch 'equilibrium' or long-term soil column testing ; MCLinc has performed both of these methodologies, with selection of method depending upon the requirements of the client or regulatory authority. Speciation of uranium in soil may be determined directly by instrumental techniques (e.g., x-ray photoelectron spectroscopy, XPS; x-ray diffraction, XRD; etc.) or by inference drawn from operational estimates. Often, the technique of choice for evaluating low-level radionuclide partitioning in soils and sediments is the sequential extraction approach. This methodology applies operationally-defined chemical treatments to selectively dissolve specific classes of macro-scale soil or sediment components. These methods recognize that total soil metal inventory is of limited use in understanding bioavailability or metal mobility, and that it is useful to estimate the amount of metal present in different solid-phase forms. Despite some drawbacks, the sequential extraction method can provide a valuable tool to distinguish among trace element fractions of different solubility related to mineral phases

  19. Uranium ore dressing process

    International Nuclear Information System (INIS)

    Floter, W.; Schoning, G.; Landsiedel, K.

    1982-01-01

    A process for dressing uranium ore containing water-soluble substances is claimed. In a mixture of crushed ore and water the solid material and water are partly separated. The first fraction, which contains the smaller grain-size portion of the solid material and the major part of the water is passed through a water softening means for removal of undesired ions. The second fraction, which contains the coarser grain-size portion of the solid material and most of the solids of the first fraction from which undesired ions have been at least partly removed is leached. The liquid is processed to obtain the dissolved uranium compounds

  20. Uranium ore deposits

    International Nuclear Information System (INIS)

    Angelelli, Victorio.

    1984-01-01

    The main uranium deposits and occurrences in the Argentine Republic are described, considering, in principle, their geologic setting, the kind of 'model' of the mineralization and its possible origin, and describing the ore species present in each case. The main uraniferous accumulations of the country include the models of 'sandstong type', veintype and impregnation type. There are also other kinds of accumulations, as in calcrete, etc. The main uranium production has been registered in the provinces of Mendoza, Salta, La Rioja, Chubut, Cordoba and San Luis. In each case, the minerals present are mentioned, having been recognized 37 different species all over the country (M.E.L.) [es

  1. Uranium exploration techniques

    International Nuclear Information System (INIS)

    Nichols, C.E.

    1984-01-01

    The subject is discussed under the headings: introduction (genetic description of some uranium deposits; typical concentrations of uranium in the natural environment); sedimentary host rocks (sandstones; tabular deposits; roll-front deposits; black shales); metamorphic host rocks (exploration techniques); geologic techniques (alteration features in sandstones; favourable features in metamorphic rocks); geophysical techniques (radiometric surveys; surface vehicle methods; airborne methods; input surveys); geochemical techniques (hydrogeochemistry; petrogeochemistry; stream sediment geochemistry; pedogeochemistry; emanometry; biogeochemistry); geochemical model for roll-front deposits; geologic model for vein-like deposits. (U.K.)

  2. Uranium conversion wastes

    International Nuclear Information System (INIS)

    Vicente, R.; Dellamano, J.C.

    1989-12-01

    A set of mathematical equations was developed and used to estimate the radiological significance of each radionuclide potentially present in the uranium refining industry effluents. The equations described the evolution in time of the radionuclides activities in the uranium fuel cycle, from mining and milling, through the yellowcake, till the conversion effluents. Some radionuclides that are not usually monitored in conversion effluents (e.g. Pa-231 and Ac-227) were found to be potentially relevant from the radiological point of view in conversion facilities, and are certainly relevant in mining and milling industry, at least in a few waste streams. (author) [pt

  3. Joining uranium to steel

    International Nuclear Information System (INIS)

    Perkins, M.A.

    1976-05-01

    A method has been devised which will allow the joining of uranium to steel by fusion welding through the use of an intermediate material. Uranium-0.5 titanium was joined to AISI 304L stainless steel by using a vanadium insert. Also, a method is now available for selecting possible filler metals when two entirely dissimilar metals need to be joined. This method allows a quantitative ranking to be made of the possible filler metals and thus the most likely candidate can be selected

  4. Irradiation effects of the zirconium oxidation and the uranium diffusion in zirconia; Effets d'irradiation sur l'oxydation du zirconium et la diffusion de l'uranium dans la zircone

    Energy Technology Data Exchange (ETDEWEB)

    Bererd, N

    2003-07-01

    The context of this study is the direct storage of spent fuel assemblies after operation in reactor. In order to obtain data on the capacities of the can as the uranium diffusion barrier, a fundamental study has been carried out for modelling the internal cladding surface under and without irradiation. The behaviour of zirconium in reactor conditions has at first been studied. A thin uranium target enriched with fissile isotope has been put on a zirconium sample, the set being irradiated by a thermal neutrons flux leading to the fission of the deposited uranium. The energetic history of the formed fission products has revealed two steps: 1)the zirconium oxidation and 2)the diffusion of uranium in the zirconia formed at 480 degrees C. A diffusion coefficient under irradiation has been measured. Its value is 10{sup -15} cm{sup 2}.s{sup -1}. In order to be able to reveal clearly the effect of the irradiation by the fission products on the zirconium oxidation, measurements of thermal oxidation and under {sup 129}Xe irradiation have been carried out. They have shown that the oxidation is strongly accelerated by the irradiation and that the temperature is negligible until 480 degrees C. On the other hand, the thermal diffusion of the uranium in zirconium and in zirconia has been studied by coupling ion implantation and Rutherford backscattering spectroscopy. This study shows that the uranium diffuses in zirconium and is trapped in zirconia in a UO{sub 3} shape. (O.M.)

  5. The uranium market 1980 - 1990

    International Nuclear Information System (INIS)

    Darmayan, Philippe

    1980-01-01

    The Supply and Demand Committee of the Uranium Institute was established to monitor continuously information and developments bearing on the uranium market and to publish from time to time reports giving its views on the supply and demand outlook. The last Uranium institute supply and demand report was published in July 1979 and a summary was given by Mr. Erkes at the last Uranium Institute symposium. Its main conclusions were that from 1979 to 1990 the flexibilities of the market were such as to offer adequate scope to producers and consumers of uranium to ensure a balance between supply and demand. Is that conclusion still valid one and a half years later [fr

  6. Uranium enriched granites in Sweden

    International Nuclear Information System (INIS)

    Wilson, M.R.; Aakerblom, G.

    1980-01-01

    Granites with uranium contents higher than normal occur in a variety of geological settings in the Swedish Precambrian, and represent a variety of granite types and ages. They may have been generated by (1) the anatexis of continental crust (2) processes occurring at a much greater depth. They commonly show enrichement in F, Sn, W and/or Mo. Only in one case is an important uranium mineralization thought to be directly related to a uranium-enriched granite, while the majority of epigenetic uranium mineralizations with economic potential are related to hydrothermal processes in areas where the bedrock is regionally uranium-enhanced. (Authors)

  7. Uranium sorption by tannin resins

    International Nuclear Information System (INIS)

    Olivares Rieumont, S.; Martinez Luzardo, J.; Torres Hernandez, J.; Lima Cazorla, D. de la Rosa.

    1998-01-01

    The sorption of uranium by immobilised Eucalyptus Saligna Sm. and Lysiloma latisiliqua L tannins was investigated. Immobilization condition were analyzed. These resins resulted suitable adsorbent for the concentration of uranium from aqueous systems. The sorption of uranium is pH dependent. At pH 5.5 maximum in sorption capacity is registered. The presence of appreciable amount of sodium chloride do not have any effect on uranium removal. Carbonate and calcium ions in concentrations similar to these that could be found in sea water and other natural water do not decrease the uranium uptake. Tannin resins can be used several times without an appreciable decay of their sorption capacity

  8. Fault rocks and uranium mineralization

    International Nuclear Information System (INIS)

    Tong Hangshou.

    1991-01-01

    The types of fault rocks, microstructural characteristics of fault tectonite and their relationship with uranium mineralization in the uranium-productive granite area are discussed. According to the synthetic analysis on nature of stress, extent of crack and microstructural characteristics of fault rocks, they can be classified into five groups and sixteen subgroups. The author especially emphasizes the control of cataclasite group and fault breccia group over uranium mineralization in the uranium-productive granite area. It is considered that more effective study should be made on the macrostructure and microstructure of fault rocks. It is of an important practical significance in uranium exploration

  9. Experiments to understand the corrosion process of fuel rod claddings

    International Nuclear Information System (INIS)

    Groeschel, F.; Hermann, A.

    1997-01-01

    Fuel rods in light water reactors have to respond to the trends in increased burn-up and extended dwelling time in reactor. Waterside corrosion of the cladding affecting wall thickness, mechanical stability due to hydriding and the heat transfer due to the low thermal conductivity of the oxide scale may become the limiting factors. The corrosion process is complex and involves a large variety of mechanisms. Understanding of the process is important for safe operation and a prerequisite for development of improved materials. A variety of analytical techniques and mechanical tests, including examination of irradiated pathfinder rods, are used to tackle the different aspects. (author) 6 figs., 1 tab., 17 refs

  10. Corrosion Resistant Cladding by YAG Laser Welding in Underwater Environment

    International Nuclear Information System (INIS)

    Tsutomi Kochi; Toshio Kojima; Suemi Hirata; Ichiro Morita; Katsura Ohwaki

    2002-01-01

    It is known that stress-corrosion cracking (SCC) will occur in nickel-base alloys used in Reactor Pressure Vessel (RPV) and Internals of nuclear power plants. A SCC sensitivity has been evaluated by IHI in each part of RPV and Internals. There are several water level instrumentation nozzles installed in domestic BWR RPV. In water level instrumentation nozzles, 182 type nickel-base alloys were used for the welding joint to RPV. It is estimated the SCC potential is high in this joint because of a higher residual stress than the yield strength (about 400 MPa). This report will describe a preventive maintenance method to these nozzles Heat Affected Zone (HAZ) and welds by a corrosion resistant cladding (CRC) by YAG Laser in underwater environment (without draining a reactor water). There are many kinds of countermeasures for SCC, for example, Induction Heating Stress Improvement (IHSI), Mechanical Stress Improvement Process (MSIP) and so on. A YAG laser CRC is one of them. In this technology a laser beam is used for heat source and irradiated through an optical fiber to a base metal and SCC resistant material is used for welding wires. After cladding the HAZ and welds are coated by the corrosion resistant materials so their surfaces are improved. A CRC by gas tungsten arc welding (GTAW) in an air environment had been developed and already applied to a couple of operating plants (16 Nozzles). This method was of course good but it spent much time to perform because of an installation of some water-proof working boxes to make a TIG-weldability environment. CRC by YAG laser welding in underwater environment has superior features comparing to this conventional TIG method as follows. At the viewpoint of underwater environment, (1) an outage term reduction (no drainage water). (2) a radioactive exposure dose reduction for personnel. At that of YAG laser welding, (1) A narrower HAZ. (2) A smaller distortion. (3) A few cladding layers. A YAG laser CRC test in underwater

  11. Standard specification for architectural flat glass clad polycarbonate

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2010-01-01

    1.1 This specification covers the quality requirements for cut sizes of glass clad polycarbonate (GCP) for use in buildings as security, detention, hurricane/cyclic wind-resistant, and blast and ballistic-resistant glazing applications. 1.2 The values stated in inch-pound units are to be regarded as the standard. The values given in parentheses are for information only. 1.3 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

  12. Embrittlement of zircaloy cladding due to oxygen uptake (CBRTTL)

    International Nuclear Information System (INIS)

    Reymann, G.A.

    1979-02-01

    A model for embrittlement of zircaloy due to oxygen uptake at high temperatures is described. The model defines limits for oxygen content and temperature which, if exceeded, give rise to zircaloy cladding which is sufficiently embrittled to cause failure either on quenching or normal handling following a transient. A significant feature of this model is that the onset of embrittlement is dependent on the cooling rate. A distinction is made between slow and fast cooling, with the boundary at 100 K/s. The material property correlations and computer subcodes described in MATPRO are developed for use in Light Water Reactor (LWR) codes

  13. Ultrasonic signal processing for sizing under-clad flaws

    International Nuclear Information System (INIS)

    Shankar, R.; Paradiso, T.J.; Lane, S.S.; Quinn, J.R.

    1985-01-01

    Ultrasonic digital data were collected from underclad cracks in sample pressure vessel specimen blocks. These blocks were weld cladded under different processes to simulate actual conditions in US Pressure Water Reactors. Each crack was represented by a flaw-echo dynamic curve which is a plot of the transducer motion on the surface as a function of the ultrasonic response into the material. Crack depth sizing was performed by identifying in the dynamic curve the crack tip diffraction signals from the upper and lower tips. This paper describes the experimental procedure, digital signal processing methods used and algorithms developed for crack depth sizing

  14. Widely tunable femtosecond solitonic radiation in photonic crystal fiber cladding

    DEFF Research Database (Denmark)

    Peng, J. H.; Sokolov, A. V.; Benabid, F.

    2010-01-01

    We report on a means to generate tunable ultrashort optical pulses. We demonstrate that dispersive waves generated by solitons within the small-core features of a photonic crystal fiber cladding can be used to obtain femtosecond pulses tunable over an octave-wide spectral range. The generation...... process is highly efficient and occurs at the relatively low laser powers available from a simple Ti:sapphire laser oscillator. The described phenomenon is general and will play an important role in other systems where solitons are known to exist....

  15. RIA simulation tests using driver tube for ATF cladding

    Energy Technology Data Exchange (ETDEWEB)

    Cinbiz, Mahmut N. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Brown, N. R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Lowden, R. R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Linton, K. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrani, K. A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-07-01

    Pellet-cladding mechanical interaction (PCMI) is a potential failure mechanism for accident-tolerant fuel (ATF) cladding candidates during a reactivity-initiated accident (RIA). This report summarizes Fiscal Year (FY) 2017 research activities that were undertaken to evaluate the PCMI-like hoop-strain-driven mechanical response of ATF cladding candidates. To achieve various RIA-like conditions, a modified-burst test (MBT) device was developed to produce different mechanical pulses. The calibration of the MBT instrument was accomplished by performing mechanical tests on unirradiated Generation-I iron-chromium-aluminum (FeCrAl) alloy samples. Shakedown tests were also conducted in both FY 2016 and FY 2017 using unirradiated hydrided ZIRLO™ tube samples. This milestone report focuses on testing of ATF materials, but the benchmark tests with hydrided ZIRLO™ tube samples are documented in a recent journal article.a For the calibration and benchmark tests, the hoop strain was monitored using strain gauges attached to the sample surface in the hoop direction. A novel digital image correlation (DIC) system composed of a single high-speed camera and an array of six mirrors was developed for the MBT instrument to better resolve the failure behavior of samples and to provide useful data for validation of high-fidelity modeling and simulation tools. The DIC system enable a 360° view of a sample’s outer surface. This feature was added to the instrument to determine the precise failure location on a sample’s surface for strain predictions. The DIC system was tested on several silicon carbide fiber/silicon carbide matrix (SiC/SiC) composite tube samples at various pressurization rates of the driver tube (which correspond to the strain rates for the samples). The hoop strains for various loading conditions were determined for the SiC/SiC composite tube samples. Future work is planned to enhance understanding of the failure behavior of the ATF cladding candidates of age

  16. Laser cladding of wear resistant metal matrix composite coatings

    International Nuclear Information System (INIS)

    Yakovlev, A.; Bertrand, Ph.; Smurov, I.

    2004-01-01

    A number of coatings with wear-resistant properties as well as with a low friction coefficient are produced by laser cladding. The structure of these coatings is determined by required performance and realized as metal matrix composite (MMC), where solid lubricant serves as a ductile matrix (e.g. CuSn), reinforced by appropriate ceramic phase (e.g. WC/Co). One of the engineered coating with functionally graded material (FGM) structure has a dry friction coefficient 0.12. Coatings were produced by coaxial injection of powder blend into the zone of laser beam action. Metallographic and tribological examinations were carried out confirming the advanced performance of engineered coatings

  17. Properties of light water reactor spent fuel cladding. Interim report

    International Nuclear Information System (INIS)

    Farwick, D.G.; Moen, R.A.

    1979-08-01

    The Commercial Waste and Spent Fuel Packaging Program will provide containment packages for the safe storage or disposal of spent Light Water Reactor (LWR) fuel. Maintaining containment of radionuclides during transportation, handling, processing and storage is essential, so the best understanding of the properties of the materials to be stored is necessary. This report provides data collection, assessment and recommendations for spent LWR fuel cladding materials properties. Major emphasis is placed on mechanical properties of the zircaloys and austenitic stainless steels. Limited information on elastic constants, physical properties, and anticipated corrosion behavior is also provided. Work is in progress to revise these evaluations as the program proceeds

  18. Recovery of uranium from seawater

    International Nuclear Information System (INIS)

    Hirotsu, Takahiro; Takagi, Norio; Katoh, Shunsaku

    1995-01-01

    Present status of the development of chelating adsorbents for the recovery of uranium from seawater is outlined with emphasis on the research by the author. Uranium is estimated to exist as stable tri (carbonate) uranylate (6) ion in seawater in a very low concentration. The adsorbent for uranium from seawater in a very low concentration. The adsorbent for uranium from seawater should have high selectivity and affinity for uranium around pH 8. The required characteristics for uranium adsorbent are examined. Various chelating adsorbents have been proposed for the uranium adsorbent and their structures are discussed. Amidoxime type adsorbents have the highest adsorbing power for uranium among the adsorbents hitherto developed and fibrous amidoxime adsorbents are most promising for the practical application. Synthesis, structure and suitable shape of the amidoxime adsorbents are discussed. Uranium adsorption behavior and the amount of saturated adsorption are examined theoretically based on the complexation of an amidoxime monomer and the formula for the adsorption equiliburium is derived. The adsorption and recovery process for uranium from seawater is composed of adsorption, desorption, separation and concentration and finally, uranium is recovered as the yellow cake. A floating body mooring system is proposed by Nobukawa. (T.H.)

  19. Types of hydrogenic uranium deposits

    International Nuclear Information System (INIS)

    Kondrat'eva, I.A.; Lisitsin, A.K.; Komarova, G.V.

    1980-01-01

    Principles of uranium deposit systematics are considered. Systematization, presented in this paper, is based on a regularity that the main industrial exogenic epigenetic uranium concentrations are formed in zones of reduction geochemical barriers. Types of uranium-bearing ground waters and nature of uranium reducing agents were taken into account during systematization. Hydrogenic uranium deposits are related to 3 types: formed by ground (1 type), stratal (2 type) and vein (3 type) waters. By the nature of uranium reducing agents 2 deposit subtypes are marked out: A - in rocks with syngenetic reducing agents, B - in rocks with epigenetic reducing agents. Uranium deposits are also differentiated by nature of reducing agents distribution in ore-containing rocks [ru

  20. Uranium deposits in granitic rocks

    International Nuclear Information System (INIS)

    Nishimori, R.K.; Ragland, P.C.; Rogers, J.J.W.; Greenberg, J.K.

    1977-01-01

    This report is a review of published data bearing on the geology and origin of uranium deposits in granitic, pegmatitic and migmatitic rocks with the aim of assisting in the development of predictive criteria for the search for similar deposits in the U.S. Efforts were concentrated on the so-called ''porphyry'' uranium deposits. Two types of uranium deposits are primarily considered: deposits in pegmatites and alaskites in gneiss terrains, and disseminations of uranium in high-level granites. In Chapter 1 of this report, the general data on the distribution of uranium in igneous and metamorphic rocks are reviewed. Chapter 2 contains some comments on the classification of uranium deposits associated with igneous rocks and a summary of the main features of the geology of uranium deposits in granites. General concepts of the behavior of uranium in granites during crustal evolution are reviewed in Chapter 3. Also included is a discussion of the relationship of uranium mineralization in granites to the general evolution of mobile belts, plus the influence of magmatic and post-magmatic processes on the distribution of uranium in igneous rocks and related ore deposits. Chapter 4 relates the results of experimental studies on the crystallization of granites to some of the geologic features of uranium deposits in pegmatites and alaskites in high-grade metamorphic terrains. Potential or favorable areas for igneous uranium deposits in the U.S.A. are delineated in Chapter 5. Data on the geology of specific uranium deposits in granitic rocks are contained in Appendix 1. A compilation of igneous rock formations containing greater than 10 ppM uranium is included in Appendix 2. Appendix 3 is a report on the results of a visit to the Roessing area. Appendix 4 is a report on a field excursion to eastern Canada