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Sample records for fuel matrix corrosion

  1. Spent fuel UO2 matrix corrosion behaviour studies through alpha-doped UO2 pellets leaching

    International Nuclear Information System (INIS)

    Muzeau, B.; Jegou, C.; Broudic, V.

    2005-01-01

    Full text of publication follows: The option of direct disposal of spent nuclear fuel in a deep geological formation raises the need to investigate the long-term behaviour of the UO 2 matrix in aqueous media subjected to α-β-γ radiations. The β-γ emitters account for the most of the activity of spent fuel at the moment it is removed from the reactor, but diminish within a millennial time frame by over three orders of magnitude to less than the long-term activity. The latter persist over much longer time periods and must therefore be taken into account over geological disposal scale. In the present investigation the UO 2 matrix corrosion under alpha radiation is studied as a function of different parameters such as: the alpha activity, the carbonates and hydrogen concentrations,.. In order to study the effect of alpha radiolysis of water on the UO 2 matrix, 238/239 Pu doped UO 2 pellets (0.22 %wt. Pu total) were fabricated with different 238 Pu/ 239 Pu ratio to reproduce the alpha activity of a 47 GWd.t HMi -1 UOX spent fuel at different milestones in time (15, 50, 1500, 10000 and 40000 years). Undoped UO 2 pellets were also available as reference sample. Leaching experiments were conducted in deionized or carbonated water (NaHCO 3 1 mM), under Argon (O 2 2 30% gas mixture. Previous experiments conducted in deionized water under argon atmosphere, have shown a good correlation between alpha activity and uranium release for the 15-, 1500- and 40000-years alpha doped UO 2 batches. Besides, uranium release in the leachate is controlled either by the kinetics, or by the thermodynamics. Provided the solubility limit of uranium is not achieved, uranium concentration increases and is only limited by the kinetics, unless precipitation occurs and the uranium concentration remains constant over time. These controls are highly dependant on the solution chemistry (HCO 3 - , pH, Eh,..), the atmosphere (Ar, Ar/H 2 ,..), and the radiolysis strength. The experimental matrix

  2. Corrosion-induced changes in pore-size distributions of fuel-matrix material

    International Nuclear Information System (INIS)

    Krautwasser, P.; Eatherly, W.P.

    1981-01-01

    In order to understand the mechanism of metallic fission-product adsorption and desorption as well as diffusion in graphitic materials, a detailed knowledge of the material microstructure is essential. Different types of grahitic matrix material used or to be used in fuel elements of the German HTR Program were measured at ORNL in cooperation with the Hahn-Meitner-Institut Berlin. Actual measurements of fission product diffusion and adsorption/desorption were performed at HMI Berlin

  3. Corrosion of fuel assembly materials

    International Nuclear Information System (INIS)

    Noe, M.; Frejaville, G.; Beslu, P.

    1985-08-01

    Corrosion of zircaloy-4 is reviewed in relation with previsions of improvement in PWRs performance: higher fuel burnup; increase coolant temperature, implying nucleate boiling on the hot clad surfaces; increase duration of the cycle due to load-follow operation. Actual knowledge on corrosion rates, based partly on laboratory tests, is insufficient to insure that external clad corrosion will not constitute a limitation to these improvements. Therefore, additional testing within representative conditions is felt necessary [fr

  4. Corrosion of Metal-Matrix Composites with Aluminium Alloy Substrate

    Directory of Open Access Journals (Sweden)

    B. Bobic

    2010-03-01

    Full Text Available The corrosion behaviour of MMCs with aluminium alloy matrix was presented. The corrosion characteristics of boron-, graphite-, silicon carbide-, alumina- and mica- reinforced aluminium MMCs were reviewed. The reinforcing phase influence on MMCs corrosion rate as well as on various corrosion forms (galvanic, pitting, stress corrosion cracking, corrosion fatique, tribocorrosion was discussed. Some corrosion protection methods of aluminium based MMCs were described

  5. Fuel corrosion processes under waste disposal conditions

    International Nuclear Information System (INIS)

    Shoesmith, D.W.

    1999-09-01

    Under the oxidizing conditions likely to be encountered in the Yucca Mountain Repository, fuel dissolution is a corrosion process involving the coupling of the anodic dissolution of the fuel with the cathodic reduction of oxidants available within the repository. The oxidants potentially available to drive fuel corrosion are environmental oxygen, supplied by the transport through the permeable rock of the mountain and molecular and radical species produced by the radiolysis of available aerated water. The mechanism of these coupled anodic and cathodic reactions is reviewed in detail. While gaps in understanding remain, many kinetic features of these reactions have been studied in considerable detail, and a reasonably justified mechanism for fuel corrosion is available. The corrosion rate is determined primarily by environmental factors rather than the properties of the fuel. Thus, with the exception of increase in rate due to an increase in surface area, pre-oxidation of the fuel has little effect on the corrosion rate

  6. Corrosion surveillance in spent fuel storage pools

    International Nuclear Information System (INIS)

    Howell, J.P.

    1996-01-01

    In mid-1991, corrosion of aluminum-clad spent nuclear fuel was observed in the light-water filled basins at the Savannah River site. A corrosion surveillance program was initiated in the P, K, L-Reactor basins and in the Receiving Basin for Offsite Fuels (RBOF). This program verified the aggressive nature of the pitting corrosion and provided recommendations for changes in basin operations to permit extended longer term interim storage. The changes were implemented during 1994--1996 and have resulted in significantly improved basin water quality with conductivity in the 1--3 microS/cm range. Under these improved conditions, no new pitting has been observed over the last three years. This paper describes the corrosion surveillance program at SRS and what has been learned about the corrosion of aluminum-clad in spent fuel storage pools

  7. 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.

  8. NON-CORROSIVE REACTOR FUEL SYSTEM

    Science.gov (United States)

    Herrick, C.C.

    1962-08-14

    A non-corrosive nuclear reactor fuel system was developed utilizing a molten plutonium-- iron alloy fuel having about 2 at.% carbon and contained in a tantalum vessel. This carbon reacts with the interior surface of the tantalum vessel to form a plutonium resistant self-healing tantalum carbide film. (AEC)

  9. Microscopic Examination of a Corrosion Front in Spent Nuclear Fuel

    International Nuclear Information System (INIS)

    J.A. Fortner; A.J. Kropf; R.J. Finch; J.C. Cunnane

    2006-01-01

    Spent uranium oxide nuclear fuel hosts a variety of trace chemical constituents, many of which must be sequestered from the biosphere during fuel storage and disposal. In this paper we present synchrotron x-ray absorption spectroscopy and microscopy findings that illuminate the resultant local chemistry of neptunium and plutonium within spent uranium oxide nuclear fuel before and after corrosive alteration in an air-saturated aqueous environment. We find the plutonium and neptunium in unaltered spent fuel to have a +4 oxidation state and an environment consistent with solid-solution in the UO 2 matrix. During corrosion in an air-saturated aqueous environment, the uranium matrix is converted to uranyl U(VI)O 2 2+ mineral assemblage that is depleted in plutonium and neptunium relative to the parent fuel. At the corrosion front interface between intact fuel and the uranyl-mineral corrosion layer, we find evidence of a thin (∼20 micrometer) layer that is enriched in plutonium and neptunium within a predominantly U 4+ environment. Available data for the standard reduction potentials for NpO 2+ /Np 4+ and UO 2 2+ /U 4+ couples indicate that Np(IV) may not be effectively oxidized to Np(V) at the corrosion potentials of uranium dioxide spent nuclear fuel in air-saturated aqueous solutions. Neptunium is an important radionuclide in dose contribution according to performance assessment models of the proposed U. S. repository at Yucca Mountain, Nevada. A scientific understanding of how the UO 2 matrix of spent nuclear fuel impacts the oxidative dissolution and reductive precipitation of neptunium is needed to predict its behavior at the fuel surface during aqueous corrosion. Neptunium would most likely be transported as aqueous Np(V) species, but for this to occur it must first be oxidized from the Np(IV) state found within the parent spent nuclear fuel [1]. In the immediate vicinity of the spent fuel's surface the redox and nucleation behavior is likely to promote

  10. High corrosion-resistant fuel spacers

    International Nuclear Information System (INIS)

    Yoshida, Toshimi; Takase, Iwao; Ikeda, Shinzo; Masaoka, Isao; Nakajima, Junjiro.

    1986-01-01

    Purpose: To enable manufacturing BWR fuel spacers by prior-art production process, using a zirconium-base alloy having very excellent corrosion resistance. Method: A highly improved nodular-resistant, corrosion-resistant zirconium alloy is devised by adding a slight amount of niobium, titanium and vanadium to zircaloy, of which fuel spacers are produced. That is, there can be obtained an alloy having much more excellent nodular resistance than conventional zircaloy, and free from a large change in plasticity, workability, and weldability, by adding to zirconium about 1.5 % of tin, about 0.15 % of iron, about 0.05 % of chromium, about 0.05 % of nickel, and 0.05 to 0.5 % of at least one or two kinds of niobium, titanium and vanadium. Using this zirconium-base alloy can manufacture fuel spacers by the same manufacturing process, thus improving economy and reliability. (Kamimura, M.)

  11. 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)

  12. Nuclear Energy Research Initiative. Development of a Stabilized Light Water Reactor Fuel Matrix for Extended Burnup

    International Nuclear Information System (INIS)

    BD Hanson; J Abrefah; SC Marschman; SG Prussin

    2000-01-01

    The main objective of this project is to develop an advanced fuel matrix capable of achieving extended burnup while improving safety margins and reliability for present operations. In the course of this project, the authors improve understanding of the mechanism for high burnup structure (HBS) formation and attempt to design a fuel to minimize its formation. The use of soluble dopants in the UO 2 matrix to stabilize the matrix and minimize fuel-side corrosion of the cladding is the main focus

  13. Silicon carbide corrosion in HTGR fuel particles

    International Nuclear Information System (INIS)

    Gruebmeier, H.; Naoumidis, A.; Thiele, B.A.

    1977-01-01

    The silicon carbide layer in TRISO-coated HTR fuel particles occasionally is attacked during irradiation by fission products generated within the kernel material. Investigations to define the cause of the SiC corrosion are described, which included ceramographic, microradiographic and microanalytical studies on irradiated and unirradiated coated particles of various kernel compositions. The results of these studies showed that the presence of chlorine within the particle, in combination with certain metallic fission products or uranium, can lead to corrosion of the SiC layer. These results provided the basis for establishing a model, which relates the transport of fission products as volatile metal chlorides and their chemical reaction with the SiC corrosion. In addition this model is consistent with the fact that under the influence of a thermal gradient corrosive attack of the SiC occurs on the cooler side of the particle. The correlation between the occurrence of SiC corrosion during irradiation and heavy-metal transport observed in thermal gradient annealing studies of unirradiated particles of the same barch constitutes the basis for a new method for quality control. (orig.) [de

  14. Drying characteristics of thorium fuel corrosion products

    Science.gov (United States)

    Smith, R.-E. Lords

    2004-07-01

    The open literature and accessible US Department of Energy-sponsored reports were reviewed for the dehydration and rehydration characteristics of potential corrosion products from thorium metal and thorium oxide nuclear fuels. Mixed oxides were not specifically examined unless data were given for performance of mixed thorium-uranium fuels. Thorium metal generally corrodes to thorium oxide. Physisorbed water is readily removed by heating to approximately 200 °C. Complete removal of chemisorbed water requires heating above 1000 °C. Thorium oxide adsorbs water well in excess of the amount needed to cover the oxide surface by chemisorption. The adsorption of water appears to be a surface phenomenon; it does not lead to bulk conversion of the solid oxide to the hydroxide. Adsorptive capacity depends on both the specific surface area and the porosity of the thorium oxide. Heat treatment by calcination or sintering reduces the adsorption capacity substantially from the thorium oxide produced by metal corrosion.

  15. Corrosion Behavior of Metal Matrix Composites

    Science.gov (United States)

    1993-01-01

    described below. Considering the anodic reaction in chloride media Cu + CI-= CuC1- + e- (1) CuCI - + C1- = CuC1- 2 (2) the corrosion process involves...step. At the lower applied potentials of region 1, the rate of formation of the CuCI is slow in comparison to the transport rate to the interface of...the CP- and the rate of CuCI dissolution, equation (2). In this region, the CuCl is dissolved during the corrosion process and the corrosion behavior

  16. Coupon Surveillance For Corrosion Monitoring In Nuclear Fuel Basin

    Energy Technology Data Exchange (ETDEWEB)

    Mickalonis, J. I.; Murphy, T. R.; Deible, R.

    2012-10-01

    Aluminum and stainless steel coupons were put into a nuclear fuel basin to monitor the effect of water chemistry on the corrosion of fuel cladding. These coupons have been monitored for over ten years. The corrosion and pitting data is being used to model the kinetics and estimate the damage that is occurring to the fuel cladding.

  17. Corrosion assessment of dry fuel storage containers

    Energy Technology Data Exchange (ETDEWEB)

    Graves, C.E.

    1994-09-01

    The structural stability as a function of expected corrosion degradation of 75 dry fuel storage containers located in the 200 Area Low-Level Waste Burial Grounds was evaluated. These containers include 22 concrete burial containers, 13 55-gal (208-l) drums, and 40 Experimental Breeder Reactor II (EBR-II) transport/storage casks. All containers are buried beneath at least 48 in. of soil and a heavy plastic tarp with the exception of 35 of the EBR-II casks which are exposed to atmosphere. A literature review revealed that little general corrosion is expected and pitting corrosion of the carbon steel used as the exterior shell for all containers (with the exception of the concrete containers) will occur at a maximum rate of 3.5 mil/yr. Penetration from pitting of the exterior shell of the 208-l drums and EBR-II casks is calculated to occur after 18 and 71 years of burial, respectively. The internal construction beneath the shell would be expected to preclude containment breach, however, for the drums and casks. The estimates for structural failure of the external shells, large-scale shell deterioration due to corrosion, are considerably longer, 39 and 150 years respectively for the drums and casks. The concrete burial containers are expected to withstand a service life of 50 years.

  18. Corrosion and Erosion-Corrosion Processes of Metal-Matrix Composites in Slurry Conditions

    Science.gov (United States)

    Flores, J. F.; Neville, A.; Kapur, N.; Gnanavelu, A.

    2012-03-01

    The corrosion and erosion-corrosion (EC) processes of four metal-matrix composites (MMCs) in a simulated cooling water environment have been assessed in this article. The MMCs consisted of two Ni-base and two Fe-base matrices alloyed with different concentrations of chromium, molybdenum, boron, silicon, and carbon; the matrices were reinforced with tungsten carbide (WC) particles. The corrosion behavior has been investigated using a combination of potentiostatic polarization and post-tests surface analysis. The EC processes were studied by in situ electrochemical techniques measuring the current density and corrosion potential response at different slurry temperatures and sand content. At static conditions it was found that as the temperature increased, there was a transition from a homogeneous corrosion of the matrix to an interfacial corrosion mechanism. The Ni-base MMCs showed a better corrosion resistance and interestingly a highly alloyed matrix did not significantly improved MMC's corrosion resistance. In the in situ EC tests, the Fe-base MMCs showed a constant increase in the current density at all sand contents. Whereas, significant changes were not observed in the Ni-base MMCs below 0.5 g/L. Although sand content had an effect on the monitored current density (the current increased as the sand content increased) this effect was less pronounced above 3 g/L.

  19. Modelling the radiolytic corrosion of α-doped UO2 and spent nuclear fuel

    Science.gov (United States)

    Liu, Nazhen; Qin, Zack; Noël, James J.; Shoesmith, David W.

    2017-10-01

    A model previously developed to predict the corrosion rate of spent fuel (UO2) inside a failed waste container has been adapted to simulate the rates measured on a wide range of α-doped UO2 and spent fuel specimens. This simulation confirms the validity of the model and demonstrates that the steady-state corrosion rate is controlled by the radiolytic production of H2O2 (which has been shown to be the primary oxidant driving fuel corrosion), irrespective of the reactivity of the UO2 matrix. The model was then used to determine the consequences of corrosion inside a failed container resealed by steel corrosion products. The possible accumulation of O2, produced by H2O2 decomposition, was found to accelerate the corrosion rate in a closed system. However, the simultaneous accumulation of radiolytic H2, which is activated as a reductant on the noble metal (ε) particles in the spent fuel, rapidly overcame this acceleration leading to the eventual suppression of the corrosion rate to insignificant values. Calculations also showed that, while the radiation dose rate, the H2O2 decomposition ratio, and the surface coverage of ε particles all influenced the short term corrosion rate, the influence of the radiolytically produced H2 was the overwhelming influence in reducing the rate to negligible level (i.e., <10-20 mol m-2 s-1).

  20. Control of corrosion in an aqueous nuclear fuel storage basin

    International Nuclear Information System (INIS)

    Zimmerman, C.A.

    1981-01-01

    Observations made during thirty years of experience in operating a nuclear fuel storage basin, used for storing a wide variety of spent nuclear fuels underwater have identified several forms of corrosion such as galvanic, pitting and crevice attack. Examples of some of the forms of corrosion observed and their causes are discussed, along with the measures taken to mitigate the corrosive attack. The paper also describes the procedure used to reduce corrosion by: surveillance of design, selection of materials for application in the basin, and inspection of items in the storage basin

  1. Corrosion resistance of a copper canister for spent nuclear fuel

    International Nuclear Information System (INIS)

    1983-04-01

    The report presents an evaluation of copper as canister material for spent nuclear fuel. The evaluation is made from the viewpoint of corrosion and applies to a concept of 1977. Supplementary corrosion studies have been performed. The report includes 9 appendices which deal with experimental data. (G.B.)

  2. Electrometallurgical treatment of aluminum-matrix fuels

    International Nuclear Information System (INIS)

    Willit, J.L.; Gay, E.C.; Miller, W.E.; McPheeters, C.C.; Laidler, J.J.

    1996-01-01

    The electrometallurgical treatment process described in this paper builds on our experience in treating spent fuel from the Experimental Breeder Reactor (EBR-II). The work is also to some degree, a spin-off from applying electrometallurgical treatment to spent fuel from the Hanford single pass reactors (SPRs) and fuel and flush salt from the Molten Salt Reactor Experiment (MSRE) in treating EBR-II fuel, we recover the actinides from a uranium-zirconium fuel by electrorefining the uranium out of the chopped fuel. With SPR fuel, uranium is electrorefined out of the aluminum cladding. Both of these processes are conducted in a LiCl-KCl molten-salt electrolyte. In the case of the MSRE, which used a fluoride salt-based fuel, uranium in this salt is recovered through a series of electrochemical reductions. Recovering high-purity uranium from an aluminum-matrix fuel is more challenging than treating SPR or EBR-II fuel because the aluminum- matrix fuel is typically -90% (volume basis) aluminum

  3. 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

  4. Electrochemical studies on spent fuel corrosion processes

    International Nuclear Information System (INIS)

    Pablo, J. de; Casas, I.; Clarens, F.; Gimenez, J.; Rovira, M.

    2003-01-01

    This presentations is mainly based on the electrochemical studies carried out by the Canadian team and the research group of the Berlin University. Electrochemical studies allow to study separately both the anodic reaction which corresponds-sources on UO 2 -electrodes response is one of to the UO 2 dissolution and the cathodic reaction that is the reduction of the oxidants. By using intensity current-potential plots a mechanisms of UO 2 corrosion has been established. At-300 mV (vs SCE), irreversible oxidation of UO 2 takes place and dissolution begins. In the absence of complexing agents like carbonate, an oxidised layer is formed at 100 mV a stoichiometry close to UO 2 . In carbonate medium, the oxidized layer is not formed because the U(VI) formed is rapidly dissolved. Results in terms of dissolution rates obtained by electrochemical measurements are similar to the ones obtained in dissolution experiments by using flow through reactors and similar kinetic laws are obtained. The effect of external α and γ-sources on UO 2 -electrodes response is one of the few available data on the effects of radiolysis on the UO 2 dissolution rate and can offer a complementary knowledge to the spent fuel and α-doped pellets dissolution experiments. (Author)

  5. Corrosion in batteries and fuel-cell power sources

    International Nuclear Information System (INIS)

    Cieslak, W.R.

    1987-01-01

    Batteries and fuel cells, as electrochemical power sources, provide energy through controlled redox reactions. Because these devices contain electrochemically active components, they place metals in contact with environments in which the metals may corrode. The shelf lives of batteries, particularly those that operate at ambient temperatures depend on very slow rates of corrosion of the electrode materials at open circuit. The means of reducing this corrosion must also be evaluated for its influence on performance. A second major corrosion consideration in electrochemical power sources involves the hardware. Again, shelf lives and service lives depend on very good corrosion resistance of the containment materials and inactive components, such as separators. In those systems in which electrolyte purity is important, even small amounts of corrosion that have not lessened structural integrity can degrade performance. There is a wide variety of batteries and fuel cells, and new systems are constantly under development. Therefore, to illustrate the types of corrosion phenomena that occur, this article will discuss the following systems: lead-acid batteries, alkaline batteries (in terms of the sintered nickel electrode only), lithium ambient-temperature batteries, aluminum/air batteries, sodium/sulfur batteries, phosphoric acid (H/sub 3/PO/sub 4/) fuel cells, and molten carbonate fuel cells

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

    International Nuclear Information System (INIS)

    Kawamura, Hirotaka; Kanbe, Hiromu; Furuya, Masahiro

    2002-01-01

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

  7. Reducing Actinide Production Using Inert Matrix Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Deinert, Mark [Colorado School of Mines, Golden, CO (United States)

    2017-08-23

    The environmental and geopolitical problems that surround nuclear power stem largely from the longlived transuranic isotopes of Am, Cm, Np and Pu that are contained in spent nuclear fuel. New methods for transmuting these elements into more benign forms are needed. Current research efforts focus largely on the development of fast burner reactors, because it has been shown that they could dramatically reduce the accumulation of transuranics. However, despite five decades of effort, fast reactors have yet to achieve industrial viability. A critical limitation to this, and other such strategies, is that they require a type of spent fuel reprocessing that can efficiently separate all of the transuranics from the fission products with which they are mixed. Unfortunately, the technology for doing this on an industrial scale is still in development. In this project, we explore a strategy for transmutation that can be deployed using existing, current generation reactors and reprocessing systems. We show that use of an inert matrix fuel to recycle transuranics in a conventional pressurized water reactor could reduce overall production of these materials by an amount that is similar to what is achievable using proposed fast reactor cycles. Furthermore, we show that these transuranic reductions can be achieved even if the fission products are carried into the inert matrix fuel along with the transuranics, bypassing the critical separations hurdle described above. The implications of these findings are significant, because they imply that inert matrix fuel could be made directly from the material streams produced by the commercially available PUREX process. Zirconium dioxide would be an ideal choice of inert matrix in this context because it is known to form a stable solid solution with both fission products and transuranics.

  8. Corrosion report for the U-Mo fuel concept

    Energy Technology Data Exchange (ETDEWEB)

    Henager, Charles H. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Bennett, Wendy D. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Doherty, Ann L. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Fuller, E. S. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Hardy, John S. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Omberg, Ronald P. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States)

    2014-08-28

    The Fuel Cycle Research and Development (FCRD) program of the Office of Nuclear Energy (NE) has implemented a program to develop a Uranium-Molybdenum (U-Mo) metal fuel for Light Water Reactors (LWR)s. Uranium-Molybdenum fuel has the potential to provide superior performance based on its thermo-physical properties, which includes high thermal conductivity for less stored heat energy. With sufficient development, it may be able to provide the Light Water industry with a melt-resistant accident tolerant fuel with improved safety response. However, the corrosion of this fuel in reactor water environments needs to be further explored and optimized by additional alloying. The Pacific Northwest National Laboratory has been tasked with performing ex-reactor corrosion testing to characterize the performance of U-Mo fuel. This report documents the results of the effort to characterize and develop the U-Mo metal fuel concept for LWRs with regard to corrosion testing. The results of a simple screening test in buffered water at 30°C using surface alloyed U-10Mo is documented and discussed. The screening test was used to guide the selection of several potential alloy improvements that were found and are recommended for further testing in autoclaves to simulate PWR water conditions more closely.

  9. Corrosion studies on retrievable spent fuel containers: a progress report

    International Nuclear Information System (INIS)

    Ludemann, W.D.; Abrego, L.; McCright, R.D.

    1978-12-01

    Spent fuel canisters stored in halite (NaCl) deposits (salt beds) are subject to a severely corrosive environment when the hot brine inclusions, rich in calcium and magnesium chlorides, migrate to the canister. Since no data base exists on corrosion in halite brines, a survey was made of the corrosion resistance of potential canister materials in other concentrated brine environments. Corrosion-resistant metals include Ta, Ti Code 12, TiPd Alloy, Inconel 625, Hastelloy C-276, and Fe-base 29-4 Alloy. Although carbon steels have cost and availability advantages, they suffer from excessive corrosion rates in brines. Corrosion-resistant nonmetals include carbon, Teflon-type fluorocarbons, epoxide coatings, and polymer cements. While these materials are not suitable for constructing the canister, they could be used as a protective coating on a carbon steel canister. On the basis of this survey, we recommend a coated carbon steel canister, used with cathodic protection. It is important to start a test program to gather a data base on the corrosion of materials in halite brines and to verify the suitability of canister materials

  10. Effects of alpha-decay on spent fuel corrosion behaviour

    International Nuclear Information System (INIS)

    Wiss, T.; Rondinella, V.V.; Cobos, J.; Wegen, D.H.; Amme, M.; Ronchi, C.

    2004-01-01

    An overview of results in the area of spent fuel characterization as nuclear waste is presented. These studies are focused on primary aspects of spent fuel corrosion, by considering different fuel compositions and burn ups, as well as a wide set of environmental conditions. The key parameter is the storage time of the fuel e.g. in view of spent fuel retrieval or in view of its final disposal. To extrapolate data obtainable from a laboratory-acceptable timescale to those expected after storage periods of interest have elapsed (amounting in the extreme case to geological ages) is a tough challenge. Emphasis is put on key aspects of fuel corrosion related to fuel properties at a given age and environmental conditions expected in the repository: e.g. the fuel activity (radiolysis effects), the effects of helium build-up and of groundwater composition. A wide range of techniques, from traditional leaching experiments to advanced electrochemistry, and of materials, including spent fuel with different compositions/burnups and analogues like the so-called alpha-doped UO 2 , are employed for these studies. The results confirm the safety of European underground repository concepts. (authors)

  11. DIMENSIONALLY STABLE, CORROSION RESISTANT NUCLEAR FUEL

    Science.gov (United States)

    Kittel, J.H.

    1963-10-31

    A method of making a uranium alloy of improved corrosion resistance and dimensional stability is described. The alloy contains from 0-9 weight per cent of an additive of zirconium and niobium in the proportions by weight of 5 to 1 1/ 2. The alloy is cold rolled, heated to two different temperatures, air-cooled, heated to a third temperature, and quenched in water. (AEC)

  12. Carbonate fuel cell endurance: Hardware corrosion and electrolyte management status

    Energy Technology Data Exchange (ETDEWEB)

    Yuh, C.; Johnsen, R.; Farooque, M.; Maru, H.

    1993-01-01

    Endurance tests of carbonate fuel cell stacks (up to 10,000 hours) have shown that hardware corrosion and electrolyte losses can be reasonably controlled by proper material selection and cell design. Corrosion of stainless steel current collector hardware, nickel clad bipolar plate and aluminized wet seal show rates within acceptable limits. Electrolyte loss rate to current collector surface has been minimized by reducing exposed current collector surface area. Electrolyte evaporation loss appears tolerable. Electrolyte redistribution has been restrained by proper design of manifold seals.

  13. Carbonate fuel cell endurance: Hardware corrosion and electrolyte management status

    Energy Technology Data Exchange (ETDEWEB)

    Yuh, C.; Johnsen, R.; Farooque, M.; Maru, H.

    1993-05-01

    Endurance tests of carbonate fuel cell stacks (up to 10,000 hours) have shown that hardware corrosion and electrolyte losses can be reasonably controlled by proper material selection and cell design. Corrosion of stainless steel current collector hardware, nickel clad bipolar plate and aluminized wet seal show rates within acceptable limits. Electrolyte loss rate to current collector surface has been minimized by reducing exposed current collector surface area. Electrolyte evaporation loss appears tolerable. Electrolyte redistribution has been restrained by proper design of manifold seals.

  14. 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

  15. Copper anode corrosion affects power generation in microbial fuel cells

    KAUST Repository

    Zhu, Xiuping

    2013-07-16

    Non-corrosive, carbon-based materials are usually used as anodes in microbial fuel cells (MFCs). In some cases, however, metals have been used that can corrode (e.g. copper) or that are corrosion resistant (e.g. stainless steel, SS). Corrosion could increase current through galvanic (abiotic) current production or by increasing exposed surface area, or decrease current due to generation of toxic products from corrosion. In order to directly examine the effects of using corrodible metal anodes, MFCs with Cu were compared with reactors using SS and carbon cloth anodes. MFCs with Cu anodes initially showed high current generation similar to abiotic controls, but subsequently they produced little power (2 mW m-2). Higher power was produced with microbes using SS (12 mW m-2) or carbon cloth (880 mW m-2) anodes, with no power generated by abiotic controls. These results demonstrate that copper is an unsuitable anode material, due to corrosion and likely copper toxicity to microorganisms. © 2013 Society of Chemical Industry.

  16. Pressurized water reactor fuel performance problems connected with fuel cladding corrosion processes

    International Nuclear Information System (INIS)

    Dobrevski, I.; Zaharieva, N.

    2008-01-01

    Generally, Pressurized Water Reactor (WWER, PWR) Fuel Element Performance is connected with fuel cladding corrosion and crud deposition processes. By transient to extended fuel cycles in nuclear power reactors, aiming to achieve higher burnup and better fuel utilization, the role of these processes increases significantly. This evolution modifies the chemical and electrochemical conditions in the reactor primary system, including change of fuel claddings' environment. The higher duty cores are always attended with increased boiling (sub-cooled nucleate boiling) mainly on the feed fuel assemblies. This boiling process on fuel cladding surfaces can cause different consequences on fuel element cladding's environment characteristics. In the case of boiling at the cladding surfaces without or with some cover of corrosion product deposition, the behavior of gases dissolved in water phase is strongly influenced by the vapor generation. The increase of vapor partial pressure will reduce the partial pressures of dissolved gases and will cause their stripping out. By these circumstances the concentrations of dissolved gases in cladding wall water layer can dramatically decrease, including also the case by which all dissolved gases to be stripped out. On the other hand it is known that the hydrogen is added to primary coolant in order to avoid the production of oxidants by radiolysis of water. It is clear that if boiling strips out dissolved hydrogen, the creation of oxidizing conditions at the cladding surfaces will be favored. In this case the local production of oxidants will be a result from local processes of water radiolysis, by which not only both oxygen (O 2 ) and hydrogen (H 2 ) but also hydrogen peroxide (H 2 O 2 ) will be produced. While these hydrogen and oxygen will be stripped out preferentially by boiling, the bigger part of hydrogen peroxide will remain in wall water phase and will act as the most important factor for creation of oxidizing conditions in fuel

  17. Uranyl peroxide enhanced nuclear fuel corrosion in seawater.

    Science.gov (United States)

    Armstrong, Christopher R; Nyman, May; Shvareva, Tatiana; Sigmon, Ginger E; Burns, Peter C; Navrotsky, Alexandra

    2012-02-07

    The Fukushima-Daiichi nuclear accident brought together compromised irradiated fuel and large amounts of seawater in a high radiation field. Based on newly acquired thermochemical data for a series of uranyl peroxide compounds containing charge-balancing alkali cations, here we show that nanoscale cage clusters containing as many as 60 uranyl ions, bonded through peroxide and hydroxide bridges, are likely to form in solution or as precipitates under such conditions. These species will enhance the corrosion of the damaged fuel and, being thermodynamically stable and kinetically persistent in the absence of peroxide, they can potentially transport uranium over long distances.

  18. Corrosion Testing of Stainless Steel Fuel Cell Hardware

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, M.S.; Zawodzinski, C.; Gottesfeld, S.

    1998-11-01

    Metal hardware is gaining increasing interest in polymer electrolyte fuel cell (PEFC) development as a possible alternative to machined graphite hardware because of its potential for low-cost manufacturing combined with its intrinsic high conductivity, minimal permeability and advantageous mechanical properties. A major barrier to more widespread use of metal hardware has been the susceptibility of various metals to corrosion. Few pure metals can withstand the relatively aggressive environment of a fuel cell and thus the choices for hardware are quite limited. Precious metals such as platinum or gold are prohibitively expensive and so tend to be utilized as coatings on inexpensive substrates such as aluminum or stainless steel. The main challenge with coatings has been to achieve pin-hole free surfaces that will remain so after years of use. Titanium has been used to some extent and though it is very corrosion-resistant, it is also relatively expensive and often still requires some manner of surface coating to prevent the formation of a poorly conducting oxide layer. In contrast, metal alloys may hold promise as potentially low-cost, corrosion-resistant materials for bipolar plates. The dozens of commercially available stainless steel and nickel based alloys have been specifically formulated to offer a particular advantage depending upon their application. In the case of austenitic stainless steels, for example, 316 SS contains molybdenum and a higher chromium content than its more common counterpart, 304 SS, that makes it more noble and increases its corrosion resistance. Likewise, 316L SS contains less carbon than 316 SS to make it easier to weld. A number of promising corrosion-resistant, highly noble alloys such as Hastelloy{trademark} or Duplex{trademark} (a stainless steel developed for seawater service) are available commercially, but are expensive and difficult to obtain in various forms (i.e. wire screen, foil, etc.) or in small amounts for R and D

  19. Potential high temperature corrosion problems due to co-firing of biomass and fossil fuels

    DEFF Research Database (Denmark)

    Montgomery, Melanie; Vilhelmsen, T.; Jensen, S.A.

    2008-01-01

    Over the past few years, considerable high temperature corrosion problems have been encountered when firing biomass in power plants due to the high content of potassium chloride in the deposits. Therefore, to combat chloride corrosion problems cofiring of biomass with a fossil fuel has been....... However, the most significant corrosion attack was sulphidation attack at the grain boundaries of 18-8 steel after 3 years exposure. The corrosion mechanisms and corrosion rates are compared with biomass firing and coal firing. Potential corrosion problems due to co-firing biomass and fossil fuels...... corrosion mechanisms appear such as sulphidation and hot corrosion due to sulphate deposits. At Studstrup power plant Unit 4, based on trials with exposure times of 3000 h using 0–20% straw co-firing with coal, the plant now runs with a fuel mix of 10% strawþcoal. Based on results from a 3 years exposure...

  20. Fuel element failures caused by iodine stress corrosion

    International Nuclear Information System (INIS)

    Videm, K.; Lunde, L.

    1976-01-01

    Sections of unirradiated cladding tubes were plugged in both ends by mechanical seals and internally pressurized with argon containing iodine. The time to failure and the strain at failure as a function of stress was determined for tubing with different heat treatments. Fully annealed tubes suffer cracking at the lowest stress but exhibit the largest strains at failure. Elementary iodine is not necessary for stress corrosion: small amounts of iodides of zirconium, iron and aluminium can also give cracking. Moisture, however, was found to act as an inhibitor. A deformation threshold exists below which stress corrosion failure does not occur regardless of the exposure time. This deformation limit is lower the harder the tube. The deformation at failure is dependent on the deformation rate and has a minimum at 0.1%/hr. At higher deformation rates the failure deformation increases, but only slightly for hard tubes. Fuel was over-power tested at ramp rates varying between 0.26 to 30 W/cm min. For one series of fuel pins the failure deformations of 0.8% at high ramp rates were in good agreement with predictions based on stress corrosion experiments. For another series of experiments the failure deformation was surprisingly low, about 0.2%. (author)

  1. Galvanic corrosion of lead coupled with titanium for nuclear fuel waste disposal

    International Nuclear Information System (INIS)

    Mani Mathew, P.; Krueger, P.A.

    1989-01-01

    In the Canadian Nuclear Fuel Waste Management Program, metals and alloys with low melting points are being evaluated for their potential application as cast matrices within used-fuel immobilization containers. This paper describes studies of galvanic corrosion between lead, candidate matrix metal and ASTM Grade-2 titanium, a candidate container-shell material. The studies were conducted under conditions that simulate a breached disposal container surrounded by a bentonite/sand mixture and emplaced in a granitic rock formation at a depth of 500-1000 m. The fractional factorial statistical design of Box Behnken was used in the tests, which covered a wide range of potential conditions that could occur in a nuclear-waste disposal vault. Test temperatures ranged from 293 to 423 K. Ionic strength and oxygen content of the simulated groundwater varied from 0.0015 to 1.37 mol and 0.02 to 8.0 mg/l, respectively. A mathematical expression was derived for the lead corrosion rates as a function of the independent variables: temperature, ionic strength, and oxygen content. This expression was used to calculate the lifetime of the additional barrier that a 25-mm thick lead layer could provide to a titanium container that eventually perforates by erosion. The results show that at least 860 years of additional barrier life could be provided by the lead matrix

  2. Application of a statistical methodology for the comprehension of corrosion phenomena on Phenix spent fuel pins

    International Nuclear Information System (INIS)

    Pantera, L.

    1992-11-01

    The maximum burnup of Phenix fuel elements is strongly conditioned by the internal corrosion of the steel cladding. This thesis is a part of a new study program on the corrosion phenomena. Based on the results of an experimental program during the years 1980-1990 its objective is the use of a statistical methodology for a better comprehension of the corrosion phenomena

  3. Mechanical and corrosion behaviors of developed copper-based metal matrix composites

    Science.gov (United States)

    Singh, Manvandra Kumar; Gautam, Rakesh Kumar; Prakash, Rajiv; Ji, Gopal

    2018-03-01

    This work investigates mechanical properties and corrosion resistances of cast copper-tungsten carbide (WC) metal matrix composites (MMCs). Copper matrix composites have been developed by stir casting technique. Different sizes of micro and nano particles of WC particles are utilized as reinforcement to prepare two copper-based composites, however, nano size of WC particles are prepared by high-energy ball milling. XRD (X-rays diffraction) characterize the materials for involvement of different phases. The mechanical behavior of composites has been studied by Vickers hardness test and compression test; while the corrosion behavior of developed composites is investigated by electrochemical impedance spectroscopy in 0.5 M H2SO4 solutions. The results show that hardness, compressive strength and corrosion resistance of copper matrix composites are very high in comparison to that of copper matrix, which attributed to the microstructural changes occurred during composite formation. SEM (Scanning electron microscopy) reveals the morphology of the corroded surfaces.

  4. Potential high temperature corrosion problems due to co-firing of biomass and fossil fuels

    DEFF Research Database (Denmark)

    Montgomery, Melanie; Vilhelmsen, T.; Jensen, S.A.

    2007-01-01

    Over the past years, considerable high temperature corrosion problems have been encountered when firing biomass in power plants due to the high content of potassium chloride in the deposits. Therefore to combat chloride corrosion problems co-firing of biomass with a fossil fuel has been undertaken...... significant corrosion attack was due to sulphidation attack at the grain boundaries of 18-8 steel after 3 years exposure. The corrosion mechanisms and corrosion rates are compared with biomass firing and coal firing. Potential corrosion problems due to co-firing biomass and fossil fuels are discussed....... appear such as sulphidation and hot corrosion due to sulphate deposits. At Studstrup power plant Unit 4, based on trials with exposure times of 3000 hours using 0-20% straw co-firing with coal, the plant now runs with a fuel of 10% straw + coal. After three years exposure in this environment...

  5. Corrosion-resistant, electrically-conductive plate for use in a fuel cell stack

    Science.gov (United States)

    Carter, J David [Bolingbrook, IL; Mawdsley, Jennifer R [Woodridge, IL; Niyogi, Suhas [Woodridge, IL; Wang, Xiaoping [Naperville, IL; Cruse, Terry [Lisle, IL; Santos, Lilia [Lombard, IL

    2010-04-20

    A corrosion resistant, electrically-conductive, durable plate at least partially coated with an anchor coating and a corrosion resistant coating. The corrosion resistant coating made of at least a polymer and a plurality of corrosion resistant particles each having a surface area between about 1-20 m.sup.2/g and a diameter less than about 10 microns. Preferably, the plate is used as a bipolar plate in a proton exchange membrane (PEMFC) fuel cell stack.

  6. The corrosion of aluminum-clad spent nuclear fuel in wet basin storage

    International Nuclear Information System (INIS)

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

    1997-01-01

    This paper discusses the corrosion of the aluminum-clad spent fuel and the improvements that have been made in the SRS basins since 1993 which have essentially mitigated new corrosion on the fuel. It presents the results of a metallographic examination of two Mk-31A target slugs stored in the L-Reactor basin for about 5 years and a summary of results from the corrosion surveillance programs through 1996

  7. Corrosion Characterization Of ZA-27 Red Mud Metal Matrix Composites In Sodium Chloride Solutions

    Directory of Open Access Journals (Sweden)

    Jayaprakash

    2017-08-01

    Full Text Available The present investigation aims to evaluate the corrosion characteristics of red mud metal in sodium chloride solution. Metal matrix composites MMC are heterogeneous systems containing matrix and reinforcement. Matrix may be alloy or metal or polymer. Reinforcement may be particulate or fiber or whisker. Their physical and mechanical properties can be tailored according to requirement. They are used in automobile aircraft and marine industries because of their increased corrosion resistance. In this study weight loss corrosion tests and Potestiodynamic polarization studies by using potestiostat are conducted on ZA-27 Red Mud metal matrix composites in different concentrated sodium chloride solutions. Both matrix and reinforcements are commercially available. Composites are prepared by liquid melt metallurgy technique using vortex method. Composites containing 2 4 and 6 percent of preheated but uncoated red mud are prepared. Cylindrical specimens and rectangular specimens are machined. Studies are carried out in 0.035 0.352 and 3.5 solutions of sodium chloride. In all the tests the composites were less prone to corrosion than the matrix. Hence the composites can be used in the marine environment.

  8. Corrosive characteristics of surface-modified stainless steel bipolar plate in solid polymer fuel cell

    Science.gov (United States)

    Zhang, Xiaowen; Wang, Lixia; Sun, Juncai

    2015-03-01

    In this paper, corrosion behavior of an AISI 304 stainless steel modified by niobium or niobium nitride (denoted as niobized 304 SS and Nb-N 304 SS, respectively) is investigated in simulated solid polymer fuel cell (SPFC) operating conditions. Potentiodynamic polarizations show that the corrosion potentials of surface modified 304 SS shift to positive direction while the corrosion current densities decrease greatly comparing with the bare 304 SS in simulated anodic SPFC environments. The order of corrosive resistance in corrosive potential, corrosive current density and pitting potential is: Nb-N 304 SS > niobized 304 SS > bare 304 SS. In the methanol-fueled SPFC operating conditions, the results show that the corrosion resistance of bare and niobized 304 SS increases with the methanol concentration increasing in the test solutions.

  9. Corrosion of copper containers prior to saturation of a nuclear fuel waste disposal vault

    International Nuclear Information System (INIS)

    King, F.; Kolar, M.

    1997-12-01

    The buffer material surrounding the containers in a Canadian nuclear fuel waste disposal vault will partially desiccate as a result of the elevated temperature at the container surface. This will lead to a period of corrosion in a moist air atmosphere. Corrosion will either take the form of slow oxidation if the container surface remains dry or aqueous electrochemical corrosion if the surface is wetted by a thin liquid film. The relevant literature is reviewed, from which it is concluded that corrosion should be uniform in nature, except if the surface is wetted, in which case localized corrosion is a possibility. A quantitative analysis of the extent and rate of uniform corrosion during the unsaturated period is presented. Two bounding cases are considered: first, the case of slow oxidation in moist air following either logarithmic or parabolic oxide-growth kinetics and, second, the case of electrochemically based corrosion occurring in a thin liquid film uninhibited by the growth of corrosion products. (author)

  10. Method of making molten carbonate fuel cell ceramic matrix tape

    Science.gov (United States)

    Maricle, Donald L.; Putnam, Gary C.; Stewart, Jr., Robert C.

    1984-10-23

    A method of making a thin, flexible, pliable matrix material for a molten carbonate fuel cell is described. The method comprises admixing particles inert in the molten carbonate environment with an organic polymer binder and ceramic particle. The composition is applied to a mold surface and dried, and the formed compliant matrix material removed.

  11. Pitting corrosion behaviour study of aluminium matrix composites (A3xx.x/SiCp)

    International Nuclear Information System (INIS)

    Pardo, A.; Merino, M. C.; Merino, S.; Lopez, M. D.; Viejo, F.; Carboneras, M.; Arrabal, R.

    2004-01-01

    The influence of the SiCp proportion on the pitting corrosion of A3xx.x/SiC/xxp composites was studies by means of potenciodinamic polarization and double cyclic polarization in saline environment at 25 degree centigrade A360/SiC/xxp matrix does not contain copper, whereas the A380/SiC/xxp matric contains 1,39-1,44 wt %Cu. The kinetic study was carried out by gravimetric measurements. The nature of corrosion products was analysed by low angle XRD and Scanning Electron Microscopy (SEM). The corrosion is due to nucleation and growth of Al 2 O 3 -3H 2 O on the material surface. The corrosion increases with the reinforcement proportion, chloride concentration and copper content. (Author) 10 refs

  12. A copper container corrosion model for the in-room emplacement of used CANDU fuel

    International Nuclear Information System (INIS)

    King, F.

    1996-11-01

    Copper containers in a Canadian nuclear fuel waste disposal vault are expected to undergo uniform corrosion and, possibly, pitting. The corrosion behaviour of the containers will be dictated by the evolution of environmental conditions within the disposal vault. The environment will evolve from an early warm, oxidizing phase, during which fast uniform corrosion and pitting may occur, to an indefinite period of cool, anoxic conditions, during which the container will only be susceptible to slow uniform corrosion. The results of corrosion and electrochemical studies of the uniform corrosion of Cu in O 2 -containing Cl - solutions are discussed and a detailed reaction mechanism presented. The relevant literature on pitting corrosion is briefly reviewed and models for the prediction of pit depth discussed. The potential for microbially influenced corrosion and stress-corrosion cracking is discussed, as are vapour-phase corrosion and the effects of β-radiation. The use of natural analogues for justifying long-term corrosion predictions is also considered. Finally, a model for uniform corrosion and pitting is presented and container lifetimes predicted. Copper containers having a minimum wall thickness of 25.4 mm are not predicted to fail by corrosion in periods 6 a. Thus, despite the assumption of poor rock quality made here, the safety of the entire disposal concept can be assured by the use of a long-lived container. (author). 125 refs., 1 tab., 24 figs

  13. Colloids from the aqueous corrosion of uranium nuclear fuel

    Science.gov (United States)

    Kaminski, M. D.; Dimitrijevic, N. M.; Mertz, C. J.; Goldberg, M. M.

    2005-12-01

    Colloids may enhance the subsurface transport of radionuclides and potentially compromise the long-term safe operation of the proposed radioactive waste repository at Yucca Mountain. Little data is available on colloid formation for the many different waste forms expected to be buried in the repository. This work expands the sparse database on colloids formed during the corrosion of metallic uranium nuclear fuel. We characterized spherical UO 2 and nickel-rich montmorilonite smectite-clay colloids formed during the corrosion of uranium metal fuel under bathtub conditions at 90 °C. Iron and chromium oxides and calcium carbonate colloids were present but were a minor population. The estimated upper concentration of the UO 2 and clays was 4 × 10 11 and 7 × 10 11-3 × 10 12 particles/L, respectively. However, oxygen eventually oxidized the UO 2 colloids, forming long filaments of weeksite K 2(UO 2) 2Si 6O 15 · 4H 2O that settled from solution, reducing the UO 2 colloid population and leaving predominantly clay colloids. The smectite colloids were not affected by oxygen. Plutonium was not directly observed within the UO 2 colloids but partitioned completely to the colloid size fraction. The plutonium concentration in the colloidal fraction was slightly higher than the value used in the viability assessment model, and does not change in concentration with exposure to oxygen. This paper provides conclusive evidence for single-phase radioactive colloids composed of UO 2. However, its impact on repository safety is probably small since oxygen and silica availability will oxidize and effectively precipitate the UO 2 colloids from concentrated solutions.

  14. Corrosion behaviour of aluminium matrix composites containing silicon carbide particles

    Energy Technology Data Exchange (ETDEWEB)

    Harris, S.J. [Nottingham Univ. (United Kingdom). Dept. of Materials Engineering and Materials Design; Noble, B. [Nottingham Univ. (United Kingdom). Dept. of Materials Engineering and Materials Design; Trowsdale, A.J. [Nottingham Univ. (United Kingdom). Dept. of Materials Engineering and Materials Design

    1996-12-31

    An examination of the pitting attack in two aluminium matrix composites (1050 and 2124) each reinforced with varying fractions (0-30 wt.%) of silicon carbide particles (SiC{sub p}) in the size range 3-40 {mu}m has been made in 1 N NaCl solution. It has been demonstrated that the existence of pores and crevices at SiC{sub p}/matrix interfaces strongly influences pit initiation. This is further aided by the cracking of large SiC{sub p}{>=}20 {mu}m, during processing. The presence of {theta}(CuAl{sub 2}) and S(CuMgAl{sub 2}) precipitates in 2124-SiC{sub p} composite also promotes pitting attack at SiC{sub p}-matrix and intermetallic-matrix interfaces. (orig.)

  15. 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

  16. 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.

  17. Corrosion Behaviour of Carbon Steel in Biodiesel–Diesel–Ethanol (BDE Fuel Blend

    Directory of Open Access Journals (Sweden)

    Thangavelu Saravana Kannan

    2015-01-01

    Full Text Available The biodiesel–diesel–ethanol blend represents an important alternative fuel for diesel engines; however, changes in the fuel composition and the introduction of new alternative fuel often results in corrosion and degradation of the automobile fuel system parts. In this present study, the corrosion behavior of carbon steel in B20D70E10 (biodiesel 20%, diesel 70% and ethanol 10% fuel blend was studied by static immersion at room temperature and 60 °C. The effect of B20D70E10 fuel blend on corrosion rate, morphology of corrosion products, and chemical structure of carbon steel were studied. In addition, the change of fuel properties, namely, total acid number, density, viscosity, calorific value, flash point, and color changes were also investigated. Moreover, fuel compositional changes, such as water content and oxidation product level in the fuel blends were examined. The results showed that the degradation of fuel properties and corrosion rate of carbon steel in B20D70E10 are lower than neat biodiesel (B100, whereas slightly higher than petro-diesel (B0

  18. 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

  19. Development of corrosion resistant materials for an electrolytic reduction process of a spent nuclear fuel

    International Nuclear Information System (INIS)

    Jong-Hyeon Lee; Soo-Haeng Cho; Jeong-Gook Oh; Eung-Ho Kim

    2008-01-01

    New alloys were designed and prepared to improve their corrosion resistance in an electrolytic reduction environment for a spent oxide fuel on the basis of a thermodynamical assessment. A considerable solubility of Si was confirmed in the Ni alloys and their corrosion resistance was drastically increased with the addition of Si. It was confirmed that a protective oxide layer was formed during a corrosion test due to a reaction among the alloying elements such as Cr, Al and Si. (authors)

  20. Impact of Aluminum on Anticipated Corrosion in a Flooded spent nuclear fuel Multi -Canister Overpack

    International Nuclear Information System (INIS)

    DUNCAN, D.R.

    1999-01-01

    Corrosion reactions in a flooded MCO are examined to determine the impact of aluminum corrosion products (from aluminum basket grids and spacers) on bound water estimates and subsequent fuel/environment reactions during storage. The mass and impact of corrosion products were determined to be insignificant, validating the choice of aluminum as an MCO component and confirming expectations that no changes to the Technical Databook or particulate mass or water content are necessary

  1. Design for Corrosion Control of Aviation Fuel Storage and Distribution Systems

    Science.gov (United States)

    1975-06-01

    Aluminum alloys such as 5052 , 5454, 5456, 5083, 5086, 6061 or 6063 are used for the distribution systems for thermally stable (JP-7) aviation fuels. They...ii 1 700-2000 Corrosive 2 2000-6000 Moderately corrosive 3 >6000 Mildly to noncorrosive 4 The corrosivity of soils with respect to aluminum alloys is...dependent upon resistivity as listed above. However, aluminum alloys are very sensitive to the concentration of chloride ion in the soil, pH and

  2. Characterization of graphite-matrix pulsed reactor fuels

    International Nuclear Information System (INIS)

    Karnes, C.H.; Marion, R.H.

    1976-01-01

    The performance of the Annular Core Pulsed Reactor (ACPR) is being upgraded in order to accommodate higher fluence experiments for fast reactor fuel element transient and safety studies. The increased fluence requires a two-zone core with the inner zone containing fuel having a high enthalpy and the capability of withstanding very high temperatures during both pulsed and steady state operation. Because the fuel is subjected to a temperature risetime of 2 to 5 ms and to a large temperature difference across the diameter, fracture due to thermal stresses is the primary failure mode. One of the fuels considered for the high enthalpy inner region is a graphite-matrix fuel containing a dispersion of uranium--zirconium carbide solid solution particles. A program was initiated to optimize the development of this class of fuel. This summary presents results on formulations of fuel which have been fabricated by the Materials Technology Group of the Los Alamos Scientific Laboratory

  3. Metal Matrix Microencapsulated Fuel Technology for LWR Applications

    International Nuclear Information System (INIS)

    Terrani, Kurt A.; Bell, Gary L.; Kiggans, Jim; Snead, Lance Lewis

    2012-01-01

    An overview of the metal matrix microencapsulated (M3) fuel concept for the specific LWR application has been provided. Basic fuel properties and characteristics that aim to improve operational reliability, enlarge performance envelope, and enhance safety margins under design-basis accident scenarios are summarized. Fabrication of M3 rodlets with various coated fuel particles over a temperature range of 800-1300 C is discussed. Results from preliminary irradiation testing of LWR M3 rodlets with surrogate coated fuel particles are also reported.

  4. Part I. Corrosion studies of continuous alumina fiber reinforced aluminum-matrix composites. Part II. Galvanic corrosion between continuous alumina fiber reinforced aluminum-matrix composites and 4340 steel

    Science.gov (United States)

    Zhu, Jun

    Part I. The corrosion performance of continuous alumina fiber reinforced aluminum-matrix composites (CF-AMCs) was investigated in both the laboratory and field environments by comparing them with their respective monolithic matrix alloys, i.e., pure Al, A1-2wt%Cu T6, and Al 6061 T6. The corrosion initiation sites were identified by monitoring the changes in the surface morphology. Corrosion current densities and pH profiles at localized corrosion sites were measured using the scanning-vibrating electrode technique and the scanning ion-selective electrode technique, respectively. The corrosion damage of the materials immersed in various electrolytes, as well as those exposed in a humidity chamber and outdoor environments, was evaluated. Potentiodynamic polarization behavior was also studied. The corrosion initiation for the composites in 3.15 wt% NaCl occurred primarily around the Fe-rich intermetallic particles, which preferentially existed around the fiber/matrix interface on the composites. The corrosion initiation sites were also caused by physical damage (e.g., localized deformation) to the composite surface. At localized corrosion sites, the buildup of acidity was enhanced by the formation of micro-crevices resulting from fibers left in relief as the matrix corroded. The composites that were tested in exposure experiments exhibited higher corrosion rates than their monolithic alloys. The composites and their monolithic alloys were subjected to pitting corrosion when anodically polarized in the 3.15 wt% NaCl, while they passivated when anodically polarized in 0.5 M Na2SO4. The experimental results indicated that the composites exhibited inferior corrosion resistance compared to their monolithic matrix alloys. Part II. Galvanic corrosion studies were conducted on CF-AMCs coupled to 4340 steel since CF-AMCs have low density and excellent mechanical properties and are being considered as potential jacketing materials for reinforcing steel gun barrels. Coupled and

  5. Corrosion on the fuel plate nucleus based on U3 O8 - Al dispersions

    International Nuclear Information System (INIS)

    Durazzo, M.

    2005-01-01

    Samples of MTR type U 3 O 8 - Al dispersion fuel plates meats were corrosion tested in deionized water at different temperatures in the range 30 to 90 deg C. In the tests the cores were exposed to the deionized water by means of an artificially produced cladding defect. The results indicate that the meat corrosion is accompanied by hydrogen evolution. (author)

  6. Corrosion resistance of metallic materials for use in nuclear fuel reprocessing

    International Nuclear Information System (INIS)

    Pinard Legry, G.; Pelras, M.; Turluer, G.

    1988-01-01

    The main of this review is to reassess the corrosion resistance properties required from metallic materials to be used in the various developments of the Purex process applied for nuclear fuel reprocessing

  7. Corrosion of research reactor aluminium clad spent fuel in water. Additional information

    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

  8. Statistical analysis of failure time in stress corrosion cracking of fuel tube in light water reactor

    International Nuclear Information System (INIS)

    Hirao, Keiichi; Yamane, Toshimi; Minamino, Yoritoshi

    1991-01-01

    This report is to show how the life due to stress corrosion cracking breakdown of fuel cladding tubes is evaluated by applying the statistical techniques to that examined by a few testing methods. The statistical distribution of the limiting values of constant load stress corrosion cracking life, the statistical analysis by making the probabilistic interpretation of constant load stress corrosion cracking life, and the statistical analysis of stress corrosion cracking life by the slow strain rate test (SSRT) method are described. (K.I.)

  9. Corrosion of the wet-seal area in molten carbonate fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Donado, R.A.; Marianowski, L.G.; Maru, H.C.; Selman, J.R.

    1984-11-01

    Severe corrosion can occur in the seal area between the cell housing and the electrolyte tile of a molten carbonate fuel cell even if an optimal selection of the active cell materials has been made. Thermodynamic calculations and an approximate mixed-potential analysis indicate that several minor corrosion cells are active at open circuit, and a major corrosion cell is established when the cell is under load. Extensive corrosion can be expected on the anode side of the wet seal, although effects at the cathode side are minor.

  10. Characterization of uranium corrosion products involved in the March 13, 1998 fuel manufacturing facility pyrophoric event

    International Nuclear Information System (INIS)

    Totemeier, T.C.

    1999-01-01

    Uranium metal corrosion products from ZPPR fuel plates involved in the March 13, 1998 pyrophoric event in the Fuel Manufacturing Facility at Argonne National Laboratory-West were characterized using thermo-gravimetric analysis, X-ray diffraction, and BET gas sorption techniques. Characterization was performed on corrosion products in several different conditions: immediately after separation from the source metal, after low-temperature passivation, after passivation and extended vault storage, and after burning in the pyrophoric event. The ignition temperatures and hydride fractions of the corrosion product were strongly dependent on corrosion extent. Corrosion products from plates with corrosion extents less than 0.7% did not ignite in TGA testing, while products from plates with corrosion extents greater than 1.2% consistently ignited. Corrosion extent is defined as mass of corrosion products divided by the total mass of uranium. The hydride fraction increased with corrosion extent. There was little change in corrosion product properties after low-temperature passivation or vault storage. The burned products were not reactive and contained no hydride; the principal constituents were UO 2 and U 3 O 7 . The source of the event was a considerable quantity of reactive hydride present in the corrosion products. No specific ignition mechanism could be conclusively identified. The most likely initiator was a static discharge in the corrosion product from the 14th can as it was poured into the consolidation can. The available evidence does not support scenarios in which the powder in the consolidation can slowly self-heated to the ignition point, or in which the powder in the 14th can was improperly passivated

  11. Characterization of uranium corrosion products involved in the March 13, 1998 fuel manufacturing facility pyrophoric event.

    Energy Technology Data Exchange (ETDEWEB)

    Totemeier, T.C.

    1999-04-26

    Uranium metal corrosion products from ZPPR fuel plates involved in the March 13, 1998 pyrophoric event in the Fuel Manufacturing Facility at Argonne National Laboratory-West were characterized using thermo-gravimetric analysis, X-ray diffraction, and BET gas sorption techniques. Characterization was performed on corrosion products in several different conditions: immediately after separation from the source metal, after low-temperature passivation, after passivation and extended vault storage, and after burning in the pyrophoric event. The ignition temperatures and hydride fractions of the corrosion product were strongly dependent on corrosion extent. Corrosion products from plates with corrosion extents less than 0.7% did not ignite in TGA testing, while products from plates with corrosion extents greater than 1.2% consistently ignited. Corrosion extent is defined as mass of corrosion products divided by the total mass of uranium. The hydride fraction increased with corrosion extent. There was little change in corrosion product properties after low-temperature passivation or vault storage. The burned products were not reactive and contained no hydride; the principal constituents were UO{sub 2} and U{sub 3}O{sub 7}. The source of the event was a considerable quantity of reactive hydride present in the corrosion products. No specific ignition mechanism could be conclusively identified. The most likely initiator was a static discharge in the corrosion product from the 14th can as it was poured into the consolidation can. The available evidence does not support scenarios in which the powder in the consolidation can slowly self-heated to the ignition point, or in which the powder in the 14th can was improperly passivated.

  12. Galvanic corrosion of Mg-Zr fuel cladding and steel immobilized in Portland cement and geopolymer at early ages

    Science.gov (United States)

    Rooses, Adrien; Lambertin, David; Chartier, David; Frizon, Fabien

    2013-04-01

    Galvanic corrosion behaviour of Mg-Zr alloy fuel cladding and steel has been studied in Ordinary Portland cement and Na-geopolymer. Portland cements implied the worse magnesium corrosion performances due to the negative effects of cement hydrates, grinding agents and gypsum on the galvanic corrosion. Galvanic corrosion in Na-geopolymer paste remains very low. Silicates and fluoride from the geopolymer activation solution significantly improve the corrosion resistance of the magnesium alloy while coupling with a cathode.

  13. Optical matrix for nuclear fuel assemblies

    International Nuclear Information System (INIS)

    Romero G, M.; Gonzaga O, A.

    1996-01-01

    In order to detect the presence of fuel rods, it was selected a reflection optical transducer, which provides a measurable electrical signal when the beam at a certain distance is interrupted then there is a reflection causing a excitation to the sensor that provides a change of state at the output of transducer. This step is coupled through an operational amplifier which drives the opto coupler circuit isolating this step of the interface and a personal computer. This work presents the description of components, designs, signal coupler and opto isolater circuit, interface circuit and tutorial assemble program. (Author)

  14. Corrosion surveillance programme for Latin American research reactor Al-clad spent fuel in water

    International Nuclear Information System (INIS)

    Ramanathan, L.V.; Haddad, R.; Ritchie, I.

    2002-01-01

    The objectives of the IAEA sponsored Regional Technical Co-operation Project for Latin America (Argentina, Brazil, Chile, Mexico, and Peru) are to provide the basic conditions to define a regional strategy for managing spent fuel and to provide solutions, taking into consideration the economic and technological realities of the countries involved. In particular, to determine the basic conditions for managing research reactor spent fuel during operation and interim storage as well as final disposal, and to establish forms of regional cooperation in the four main areas: spent fuel characterization, safety, regulation and public communication. This paper reports the corrosion surveillance activities of the Regional Project and these are based on the IAEA sponsored co-ordinated research project (CRP) on 'Corrosion of research reactor Al-clad spent fuel in water'. The overall test consists of exposing corrosion coupon racks at different spent fuel basins followed by evaluation. (author)

  15. Corrosion

    Science.gov (United States)

    Slabaugh, W. H.

    1974-01-01

    Presents some materials for use in demonstration and experimentation of corrosion processes, including corrosion stimulation and inhibition. Indicates that basic concepts of electrochemistry, crystal structure, and kinetics can be extended to practical chemistry through corrosion explanation. (CC)

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

    International Nuclear Information System (INIS)

    Baney, Ronald

    2008-01-01

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

  17. Corrosion of aluminum-clad alloys in wet spent fuel storage

    International Nuclear Information System (INIS)

    Howell, J.P.

    1995-09-01

    Large quantities of Defense related spent nuclear fuels are being stored in water basins around the United States. Under the non-proliferation policy, there has been no processing since the late 1980's and these fuels are caught in the pipeline awaiting processing or other disposition. At the Savannah River Site, over 200 metric tons of aluminum clad fuel are being stored in four water filled basins. Some of this fuel has experienced significant pitting corrosion. An intensive effort is underway at SRS to understand the corrosion problems and to improve the basin storage conditions for extended storage requirements. Significant improvements have been accomplished during 1993-1995, but the ultimate solution is to remove the fuel from the basins and to process it to a more stable form using existing and proven technology. This report presents a discussion of the fundamentals of aluminum alloy corrosion as it pertains to the wet storage of spent nuclear fuel. It examines the effects of variables on corrosion in the storage environment and presents the results of corrosion surveillance testing activities at SRS, as well as other fuel storage basins within the Department of Energy production sites

  18. Corrosion surveillance program of aluminum spent fuel elements in wet storage sites

    International Nuclear Information System (INIS)

    Linardi, E; Haddad, R

    2012-01-01

    Due to different degradation issues observed in aluminum-clad spent fuel during long term storage in water, the IAEA implemented in 1996 a Coordinated Research Project (CRP) and a Regional Project for Latin America, on Corrosion of Research Reactor Aluminum Clad Spent Fuel in Water. Argentine has been among the participant countries of these projects, carrying out spent fuel corrosion surveillance activities in its storage facilities. As a result of the research a large database on corrosion of aluminum-clad fuel has been generated. It was determined that the main types of corrosion affecting the spent fuel are pitting and galvanic corrosion due to contact with stainless steel. It was concluded that the quality of the water is the critical factor to control in a spent fuel storage facility. Another phase of the program is being conducted currently, which began in 2011 with the immersion of test racks in the RA1 reactor pool, and in the Research Reactor Spent Fuel Storage Facility (FACIRI), located in Ezeiza Atomic Center. This paper presents the results of the chemical analysis of the water performed so far, and its relationship with the examination of the coupons extracted from the sites (author)

  19. Technical investigation of a pyrophoric event involving corrosion products from HEU ZPPR fuel plates

    International Nuclear Information System (INIS)

    Totemeier, T. C.

    2000-01-01

    A pyrophoric event recently occurred which involved corrosion products collected from highly-enriched uranium (HEU) fuel plates used in the Zero Power Physics Reactor (ZPPR). This paper summarizes the event and its background, and presents the results of an investigation into its source and mechanism. The investigation focused on characterization of corrosion product samples similar to those involved in the event using thermo-gravimetric analysis (TGA). Burning curve TGA tests were performed to measure the ignition temperature and hydride fractions of corrosion products in several different conditions to assess the effects of passivation treatment and long-term storage on chemical reactivity. The hydride fraction and ignition temperature of the corrosion products were found to be strongly dependent on the corrosion extent of the source metal. The results indicate that the energy source for the event was a considerable quantity of uranium hydride present in the corrosion products, but the specific ignition mechanism could not be identified

  20. High Temperature Corrosion Problem of Boiler Components in presence of Sulfur and Alkali based Fuels

    Science.gov (United States)

    Ghosh, Debashis; Mitra, Swapan Kumar

    2011-04-01

    Material degradation and ageing is of particular concern for fossil fuel fired power plant components. New techniques/approaches have been explored in recent years for Residual Life assessment of aged components and material degradation due to different damage mechanism like creep, fatigue, corrosion and erosion etc. Apart from the creep, the high temperature corrosion problem in a fossil fuel fired boiler is a matter of great concern if the fuel contains sulfur, chlorine sodium, potassium and vanadium etc. This paper discusses the material degradation due to high temperature corrosion in different critical components of boiler like water wall, superheater and reheater tubes and also remedial measures to avoid the premature failure. This paper also high lights the Residual Life Assessment (RLA) methodology of the components based on high temperature fireside corrosion. of different critical components of boiler.

  1. Corrosion of Spent Nuclear Fuel: The Long-Term Assessment

    Energy Technology Data Exchange (ETDEWEB)

    Ewing, Rodney C.

    2003-09-14

    public that there is a reasonable basis for accepting the long-term extrapolations of spent fuel behavior. In recent years ''natural analogues'' for both the repository environment (e.g., the Oklo natural reactors) and nuclear waste form behavior (e.g., corrosion and alteration of uraninite, UO{sub 2+x}) have been cited as a fundamental means of achieving confirmation of long-term extrapolations. In particular, considerable effort has already been made to establish that uraninite, UO{sub 2+x}, with its impurities, is a good structural and chemical analogue for the analysis of the long-term behavior of the UO{sub 2} in spent nuclear fuel. This proposal is based on the study of uraninite and the naturally occurring alteration products of UO{sub 2+x} under oxidizing and reducing conditions. The UO{sub 2} in spent nuclear fuel is not stable under oxidizing conditions. In oxic solutions, uranium has a strong tendency to exist as U{sup 6+} in the uranyl molecule, UO{sub 2}{sup 2+}. Uranyl ions react with a wide variety of inorganic and organic anions to form complexes. Throughout most of the natural range of pH, U{sup 6+} forms strong complexes with oxygen-bearing ions like CO{sub 3}{sup 2-}, HCO{sup 3-}, SO{sub 4}{sup 2-}, PO{sub 4}{sup 3-}, and AsO{sub 4}{sup 3-}, which are present in most oxidized stream and subsurface waters. In arid environments, the U{sup 6+} ion can precipitate as a wide variety of uranyl oxide hydrates, uranyl silicates and uranyl phosphates. This is well demonstrated in experimental work, e.g., in long term drip tests on UO{sub 2} and is confirmed by natural occurrences of UO{sub 2} in which a wide variety of uranyl phases form as alteration products. The most striking feature of these studies is the very close parallel in the paragenetic sequences (i.e. phase formation sequence) between the very long term (10 year tests) and the young (therefore, low-Pb uraninites) of the Nopal I deposit in Mexico.

  2. Spent nuclear fuel project recommended reaction rate constants for corrosion of N-Reactor fuel

    International Nuclear Information System (INIS)

    Cooper, T.D.; Pajunen, A.L.

    1998-01-01

    The US Department of Energy (DOE) established the Spent Nuclear Fuel Project (SNF Project) to address safety and environmental concerns associated with deteriorating spent nuclear fuel presently stored in the Hanford Site's K Basins. The SNF Project has been tasked by the DOE with moving the spent N-Reactor fuel from wet storage to contained dry storage in order to reduce operating costs and environmental hazards. The chemical reactivity of the fuel must be understood at each process step and during long-term dry storage. Normally, the first step would be to measure the N-fuel reactivity before attempting thermal-hydraulic transfer calculations; however, because of the accelerated project schedule, the initial modeling was performed using literature values for uranium reactivity. These literature values were typically found for unirradiated, uncorroded metal. It was fully recognized from the beginning that irradiation and corrosion effects could cause N-fuel to exhibit quite different reactivities than those commonly found in the literature. Even for unirradiated, uncorroded uranium metal, many independent variables affect uranium metal reactivity resulting in a wide scatter of data. Despite this wide reactivity range, it is necessary to choose a defensible model and estimate the reactivity range of the N-fuel until actual reactivity can be established by characterization activities. McGillivray, Ritchie, and Condon developed data and/or models that apply for certain samples over limited temperature ranges and/or reaction conditions (McGillivray 1994, Ritchie 1981 and 1986, and Condon 1983). These models are based upon small data sets and have relatively large correlation coefficients

  3. LWRS Fuels Pathway: Engineering Design and Fuels Pathway Initial Testing of the Hot Water Corrosion System

    Energy Technology Data Exchange (ETDEWEB)

    Dr. John Garnier; Dr. Kevin McHugh

    2012-09-01

    The Advanced LWR Nuclear Fuel Development R&D pathway performs strategic research focused on cladding designs leading to improved reactor core economics and safety margins. The research performed is to demonstrate the nuclear fuel technology advancements while satisfying safety and regulatory limits. These goals are met through rigorous testing and analysis. The nuclear fuel technology developed will assist in moving existing nuclear fuel technology to an improved level that would not be practical by industry acting independently. Strategic mission goals are to improve the scientific knowledge basis for understanding and predicting fundamental nuclear fuel and cladding performance in nuclear power plants, and to apply this information in the development of high-performance, high burn-up fuels. These will result in improved safety, cladding, integrity, and nuclear fuel cycle economics. To achieve these goals various methods for non-irradiated characterization testing of advanced cladding systems are needed. One such new test system is the Hot Water Corrosion System (HWCS) designed to develop new data for cladding performance assessment and material behavior under simulated off-normal reactor conditions. The HWCS is capable of exposing prototype rodlets to heated, high velocity water at elevated pressure for long periods of time (days, weeks, months). Water chemistry (dissolved oxygen, conductivity and pH) is continuously monitored. In addition, internal rodlet heaters inserted into cladding tubes are used to evaluate repeated thermal stressing and heat transfer characteristics of the prototype rodlets. In summary, the HWCS provides rapid ex-reactor evaluation of cladding designs in normal (flowing hot water) and off-normal (induced cladding stress), enabling engineering and manufacturing improvements to cladding designs before initiation of the more expensive and time consuming in-reactor irradiation testing.

  4. Corrosion of aluminum alloy 2024 by microorganisms isolated from aircraft fuel tanks.

    Science.gov (United States)

    McNamara, Christopher J; Perry, Thomas D; Leard, Ryan; Bearce, Ktisten; Dante, James; Mitchell, Ralph

    2005-01-01

    Microorganisms frequently contaminate jet fuel and cause corrosion of fuel tank metals. In the past, jet fuel contaminants included a diverse group of bacteria and fungi. The most common contaminant was the fungus Hormoconis resinae. However, the jet fuel community has been altered by changes in the composition of the fuel and is now dominated by bacterial contaminants. The purpose of this research was to determine the composition of the microbial community found in fuel tanks containing jet propellant-8 (JP-8) and to determine the potential of this community to cause corrosion of aluminum alloy 2024 (AA2024). Isolates cultured from fuel tanks containing JP-8 were closely related to the genus Bacillus and the fungi Aureobasidium and Penicillium. Biocidal activity of the fuel system icing inhibitor diethylene glycol monomethyl ether is the most likely cause of the prevalence of endospore forming bacteria. Electrochemical impedance spectroscopy and metallographic analysis of AA2024 exposed to the fuel tank environment indicated that the isolates caused corrosion of AA2024. Despite the limited taxonomic diversity of microorganisms recovered from jet fuel, the community has the potential to corrode fuel tanks.

  5. The corrosion of aluminum-clad spent nuclear fuel in wet basin storage

    International Nuclear Information System (INIS)

    Howell, J.P.; Burke, S.D.

    1996-01-01

    Large quantities of Defense related spent nuclear fuels are being stored in water basins around the United States. Under the non-proliferation policy, there has been no processing since the late 1980's and these fuels are caught in the pipeline awaiting stabilization or other disposition. At the Savannah River Site, over 200 metric tons of aluminum clad fuel are being stored in four water filled basins. Some of this fuel has experienced visible pitting corrosion. An intensive effort is underway at SRS to understand the corrosion problems and to improve the basin storage conditions for extended storage requirements. Significant improvements have been accomplished during 1993-1996. This paper presents a discussion of the fundamentals of aluminum alloy corrosion as it pertains to the wet storage of spent nuclear fuel. It examines the effects of variables on corrosion in the storage environment and presents the results of corrosion surveillance testing activities at SRS, as well as discussions of fuel storage basins at other production sites of the Department of Energy

  6. The corrosion of aluminum-clad spent nuclear fuel in wet basin storage

    Energy Technology Data Exchange (ETDEWEB)

    Howell, J.P.; Burke, S.D.

    1996-02-20

    Large quantities of Defense related spent nuclear fuels are being stored in water basins around the United States. Under the non-proliferation policy, there has been no processing since the late 1980`s and these fuels are caught in the pipeline awaiting stabilization or other disposition. At the Savannah River Site, over 200 metric tons of aluminum clad fuel are being stored in four water filled basins. Some of this fuel has experienced visible pitting corrosion. An intensive effort is underway at SRS to understand the corrosion problems and to improve the basin storage conditions for extended storage requirements. Significant improvements have been accomplished during 1993-1996. This paper presents a discussion of the fundamentals of aluminum alloy corrosion as it pertains to the wet storage of spent nuclear fuel. It examines the effects of variables on corrosion in the storage environment and presents the results of corrosion surveillance testing activities at SRS, as well as discussions of fuel storage basins at other production sites of the Department of Energy.

  7. Effect of reactor chemistry and operating variables on fuel cladding corrosion in PWRs

    International Nuclear Information System (INIS)

    Park, Moon Ghu; Lee, Sang Hee

    1997-01-01

    As the nuclear industry extends the fuel cycle length, waterside corrosion of zircaloy cladding has become a limiting factor in PWR fuel design. Many plant chemistry factors such as, higher lithium/boron concentration in the primary coolant can influence the corrosion behavior of zircaloy cladding. The chemistry effect can be amplified in higher duty fuel, particularlywhen surface boiling occurs. Local boiling can result in increased crud deposition on fuel cladding which may induce axial power offset anomalies (AOA), recently reported in several PWR units. In this study, the effect of reactor chemistry and operating variables on Zircaloy cladding corrosion is investigated and simulation studies are performed to evaluate the optimal primary chemistry condition for extended cycle operation. (author). 8 refs., 3 tabs., 16 figs

  8. Current state of knowledge of water radiolysis effects on spent nuclear fuel corrosion

    International Nuclear Information System (INIS)

    Christensen, H.; Sunder, S.

    2000-07-01

    Literature data on the effect of water radiolysis products on spent-fuel oxidation and dissolution are reviewed. Effects of gamma radiolysis, alpha radiolysis, and dissolved O 2 or H 2 O 2 in unirradiated solutions are discussed separately. Also, the effect of carbonate in gamma-irradiated solutions and radiolysis effects on leaching of spent fuel are reviewed. In addition, a kinetic model for calculating the corrosion rates of UO 2 in solutions undergoing radiolysis is discussed. The model gives good agreement between calculated and measured corrosion rates in the case of gamma radiolysis and in unirradiated solutions containing dissolved oxygen or hydrogen peroxide. However, the model fails to predict the results of alpha radiolysis. In a recent study , it was shown that the model gave good agreement with measured corrosion rates of spent fuel exposed in deionized water. The applications of radiolysis studies for geologic disposal of used nuclear fuel are discussed. (author)

  9. Evaluation of corrosion on the fuel performance of stainless steel cladding

    Directory of Open Access Journals (Sweden)

    de Souza Gomes Daniel

    2016-01-01

    Full Text Available In nuclear reactors, the use of stainless steel (SS as the cladding material offers some advantages such as good mechanical and corrosion resistance. However, its main advantage is the reduction in the amount of the hydrogen released during loss-of-coolant accident, as observed in the Fukushima Daiichi accident. Hence, research aimed at developing accident tolerant fuels should consider SS as an important alternative to existing materials. However, the available computational tools used to analyze fuel rod performance under irradiation are not capable of assessing the effectiveness of SS as the cladding material. This paper addresses the SS corrosion behavior in a modified fuel performance code in order to evaluate its effect on the global fuel performance. Then, data from the literature concerning to SS corrosion are implemented in the specific code subroutines, and the results obtained are compared to those for Zircaloy-4 (Zy-4 under the same power history. The results show that the effects of corrosion on SS are considerably different from those on Zy-4. The thickness of the oxide layer formed on the SS surface is considerably lower than that formed on Zy-4. As a consequence of this, the global fuel performance of SS under irradiation should be less affected by the corrosion.

  10. Graphitic matrix materials for spherical HTR fuel elements

    International Nuclear Information System (INIS)

    Rind, W.; Schulze, R.E.; Schulze, H.A.

    1981-12-01

    The present report comprises the essential results of material development and irradiation testing of graphitic matrix materials for spherical HTR fuel elements and completes the documentation of the irradiation data for 20 matrix materials (Juel-1702). The main emphasis is given to the matrices A3-3 (standard matrix) and A3-27 (matrix synthesized resin), both of which are being used as structural materials for the fuel elements of the AVR and the THTR respectively. In addition, comparisons are made between 18 A3-variants and the standard matrix A3-3. It is shown that three of the variants come into question as a potential for use. The results described were obtained in the framework of the HTR project 'Hochtemperaturreaktor-Brennstoffkreislauf' (HBK), in which are involved the Gesellschaft fuer Hochtemperaturreaktor-Technik mbH, Hochtemperaturreaktor-Brennelemente GmbH, Hochtemperatur-Reaktorbau GmbH, Kernforschungsanlage Juelich GmbH, NUKEM GmbH, and Sigri Elektrographit GmbH/Ringsdorff-Werke GmbH. The project is sponsored by the 'Bundesministerium fuer Forschung und Technologie' and by the state of 'Nordrhein-Westfalen'. (orig.) [de

  11. Graphitic matrix materials for spherical HTR fuel elements

    International Nuclear Information System (INIS)

    Schulze, R.E.; Schulze, H.A.

    1981-02-01

    The report comprises the graphical documentation of irradiation results on graphitic matrix materials for spherical HTR fuel elements. The plotted results are based on data analyses of the series of exposures in the High Flux Reactor Petten (HFR). The documentation includes information about the changes of - the dimensions - the dynamic modulus of elasticity - the coefficient of thermal expansion of the materials after irradiation with fast neutrons. The irradiation experiments and the data analyses are part of the matrix development and irradiation programme, whose objective, realization and results obtained are summarized. (orig./IHOE) [de

  12. Corrosion of materials in Molten Carbonate Fuel Cells (MCFC): Literature review

    Science.gov (United States)

    1987-01-01

    Molten carbonate fuel cell (MCFC) separator plates are subjected to corrosive environments. This review focuses on the most comprehensive compiled available data from the major government contractors like ERC and UTC, together with a basic review of alloys resistant to oxidation, sulfidation, carburization, and hot corrosion. Results of this review are used in selecting test materials for the current program. A number of metal oxides were examined. This review discusses the tests performed by UTC and ERC on separator plate materials.

  13. GRH 12-01 Fireside Corrosion in Oxy-fuel Combustion Poster 0108

    Energy Technology Data Exchange (ETDEWEB)

    G. R. Holcomb; J. Tylczak; G. H. Meier; B. Lutz; K. Jung; N. Mu; N. M. Yanar; F. S. Pettit; J. Zhu; A. Wise; D. Laughlin; S. Sridhar

    2012-05-20

    The goals are to: (1) Achieve 90% CO{sub 2} capture at no more than a 35% increase in levelized cost of electricity of post-combustion capture for new and existing conventional coal-fired power plants; (2) Provide high-temperature corrosion information to aid in materials development and selection for oxy-fuel combustion; and (3) Identify corrosion mechanism and behavior differences between air- and oxy-firing.

  14. Status of the inert matrix fuel program at PSI

    International Nuclear Information System (INIS)

    Ledergerber, G.; Degueldre, C.; Kasemeyer, U.; Stanculescu, A.; Paratte, J.M.; Chawla, R.

    1997-01-01

    Incineration of plutonium by a once-through cycle in LWRs utilising an inert matrix based fuel may prove to be an attractive way of making use of the energy of fissile plutonium and reducing both the hazard potential and the volumes of the waste. Yttria stabilised zirconia forms a solid solution with oxides of rare earth elements (e.g. erbium, cerium) and some actinides. The small absorption cross section, the excellent stability under irradiation, and the insolubility in acids and water recommends this material as an inert matrix. Neutronics calculations with erbium as burnable poison show that these compositions would be optimal from the reactivity point of view. A fuel element with an improved reactivity behaviour over its life cycle has been designed for possible introduction into a heterogeneous LWR core. (author). 16 refs., 1 tab., 10 figs

  15. Corrosion Problems in Incinerators and Biomass-Fuel-Fired Boilers

    Directory of Open Access Journals (Sweden)

    Deepa Mudgal

    2014-01-01

    Full Text Available Incinerators are widely used to burn the municipal waste, biowaste, wood, straw, and biomedical waste. Combustion of these types of waste results in generation of chlorides of sodium and potassium which may attack the metallic part of the incinerator. In biofuel-fired boilers, similar type of highly corrosive environment is present. Attempt has been made to review the corrosion problems and their solutions as per the available literature.

  16. Corrosion Problems in Incinerators and Biomass-Fuel-Fired Boilers

    OpenAIRE

    Mudgal, Deepa; Singh, Surendra; Prakash, Satya

    2014-01-01

    Incinerators are widely used to burn the municipal waste, biowaste, wood, straw, and biomedical waste. Combustion of these types of waste results in generation of chlorides of sodium and potassium which may attack the metallic part of the incinerator. In biofuel-fired boilers, similar type of highly corrosive environment is present. Attempt has been made to review the corrosion problems and their solutions as per the available literature.

  17. Corrosion of research reactor aluminium-clad spent fuel in water-chemical and microbiological influenced

    International Nuclear Information System (INIS)

    Maksin, T.N.; Dobrijevic, R.P.; Idjakovic, Z.E.; Pesic, M.P.

    2002-01-01

    Spent fuel resulting from 25 years of operating research reactor RA at the Vinca Institute is presently all stored in the temporary spent fuel storage pool. It has been left in the ambient temperature and humidity for more then fifteen years so intensive corrosion processes were notice. We have spent fuel pools under control, after first research coordination meeting (RCM), of the first CRP, by monitoring of physical and chemical parameters of water in the pools, including temperature, pH-factor, electrical conductivity, mass concentration of corrosion products in the water and mud, mass concentration of relevant ions etc. The rack of standard corrosion coupons, was given at that time, has been in poor quality water for six years. We pick up rack assembly from basin and analysed. The results of this investigation are present in this article. (author)

  18. Current state of knowledge in radiolysis effects on spent fuel corrosion

    International Nuclear Information System (INIS)

    Christensen, H.; Sunder, S.

    1998-09-01

    Literature data on the effect of water radiolysis products on spent fuel oxidation and dissolution have been reviewed. Effects of γ-radiolysis, α-radiolysis and dissolved O 2 or H 2 O 2 in unirradiated solutions have been discussed separately. Also the effect of carbonate in γ-irradiated solutions and radiolysis effects on leaching of spent fuels have been reviewed. In addition a radiolysis model for calculation of corrosion rates of UO 2 , presented previously, has been discussed. The model has been shown to give a good agreement between calculated and measured corrosion rates in the case of γ-radiolysis and in unirradiated solutions of dissolved oxygen or hydrogen peroxide. The model has failed to predict the results of α-radiolysis. In a recent study it was shown that the model gave a good agreement with measured corrosion rates of spent fuel exposed in deionized water

  19. Microbiologically induced corrosion of aluminum alloys in fuel-oil/aqueous system.

    Science.gov (United States)

    Yang, S S; Lin, J Y; Lin, Y T

    1998-09-01

    To investigate the microbiologically induced corrosion of aluminum alloys in fuel-oil/aqueous system, aluminum alloys A356, AA 5052, AA 5083 and AA 6061 were chosen as the test alloys and Cladosporium and several fuel-oil contaminated microbes isolated in Taiwan were used as test organisms. Aluminum alloy AA 5083 in fuel-oil/aqueous system was the most susceptible material for microbial corrosion, then followed by aluminum alloys AA 5052 and A356, and AA 6061 was more resistant to microbial aggression. Mixed culture had high capability of corrosion, then followed by Penicillium sp. AM-F5, Fusarium sp. AM-F1, Pseudomonas aeruginosa AM-B5, Ps. fluorescens AM-B9, C. resinae ATCC 22712, Penicillium sp. AM-F2, Candida sp. AM-Y1 and Ps. aeruginosa AM-B11. From energy dispersive spectrometer analysis, aluminum and magnesium contents decreased in the corrosion area, while chlorine and sulfur contents increased. The major organic acid produced in fuel-oil/aqueous system was acetic acid, and the total organic acids content had a positive correlation with the degree of microbial corrosion.

  20. Corrosion of metal bipolar plates for PEM fuel cells: A review

    Energy Technology Data Exchange (ETDEWEB)

    Antunes, Renato A. [Engenharia de Materiais, Universidade Federal do ABC (UFABC), 09210-170 Santo Andre, SP (Brazil); Oliveira, Mara Cristina L.; Ett, Gerhard; Ett, Volkmar [Electrocell Ind. Com. Equip. Elet. LTDA, Centro de Inovacao, Empreendedorismo e Tecnologia (CIETEC), 05508-000 Sao Paulo, SP (Brazil)

    2010-04-15

    PEM fuel cells are of prime interest in transportation applications due to their relatively high efficiency and low pollutant emissions. Bipolar plates are the key components of these devices as they account for significant fractions of their weight and cost. Metallic materials have advantages over graphite-based ones because of their higher mechanical strength and better electrical conductivity. However, corrosion resistance is a major concern that remains to be solved as metals may develop oxide layers that increase electrical resistivity, thus lowering the fuel cell efficiency. This paper aims to present the main results found in recent literature about the corrosion performance of metallic bipolar plates. (author)

  1. Surface modification techniques for increased corrosion tolerance of zirconium fuel cladding

    Science.gov (United States)

    Carr, James Patrick, IV

    Corrosion is a major issue in applications involving materials in normal and severe environments, especially when it involves corrosive fluids, high temperatures, and radiation. Left unaddressed, corrosion can lead to catastrophic failures, resulting in economic and environmental liabilities. In nuclear applications, where metals and alloys, such as steel and zirconium, are extensively employed inside and outside of the nuclear reactor, corrosion accelerated by high temperatures, neutron radiation, and corrosive atmospheres, corrosion becomes even more concerning. The objectives of this research are to study and develop surface modification techniques to protect zirconium cladding by the incorporation of a specific barrier coating, and to understand the issues related to the compatibility of the coatings examined in this work. The final goal of this study is to recommend a coating and process that can be scaled-up for the consideration of manufacturing and economic limits. This dissertation study builds on previous accident tolerant fuel cladding research, but is unique in that advanced corrosion methods are tested and considerations for implementation by industry are practiced and discussed. This work will introduce unique studies involving the materials and methods for accident tolerant fuel cladding research by developing, demonstrating, and considering materials and processes for modifying the surface of zircaloy fuel cladding. This innovative research suggests that improvements in the technique to modify the surface of zirconium fuel cladding are likely. Three elements selected for the investigation of their compatibility on zircaloy fuel cladding are aluminum, silicon, and chromium. These materials are also currently being investigated at other labs as alternate alloys and coatings for accident tolerant fuel cladding. This dissertation also investigates the compatibility of these three elements as surface modifiers, by comparing their microstructural and

  2. Influence of processing variables and alloy chemistry on the corrosion behavior of ZIRLO nuclear fuel cladding

    International Nuclear Information System (INIS)

    Comstock, R.J.; Sabol, G.P.; Schoenberger, G.

    1996-01-01

    Variations in the thermal heat treatments used during the fabrication of ZIRLO (Zr-1Nb-1Sn-0.1Fe) fuel clad tubing and in ZIRLO alloy chemistry were explored to develop a further understanding of the relationship between processing, microstructure, and cladding corrosion performance. Heat treatment variables included intermediate tube annealing temperatures as well as a beta-phase heat treatment during the latter stages of the tube reduction schedule. Chemistry variables included deviations in niobium and tin content from the nominal composition. The effects of both heat treatment and chemistry on corrosion behavior were assessed by autoclave tests in both pure and lithiated water and high-temperature steam. Analytical electron microscopy demonstrated that the best out-reactor corrosion performance is obtained for microstructures containing a fine distribution of beta-niobium and Zr-Nb-Fe particles. Deviations from this microstructure, such as the presence of beta-zirconium phase, tend to degrade corrosion resistance. ZIRLO fuel cladding was irradiated in four commercial reactors. In all cases, the microstructure in the cladding included beta-niobium and Zr-Nb-Fe particles. ZIRLO fuel cladding processed with a late-stage beta heat treatment to further refine the second-phase particle size exhibited in-reactor corrosion behavior that was similar to reference ZIRLO cladding. Variations of the in-reactor corrosion behavior of ZIRLO were correlated to tin content, with higher oxide thickness observed in the ZIRLO cladding containing higher tin. The results of these studies indicate that optimum corrosion performance of ZIRLO is achieved by maintaining a uniform distribution of fine second-phase particles and controlled levels of tin

  3. An assessment of alternative diesel fuels: microbiological contamination and corrosion under storage conditions.

    Science.gov (United States)

    Lee, Jason S; Ray, Richard I; Little, Brenda J

    2010-08-01

    Experiments were designed to evaluate the nature and extent of microbial contamination and the potential for microbiologically influenced corrosion of alloys exposed in a conventional high sulfur diesel (L100) and alternative fuels, including 100% biodiesel (B100), ultra-low sulfur diesel (ULSD) and blends of ULSD and B100 (B5 and B20). In experiments with additions of distilled water, all fuels supported biofilm formation. Changes in the water pH did not correlate with observations related to corrosion. In all exposures, aluminum 5052 was susceptible to pitting while stainless steel 304L exhibited passive behavior. Carbon steel exhibited uniform corrosion in ULSD and L100, and passive behavior in B5, B20, and B100.

  4. Corrosion-resistant fuel cladding allow for liquid metal fast breeder reactors

    Science.gov (United States)

    Brehm, Jr., William F.; Colburn, Richard P.

    1982-01-01

    An aluminide coating for a fuel cladding tube for LMFBRs (liquid metal fast breeder reactors) such as those using liquid sodium as a heat transfer agent. The coating comprises a mixture of nickel-aluminum intermetallic phases and presents good corrosion resistance to liquid sodium at temperatures up to 700.degree. C. while additionally presenting a barrier to outward diffusion of .sup.54 Mn.

  5. Gel structure of the corrosion layer on cladding pipes of nuclear fuels

    Czech Academy of Sciences Publication Activity Database

    Medek, Jiří; Weishauptová, Zuzana

    2009-01-01

    Roč. 393, č. 2 (2009), s. 306-310 ISSN 0022-3115 R&D Projects: GA ČR GA106/04/0043 Institutional research plan: CEZ:AV0Z30460519 Keywords : cladding pipes of nuclear fuels * corrosion layer * zirconium alloys Subject RIV: JF - Nuclear Energetics Impact factor: 1.933, year: 2009

  6. Experimental studies of laser-ablated zirconium carbide plasma plumes: Fuel corrosion diagnostic development

    International Nuclear Information System (INIS)

    Wantuck, P.J.; Butt, D.P.; Sappey, A.D.

    1992-01-01

    Understanding the corrosion behavior of nuclear fuel materials, such as refractory carbides, in a high temperature hydrogen environment is critical for several proposed nuclear thermal propulsion (NTP) concepts. Monitoring the fuel corrosion products is important not only for understanding corrosion characteristics, but to assess the performance of an actual, operating nuclear propulsion system as well. In this paper, we describe an experimental study initiated to develop, test, and subsequently utilize non-intrusive, laser-based diagnostics to characterize the gaseous product species which are expected to evolve during the exposure of representative fuel samples to hydrogen. Laser ablation is used to produce high temperature, vapor plumes from solid solution, uranium-free, zirconium carbide (ZrC) forms for probing by other laser diagnostic methods; predominantly laser-induced fluorescence (LIF). We discuss the laser ablation technique, results of plume emission measurements, as well as the use of planar LIF to image both the ZrC plumes and actual NTP fuel corrosion constituents

  7. Corrosion resistance of metallic materials for use in nuclear fuel reprocessing

    International Nuclear Information System (INIS)

    Legry, J.P.; Pelras, M.; Turluer, G.

    1989-01-01

    This paper reviews the corrosion resistance properties required from metallic materials to be used in the various developments of the PUREX process for nuclear fuel reprocessing. Stainless steels, zirconium or titanium base alloys are considered for the various plant components, where nitric acid is the main electrolyte with differing acid and nitrate concentrations, temperature and oxidizing species. (author)

  8. Corrosion of Zircaloy-clad fuel rods in high-temperature PWRs: Measurement of waterside corrosion in North Anna Unit 1

    International Nuclear Information System (INIS)

    Balfour, M.G.; Kilp, G.R.; Comstock, R.J.; McAtee, K.R.; Thornburg, D.R.

    1992-03-01

    Twenty-four peripheral rods and two interior rods from North Anna Unit 1, End-of-Cycle 7, were measured at poolside for waterside corrosion on four-cycle Region 6 assemblies F35 and F66, with rod average burnups of 60 GWD/MTU. Similar measurements were obtained on 24 two-cycle fuel rods from Region 8A assemblies H02 and H10 with average burnups of about 40 GWD/MTU. The Region 6 peripheral rods had been corrosion measured previously after three cycles, at 45 GWD/MTU average burnup. The four-cycle Region 6 fuel rods showed high corrosion, compared to only intermediate corrosion level after three cycles. The accelerated corrosion rate in the fourth cycle was accompanied by extensive laminar cracking and spalling of the oxide film in the thickest regions. The peak corrosion of the two-cycle region 8A rods was 32 μm to 53 μm, with some isolated incipient oxide spalling. In conjunction with the in-reactor corrosion measurements, extensive characterization tests plus long-term autoclave corrosion tests were performed on archive samples of the three major tubing lots represented in the North Anna measurements. The autoclave tests generally showed the same ordering of corrosion by tubing lot as in the reactor; the chief difference between the archive tubing samples was a lower tin content (1.38 percent) for the lot with the lowest corrosion rate compared with a higher tin content (1.58) for the lot with the highest corrosion rate. There was no indication in the autoclave tests of an accelerated rate of corrosion as observed in the reactor

  9. Effect of water α radiolysis on the spent nuclear fuel UO2 matrix alteration

    International Nuclear Information System (INIS)

    Lucchini, J.F.

    2001-01-01

    In the option of long term storage or direct disposal of nuclear spent fuel, it is essential to study the long-term behaviour of the spent fuel matrix (UO 2 ) in water, in presence of ionizing radiations. This work gives some knowledge elements about the impact of aerated water alpha radiolysis on UO 2 alteration. An original experiment method was used in this study. UO 2 /water interfaces were irradiated by an external He 2+ ions beam. The sequential batch dissolution tests on UO 2 samples were performed in aerated deionized water, before, during and after a-irradiation under high fluxes. A corrosion product, identified as hydrated uranium peroxide, was formed on the UO 2 surface. The uranium release was 3 to 4 orders of magnitude higher under irradiation than out of irradiation. The concentrations of the radiolysis products H 2 O 2 and H 3 O + were affected by the uranium oxide surface. They could not only explain the whole uranium release reached during irradiation in water. Leaching experiments on UO X spent fuel samples (with or without the Zircaloy clad) were also performed, in hot cells. The uranium release was relatively small, and H 2 O 2 was not detected in solution. The rates of uranium release in aerated water during one hour were calculated. They were about mg -1 .m -2 .d -1 for spent fuel and for UO 2 , and about g -1 .m -2 .d -1 for UO 2 irradiated by He 2+ ions. The comparison of the results between the two kinds of experiment shows a difference of the behaviour in water between UO 2 irradiated by He 2+ ions and spent fuel. Some hypothesis are given to explain this difference. (author)

  10. Influence of reinforcement proportion and matrix composition on pitting corrosion behaviour of cast aluminium matrix composites (A3xx.x/SiCp)

    International Nuclear Information System (INIS)

    Pardo, A.; Merino, M.C.; Merino, S.; Viejo, F.; Carboneras, M.; Arrabal, R.

    2005-01-01

    The influence of silicon carbide (SiCp) proportion and matrix composition on four aluminium metal matrix composites (A360/SiC/10p, A360/SiC/20p, A380/SiC/10p, A380/SiC/20p) immersed in 1-3.5 wt% NaCl at 22 deg C was investigated by potentiodynamic polarization. The kinetics of the corrosion process was studied on the basis of gravimetric measurements. The nature of corrosion products was analysed by scanning electron microscopy (SEM) and low angle X-ray diffraction (XRD). The corrosion damage in Al/SiCp composites was caused by pitting attack and by nucleation and growth of Al 2 O 3 . 3H 2 O on the material surface. The main attack nucleation sites were the interface region between the matrix and the reinforcement particles. The corrosion process was influenced more by the concentration of alloy elements in the matrix than by the proportion of SiCp reinforcement and saline concentration

  11. Oxidation of ZPPR fuel corrosion products : National Spent Nuclear Fuel program FY 1999 final report

    International Nuclear Information System (INIS)

    Totemeier, T. C.

    1999-01-01

    The oxidation behavior of hydride-bearing corrosion products from uranium metal ZPPR fuel plates was studied in Ar-O 2 , Ar-H 2 O, Ar-O 2 -H 2 O, dry air, and moist air environments. Both isothermal and burning curve tests in the different environments were performed using a thermo-gravimetric analyzer. The effect of pre-oxidation in each environment on subsequent ignition temperature was investigated by performing burning curve tests on samples after isothermal oxidation. Low-temperature oxidation rates in Ar-O 2 and dry air environments were identical. Oxidation rates in moist environments were slightly higher, but the difference was not statistically significant at 95% confidence. Oxygen contamination was suspected to have lowered rates measured in the Ar-H 2 O environment. ignition temperatures measured in air were 10-15 C higher than those measured in Ar-20% O 2 ; the ignition temperatures of samples pre-oxidized in moist gas environments appeared to be slightly lower than those of unreacted samples at equivalent hydride fractions. Burning rates in all environments were linearly dependent on hydride surface area for surface areas less than 200 cm 2 . Burning rates were constant at higher surface areas

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

    International Nuclear Information System (INIS)

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

    1997-01-01

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

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

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  14. Reprocessability of molybdenum and magnesia based inert matrix fuels

    Directory of Open Access Journals (Sweden)

    Ebert Elena L.

    2015-12-01

    Full Text Available This work focuses on the reprocessability of metallic 92Mo and ceramic MgO, which is under investigation for (Pu,MA-oxide (MA = minor actinide fuel within a metallic 92Mo matrix (CERMET and a ceramic MgO matrix (CERCER. Magnesium oxide and molybdenum reference samples have been fabricated by powder metallurgy. The dissolution of the matrices was studied as a function of HNO3 concentration (1-7 mol/L and temperature (25-90°C. The rate of dissolution of magnesium oxide and metallic molybdenum increased with temperature. While the MgO rate was independent of the acid concentration (1-7 mol/L, the rate of dissolution of Mo increased with acid concentration. However, the dissolution of Mo at high temperatures and nitric acid concentrations was accompanied by precipitation of MoO3. The extraction of uranium, americium, and europium in the presence of macro amounts of Mo and Mg was studied by three different extraction agents: tri-n-butylphosphate (TBP, N,Nʹ-dimethyl-N,Nʹ-dioctylhexylethoxymalonamide (DMDOHEMA, and N,N,N’,N’- -tetraoctyldiglycolamide (TODGA. With TBP no extraction of Mo and Mg occurred. Both matrix materials are partly extracted by DMDOHEMA. Magnesium is not extracted by TODGA (D < 0.1, but a weak extraction of Mo is observed at low Mo concentration.

  15. The Impact of Microbially Influenced Corrosion on Spent Nuclear Fuel and Storage Life

    International Nuclear Information System (INIS)

    Wolfram, J. H.; Mizia, R. E.; Jex, R.; Nelson, L.; Garcia, K. M.

    1996-01-01

    A study was performed to evaluate if microbial activity could be considered a threat to spent nuclear fuel integrity. The existing data regarding the impact of microbial influenced corrosion (MIC) on spent nuclear fuel storage does not allow a clear assessment to be made. In order to identify what further data are needed, a literature survey on MIC was accomplished with emphasis on materials used in nuclear fuel fabrication, e.g., A1, 304 SS, and zirconium. In addition, a survey was done at Savannah River, Oak Ridge, Hanford, and the INEL on the condition of their wet storage facilities. The topics discussed were the SNF path forward, the types of fuel, ramifications of damaged fuel, involvement of microbial processes, dry storage scenarios, ability to identify microbial activity, definitions of water quality, and the use of biocides. Information was also obtained at international meetings in the area of biological mediated problems in spent fuel and high level wastes. Topics dis cussed included receiving foreign reactor research fuels into existing pools, synergism between different microbes and other forms of corrosion, and cross contamination

  16. Tri-fuel (diesel-biodiesel-ethanol) emulsion characterization, stability and the corrosion effect

    Science.gov (United States)

    Low, M. H.; Mukhtar, N. A. M.; Yohaness Hagos, Ftwi; Noor, M. M.

    2017-10-01

    This paper presents the result of experimenting emulsified tri-fuel in term of stability, physico-chemical properties and corrosion effect on three common metals. The results were interpreted in terms of the impact of five minutes emulsification approach. Tri-fuel emulsions were varied in proportion ratio consist of biodiesel; 0%, 5%, 10%, and ethanol; 5%, 10%, 15%. Fuel characterization includes density, calorific value, flash point, and kinematic viscosity. Flash point of tri-fuel emulsion came with range catalog. Calorific value of tri-fuel emulsion appeared in declining pattern as more ethanol and biodiesel were added. Biodiesel promoted flow resistance while ethanol with opposite effect. 15% ethanol content in tri-fuel emulsion separated faster than 10% ethanol content but ethanol content with 5% yield no phase separation at all. Close cap under static immersion with various ratio of tri-fuel emulsions for over a month, corrosiveness attack was detected via weight loss technique on aluminum, stainless steel and mild steel.

  17. The Impact of Microbially Influenced Corrosion on Spent Nuclear Fuel and Storage Life

    Energy Technology Data Exchange (ETDEWEB)

    J. H. Wolfram; R. E. Mizia; R. Jex; L. Nelson; K. M. Garcia

    1996-10-01

    A study was performed to evaluate if microbial activity could be considered a threat to spent nuclear fuel integrity. The existing data regarding the impact of microbial influenced corrosion (MIC) on spent nuclear fuel storage does not allow a clear assessment to be made. In order to identify what further data are needed, a literature survey on MIC was accomplished with emphasis on materials used in nuclear fuel fabrication, e.g., A1, 304 SS, and zirconium. In addition, a survey was done at Savannah River, Oak Ridge, Hanford, and the INEL on the condition of their wet storage facilities. The topics discussed were the SNF path forward, the types of fuel, ramifications of damaged fuel, involvement of microbial processes, dry storage scenarios, ability to identify microbial activity, definitions of water quality, and the use of biocides. Information was also obtained at international meetings in the area of biological mediated problems in spent fuel and high level wastes. Topics dis cussed included receiving foreign reactor research fuels into existing pools, synergism between different microbes and other forms of corrosion, and cross contamination.

  18. Computational simulation of the microstructure of irradiation damaged regions for the plate type fuel of UO2 microspheres dispersed in stainless steel matrix

    International Nuclear Information System (INIS)

    Reis, S.C. dos; Lage, A.F.; Braga, D.; Ferraz, W.B.

    2006-01-01

    Plate type fuel elements have high efficiency of thermal transference what benefits the heat flux with high rates of power output. In reactor cores, fuel elements, in general, are subject to a high neutrons flux, high working temperatures, severe corrosion conditions, direct interference of fission products that result from nuclear reactions and radiation interaction-matter. For plate type fuels composed of ceramic particles dispersed in metallic matrix, one can observe the damage regions that arise due to the interaction fission products in the metallic matrix. Aiming at evaluating the extension of the damage regions in function of the particles and its diameters, in this paper, computational geometric simulations structure of plate type fuel cores, composed of UO 2 microspheres dispersed in stainless steel in several fractions of volume and diameters were carried out. The results of the simulations were exported to AutoCAD R where it was possible its visualization and analysis. (author)

  19. Water Chemistry and Clad Corrosion/Deposition Including Fuel Failures. Proceedings of a Technical Meeting

    International Nuclear Information System (INIS)

    2013-03-01

    Corrosion is a principal life limiting degradation mechanism in nuclear steam supply systems, particularly taking into account the trends in increasing fuel burnup, thermal ratings and cycle length. Further, many plants have been operating with varying water chemistry regimes for many years, and issues of crud (deposition of corrosion products on other surfaces in the primary coolant circuit) are of significant concern for operators. At the meeting of the Technical Working Group on Fuel Performance and Technology (TWGFPT) in 2007, it was recommended that a technical meeting be held on the subject of water chemistry and clad corrosion and deposition, including the potential consequences for fuel failures. This proposal was supported by both the Technical Working Group on Advanced Technologies for Light Water Reactors (TWG-LWR) and the Technical Working Group on Advanced Technologies for Heavy Water Reactors (TWG-HWR), with a recommendation to hold the meeting at the National Nuclear Energy Generating Company ENERGOATOM, Ukraine. This technical meeting was part of the IAEA activities on water chemistry, which have included a series of coordinated research projects, the most recent of which, Optimisation of Water Chemistry to Ensure Reliable Water Reactor Fuel Performance at High Burnup and in Ageing Plant (FUWAC) (IAEATECDOC-1666), concluded in 2010. Previous technical meetings were held in Cadarache, France (1985), Portland, Oregon, USA (1989), Rez, Czech Republic (1993), and Hluboka nad Vltavou, Czech Republic (1998). This meeting focused on issues associated with the corrosion of fuel cladding and the deposition of corrosion products from the primary circuit onto the fuel assembly, which can cause overheating and cladding failure or lead to unplanned power shifts due to boron deposition in the clad deposits. Crud deposition on other surfaces increases radiation fields and operator dose and the meeting considered ways to minimize the generation of crud to avoid

  20. Microbially influenced corrosion communities associated with fuel-grade ethanol environments.

    Science.gov (United States)

    Williamson, Charles H D; Jain, Luke A; Mishra, Brajendra; Olson, David L; Spear, John R

    2015-08-01

    Microbially influenced corrosion (MIC) is a costly problem that impacts hydrocarbon production and processing equipment, water distribution systems, ships, railcars, and other types of metallic infrastructure. In particular, MIC is known to cause considerable damage to hydrocarbon fuel infrastructure including production, transportation, and storage systems, often times with catastrophic environmental contamination results. As the production and use of alternative fuels such as fuel-grade ethanol (FGE) increase, it is important to consider MIC of engineered materials exposed to these "newer fuels" as they enter existing infrastructure. Reports of suspected MIC in systems handling FGE and water prompted an investigation of the microbial diversity associated with these environments. Small subunit ribosomal RNA gene pyrosequencing surveys indicate that acetic-acid-producing bacteria (Acetobacter spp. and Gluconacetobacter spp.) are prevalent in environments exposed to FGE and water. Other microbes previously implicated in corrosion, such as sulfate-reducing bacteria and methanogens, were also identified. In addition, acetic-acid-producing microbes and sulfate-reducing microbes were cultivated from sampled environments containing FGE and water. Results indicate that complex microbial communities form in these FGE environments and could cause significant MIC-related damage that may be difficult to control. How to better manage these microbial communities will be a defining aspect of improving mitigation of global infrastructure corrosion.

  1. Thermal durability of modified Synroc material as reactor fuel matrix

    International Nuclear Information System (INIS)

    Kikuchi, Akira; Kanazawa, Hiroyuki; Togashi, Yoshihiro; Matumoto, Seiichiro; Nishino, Yasuharu; Ohwada, Isao; Nakata, Masahito; Amano, Hidetoshi; Mitamura, Hisayoshi

    1994-08-01

    A Synroc, a polyphase titanate ceramics composed of three mineral phases (perovskite, hollandite and zirconolite), has an excellent performance of immobilization of high level nuclear waste. A working group in the Department of Hot Laboratories paid special attention to this merit and started a development study on a LWR fuel named 'Waste Disposal Possible (WDP) Fuel', which has the two functions of a reactor fuel and a waste form. The present paper mainly describes thermal durability of a modified Synroc material, which is essentially important for applying the material to a fuel matrix. The two kinds of Synroc specimens, designated 'SM' as modified and 'SB' as a reference, were prepared by hot-pressing and annealed at 1200degC to 1500degC for 30 min in air. Unexpected and peculiar spherical voids were observed in the specimen SM at 1400degC and 1500degC, which caused the specimen swelling. The formation of the voids depends significantly on the existence of spherical precipitates seen in the as-fabricated specimen including latent micropores with high pressure. On the other hand, the heat treatment at 1500degC formed additional new phases, designated 'Phase A' for the specimen SB and 'Phase X' for SM. Phase A is a decomposition product of hollandite and Phase X a reaction product of Phase A and perovskite in the spherical voids. Furthermore, additional information and thermal properties examined are presented in Appendix 1 and Appendix 2, respectively. It was recognized that the modified Synroc specimen SM had excellent thermal properties. (author)

  2. Proceedings of a workshop on corrosion of Nuclear fuel waste containers

    International Nuclear Information System (INIS)

    Shoesmith, D.W.

    1990-01-01

    The 23 papers presented at this conference review the technical merits, and particularly corrosion performance, of the three main materials used for nuclear fuel waste containers: titanium and its alloys, copper and its alloys, and iron and carbon steels. The specific questions posed to the Workshop were: 1) Can we predict the lifetime of container materials in a variety of vault environments? 2) Is there a limiting range of conditions beyond which a specific material cannot be used? 3) Do we have the necessary corrosion rate data and/or mechanistic models required to make predictions? 4) Can we justify the use of titanium on the basis of propagation rate measurements for crevice corrosion, or do we need to prove initiation cannot occur? 5) Will the pitting of copper be significant? 6) How thick a carbon steel container would be required, and can it be fabricated and stress-relieved? 7) Are radiation fields of any consequence at the dose rates expected?

  3. Carbon Corrosion at Pt/C Interface in Proton Exchange Membrane Fuel Cell Environment

    International Nuclear Information System (INIS)

    Choi, Min Ho; Beam, Won Jin; Park, Chan Jin

    2010-01-01

    This study examined the carbon corrosion at Pt/C interface in proton exchange membrane fuel cell environment. The Pt nano particles were electrodeposited on carbon substrate, and then the corrosion behavior of the carbon electrode was examined. The carbon electrodes with Pt nano electrodeposits exhibited the higher oxidation rate and lower oxidation overpotential compared with that of the electrode without Pt. This phenomenon was more active at 75 .deg. C than 25 .deg. C. In addition, the current transients and the corresponding power spectral density (PSD) of the carbon electrodes with Pt nano electrodeposits were much higher than those of the electrode without Pt. The carbon corrosion at Pt/C interface was highly accelerated by Pt nano electrodeposits. Furthermore, the polarization and power density curves of PEMFC showed degradation in the performance due to a deterioration of cathode catalyst material and Pt dissolution

  4. Increased local corrosion of SVEA-96 fuel assemblies in KKL. Final report

    International Nuclear Information System (INIS)

    2001-11-01

    In February 1997 it was noted that the cladding surface below the spacers of SVEA-96 fuel assemblies showed increased local corrosion. This phenomenon called 'Enhanced Spacer Shadow Corrosion' (ESSC) by the fuel supplier was carefully monitored during the following 4 years. Several measures have been taken in order to counteract this ESSC. In this report of the Swiss Federal Agency for the Safety of Nuclear Installations (HSK) a summary is given of the technical and licensing aspects of ESSC. Although the fundamental mechanisms for the occurrence of ESSC are not yet sufficiently understood, short-term modification to water chemistry and the increasing use of improved cladding materials have effectively reduced this phenomenon. For the justification of the use of ESSC-damaged SVEA-96 fuel assemblies, HSK established temporary criteria which are based on technical investigations by the fuel assembly supplier. Among these, a special mention can be made of the more restrictive thermo-mechanical operation limit (TMOL) curve. As proof with respect of the HSK criteria, the plant operator conducted extended inspections on fuel assemblies during serving periods in 1997-2001 (measurement of oxide thickness). The conservative aspect of the measurements was assured through destructive examinations carried out at the Hot Laboratory of the Paul Scherrer Institute (PSI). Based on the modified water chemistry and the design of the core loading for cycle 18 (2001/2002) which contains only ESSC resistant cladding materials (LK2+, LK3), the original licence basis concerning the tolerable oxide thickness on the cladding could be guarantied. This has been verified by the results of a fuel assembly examination in August 2001. Therefore, the problem of the increased corrosion of the cladding of the SVEA-96 fuel assemblies is considered as being solved

  5. Factors affecting the silver corrosion performance of jet fuel from the Merox process

    Energy Technology Data Exchange (ETDEWEB)

    Viljoen, C.L. [Sasol Oil R& D (South Africa); Hietkamp, S. [CSIR, Pretoria (South Africa); Marais, B.; Venter, J.J. [National Petroleum Refineries of South Africa, Sasolburg (South Africa)

    1995-05-01

    The Natref refinery at Sasolburg, South Africa, which is 63,6% owned by Sasol and 36,5% by Total, is producing Jet A-1 fuel at a rate of 80 m{sup 3}/h by means of a UOP Merox process. A substantial part of the crude oil slate is made up from crudes which have been stored for considerable times in underground mines. Since the 1970`s, Natref has experienced sporadic non-conformance of its treated jet fuel to the silver corrosion (IP 227) test. Various causes and explanations for the sporadic silver corrosion occurrence have been put forward but a direct causal link has remained obscure. The paper addresses these possible causes for silver corrosion and some of the process changes which have been made to alleviate the problem. Emphasis is placed on the most recent approaches which were taken to identify the origin of the sporadic silver corrosion. An inventory of all the potential causes was made, such a bacterial action, elemental sulphur formation in storage, etc. and experiments designed to test the validity of these causes, are discussed. A statistical evaluation which was done of the historical process data over a 2 year period, failed to link the use of mine crudes directly to Ag-corrosion occurrence. However, a correlation between elemental sulphur and H{sub 2}S levels in the feed to the Merox reactor and Ag-corrosion was observed. Finally, the outcome of the experiments are discussed, as well as the conclusions which were reached from the observed results.

  6. Corrosion and pyrophoricity of ZPPR fuel plates: Implications for basin storage

    International Nuclear Information System (INIS)

    Totemeier, T.C.; Hayes, S.L.; Pahl, R.G.; Crawford, D.C.

    1997-01-01

    This paper presents the results of recent experimentation and analysis of the pyrophoric behavior of corroded Zero Power Physics Reactor (ZPPR) HEU fuel plates and the implications of these results for the handling, drying, and passivation of uranium metal fuels stored in water basins. The ZPPR plates were originally clad in 1980; crevice corrosion of the uranium metal in a dry storage environment has occurred due to the use of porous cladding end plugs. The extensive corrosion has resulted in bulging and, in some cases, breaching of the cladding over a 15 year storage period. Processing of the plates has been initiated to recover the highly enriched uranium metal and remove the storage vulnerability identified with the corroded plates, which have been shown to contain significant quantities of the pyrophoric compound uranium hydride (UH 3 ). Experiments were undertaken to determine effective passivation techniques for the corrosion product; analysis and modeling was performed to determine whether heat generated by rapid hydride re-oxidation could ignite the underlying metal plates. The results of the initial passivation experiment showed that simple exposure of the hydride-containing corrosion product to an Ar-3 vol.% O 2 environment was insufficient to fully passivate the hydride--flare-up of the product occurred during subsequent vigorous handling in air. A second experiment demonstrated that corrosion product was fully stable following grinding of the product to a fine powder in the Ar-3 vol.% O 2 atmosphere. Numerical modeling of a corroded plate indicated that ignition of the plate due to the heat from hydride re-oxidation was likely if hydride fractions in the corrosion product exceeded 30%

  7. Training implementation matrix. Spent Nuclear Fuel Project (SNFP)

    International Nuclear Information System (INIS)

    EATON, G.L.

    2000-01-01

    This Training Implementation Matrix (TIM) describes how the Spent Nuclear Fuel Project (SNFP) implements the requirements of DOE Order 5480.20A, Personnel Selection, Qualification, and Training Requirements for Reactor and Non-Reactor Nuclear Facilities. The TIM defines the application of the selection, qualification, and training requirements in DOE Order 5480.20A at the SNFP. The TIM also describes the organization, planning, and administration of the SNFP training and qualification program(s) for which DOE Order 5480.20A applies. Also included is suitable justification for exceptions taken to any requirements contained in DOE Order 5480.20A. The goal of the SNFP training and qualification program is to ensure employees are capable of performing their jobs safely and efficiently

  8. Corrosion of MTR type fuel plates containing U3O8-Al cermet cores

    International Nuclear Information System (INIS)

    Durazzo, M.

    1985-01-01

    The fuel plate samples containing U 3 O 8 -Al cermet cores with concentrations from 10 to 90% of U 3 O 8 weight were fabricated. Samples with 58% of U 3 O 8 eight were fabricated using compacts with densities from 75 to 95% of theoretical density. The influences of U 3 O 8 concentration and porosity of compacted core on porosity and uniformity of core thickness are discussed. The U 3 O 8 -Al cores were submitted to corrosion tests and exposed to deionized water at temperatures of 30, 50, 70 and 90 0 C by cladding deffect produced artificially. The results shown that core corrosion is accompanied by hydrogen release. The total volum of released hydrogen and the time interval to observe the initiation of hydrogen releasing (incubation time) are depending on core pososity and absolute temperature. A mechanism for U 3 O 8 -Al core corrosion process is proposed and discussed. The cladding of fuel plate samples was submitted to corrosion tests under similar conditons of the IAE-R1 reactor operating at 2, 5 and 10 MW. (Author) [pt

  9. Ash deposition and high temperature corrosion at combustion of aggressive fuels

    Energy Technology Data Exchange (ETDEWEB)

    Hede Larsen, O. [I/S Fynsvaerket, Faelleskemikerne, Odense (Denmark); Henriksen, N. [Elsamprojekt A/S, Faelleskemikerne, Fredericia (Denmark)

    1996-12-01

    In order to reduce CO{sub 2} emission, ELSAM is investigating the possibilities of using biomass - mainly straw - for combustion in high efficiency power plants. As straw has very high contents of chlorine and potassium, a fuel with high corrosion and ash deposition propensities has been introduced. ELSAM has investigated 3 ultra supercritical boiler concepts for combustion of straw alone or together with coal: (1) PF boilers with a relatively low share of straw, (2) CFB boilers with low to high share of straw and (3) vibrating grate boilers with 100% straw. These investigations has mainly been full-scale tests with straw fed into existing boilers. Corrosion tests have been performed in these boilers using temperature regulated probes and in-plant test tubes in existing superheaters. The corrosion has been determined by detailed measurements of wall thickness reduction and light optical microscopic measurements of the material degradation due to high temperature corrosion. Corrosion mechanisms have been evaluated using SEM/EDX together with thermodynamical considerations based on measurements of the chemical environment in the flue gas. Ash deposition is problematic in CFB boilers and in straw fired boilers, especially in years with high potassium and chlorine content of the straw. This ash deposition also is related to condensation of KCl and can probably only be handled by improved cleaning devices. (EG)

  10. A Semi-Empirical Two Step Carbon Corrosion Reaction Model in PEM Fuel Cells

    Energy Technology Data Exchange (ETDEWEB)

    Young, Alan; Colbow, Vesna; Harvey, David; Rogers, Erin; Wessel, Silvia

    2013-01-01

    The cathode CL of a polymer electrolyte membrane fuel cell (PEMFC) was exposed to high potentials, 1.0 to 1.4 V versus a reversible hydrogen electrode (RHE), that are typically encountered during start up/shut down operation. While both platinum dissolution and carbon corrosion occurred, the carbon corrosion effects were isolated and modeled. The presented model separates the carbon corrosion process into two reaction steps; (1) oxidation of the carbon surface to carbon-oxygen groups, and (2) further corrosion of the oxidized surface to carbon dioxide/monoxide. To oxidize and corrode the cathode catalyst carbon support, the CL was subjected to an accelerated stress test cycled the potential from 0.6 VRHE to an upper potential limit (UPL) ranging from 0.9 to 1.4 VRHE at varying dwell times. The reaction rate constants and specific capacitances of carbon and platinum were fitted by evaluating the double layer capacitance (Cdl) trends. Carbon surface oxidation increased the Cdl due to increased specific capacitance for carbon surfaces with carbon-oxygen groups, while the second corrosion reaction decreased the Cdl due to loss of the overall carbon surface area. The first oxidation step differed between carbon types, while both reaction rate constants were found to have a dependency on UPL, temperature, and gas relative humidity.

  11. Protection of porous carbon fuel particles from boudouard corrosion

    Science.gov (United States)

    Cooper, John F.

    2015-05-26

    A system for producing energy that includes infusing porous carbon particles produced by pyrolysis of carbon-containing materials with an off-eutectic salt composition thus producing pore-free carbon particles, and reacting the carbon particles with oxygen in a fuel cell according to the reaction C+O.sub.2=CO.sub.2 to produce electrical energy.

  12. Analytical assessment for stress corrosion fatigue of CANDU fuel elements under load following conditions

    International Nuclear Information System (INIS)

    Horhoianu, Grigore; Ionescu, Drags; Pauna, Eduard

    2012-01-01

    When nuclear power reactors are operated in a load following (LF) mode, the nuclear fuel may be subjected to step changes in power on weekly, daily, or even hourly basis, depending on the grid's needs. Two load following tests performed in TRIGA Research Reactor of Institute for Nuclear Research (INR) Pitesti were simulated with finite elements computer codes in order to evaluate Stress Corrosion Fatigue (SCF) of the sheath arising from expansion and contraction of the pellets in the corrosive environment. The 3D finite element analyses show that the cyclic strains give highly multiaxial stresses in the sheath at ridge region. This paper summarizes the results of the analytical assessment for SCF and their relation to CANDU fuel performance in LF tests conditions. (orig.)

  13. Comparison of matrix exponential methods for fuel burnup calculations

    International Nuclear Information System (INIS)

    Oh, Hyung Suk; Yang, Won Sik

    1999-01-01

    Series expansion methods to compute the exponential of a matrix have been compared by applying them to fuel depletion calculations. Specifically, Taylor, Pade, Chebyshev, and rational Chebyshev approximations have been investigated by approximating the exponentials of bum matrices by truncated series of each method with the scaling and squaring algorithm. The accuracy and efficiency of these methods have been tested by performing various numerical tests using one thermal reactor and two fast reactor depletion problems. The results indicate that all the four series methods are accurate enough to be used for fuel depletion calculations although the rational Chebyshev approximation is relatively less accurate. They also show that the rational approximations are more efficient than the polynomial approximations. Considering the computational accuracy and efficiency, the Pade approximation appears to be better than the other methods. Its accuracy is better than the rational Chebyshev approximation, while being comparable to the polynomial approximations. On the other hand, its efficiency is better than the polynomial approximations and is similar to the rational Chebyshev approximation. In particular, for fast reactor depletion calculations, it is faster than the polynomial approximations by a factor of ∼ 1.7. (author). 11 refs., 4 figs., 2 tabs

  14. Study of (U,Pu)O2 spent fuel matrix alteration under geological disposal conditions: Experimental approach and geochemical modeling

    International Nuclear Information System (INIS)

    Odorowski, Melina

    2015-01-01

    To assess the performance of direct disposal of spent fuel in a nuclear waste repository, researches are performed on the long-term behavior of spent fuel (UO x and MO x ) under environmental conditions close to those of the French disposal site. The objective of this study is to determine whether the geochemistry of the Callovian-Oxfordian (CO x ) clay geological formation and the steel overpack corrosion (producing iron and hydrogen) have an impact on the oxidative dissolution of the (U,Pu)O 2 matrix under alpha radiolysis of water. Leaching experiments have been performed with UO 2 pellets doped with alpha emitters (Pu) and MIMAS MO x fuel (un-irradiated or spent fuel) to study the effect of the CO x groundwater and of the presence of metallic iron upon the oxidative dissolution of these materials induced by the radiolysis of water. Results indicate an inhibiting effect of the CO x water on the oxidative dissolution. In the presence of iron, two different behaviors are observed. Under alpha irradiation as the one expected in the geological disposal, the alteration of UO 2 matrix and MO x fuel is very strongly inhibited because of the consumption of radiolytic oxidative species by iron in solution leading to the precipitation of Fe(III)-hydroxides on the pellets surface. On the contrary, under a strong beta/gamma irradiation field, alteration tracers indicate that the oxidative dissolution goes on and that uranium concentration in solution is controlled by the solubility of UO 2 (am,hyd). This is explained by the shifting of the redox front from the fuel surface to the bulk solution not protecting the fuel anymore. The developed geochemical (CHESS) and reactive transport (HYTEC) models correctly represent the main results and occurring mechanisms. (author) [fr

  15. Corrosion experiments on stainless steels used in dry storage canisters of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Ryskamp, J.M.; Adams, J.P.; Faw, E.M.; Anderson, P.A.

    1996-09-01

    Nonradioactive (cold) experiments have been set up in the Idaho Chemical Processing Plant (ICPP)-1634, and radioactive (hot) experiments have been set up in the Irradiated Fuel Storage Facility (IFSF) at ICPP. The objective of these experiments is to provide information on the interactions (corrosion) between the spent nuclear fuel currently stored at the ICPP and the dry storage canisters and containment materials in which this spent fuel will be stored for the next several decades. This information will be used to help select canister materials that will retain structural integrity over this period within economic, criticality, and other constraints. The two purposes for Dual Purpose Canisters (DPCs) are for interim storage of spent nuclear fuel and for shipment to a final geological repository. Information on how corrosion products, sediments, and degraded spent nuclear fuel may corrode DPCs will be required before the DPCs will be allowed to be shipped out of the State of Idaho. The information will also be required by the Nuclear Regulatory Commission (NRC) to support the licensing of DPCs. Stainless steels 304L and 316L are the most likely materials for dry interim storage canisters. Welded stainless steel coupons are used to represent the canisters in both hot and cold experiments.

  16. Corrosion experiments on stainless steels used in dry storage canisters of spent nuclear fuel

    International Nuclear Information System (INIS)

    Ryskamp, J.M.; Adams, J.P.; Faw, E.M.; Anderson, P.A.

    1996-09-01

    Nonradioactive (cold) experiments have been set up in the Idaho Chemical Processing Plant (ICPP)-1634, and radioactive (hot) experiments have been set up in the Irradiated Fuel Storage Facility (IFSF) at ICPP. The objective of these experiments is to provide information on the interactions (corrosion) between the spent nuclear fuel currently stored at the ICPP and the dry storage canisters and containment materials in which this spent fuel will be stored for the next several decades. This information will be used to help select canister materials that will retain structural integrity over this period within economic, criticality, and other constraints. The two purposes for Dual Purpose Canisters (DPCs) are for interim storage of spent nuclear fuel and for shipment to a final geological repository. Information on how corrosion products, sediments, and degraded spent nuclear fuel may corrode DPCs will be required before the DPCs will be allowed to be shipped out of the State of Idaho. The information will also be required by the Nuclear Regulatory Commission (NRC) to support the licensing of DPCs. Stainless steels 304L and 316L are the most likely materials for dry interim storage canisters. Welded stainless steel coupons are used to represent the canisters in both hot and cold experiments

  17. Influence of radiolysis on UO2 fuel matrix dissolution under disposal conditions. Literature Study

    International Nuclear Information System (INIS)

    Ollila, K.

    2011-05-01

    The objective of this study was to examine the recent published literature on the influence of water radiolysis on UO 2 fuel matrix dissolution under the disposal conditions. The α radiation is considered to be dominating over the other types of radiations at times longer than 1000 years. The presence of the anaerobic corrosion products of iron, especially of hydrogen, has been observed to play an important role under radiolysis conditions. It is not possible to exclude gamma/beta radiolysis effects in the experiments with spent fuel, since there is not available a fuel over 100 years old. More direct measurements of α radiolysis effects have been conducted with α doped UO 2 materials. On the basis of the results of these experiments, a specific activity threshold to observe α radiolysis effects has been presented. The threshold is 1.8 x 10 7 to 3.3 x 10 7 Bq/g in anoxic 10 -3 M carbonate solution. It is dependent on the environmental conditions, such as the reducing buffer capacity of the conditions. The results of dissolution rate measurements at VTT with 233 U-doped UO 2 samples in 0.01 to 0.1 M NaCl solutions under anoxic conditions did not show any effect of α radiolysis with doping levels of 5 and 10% 233 U (3.2 x 10 7 and 6.3 x 10 7 Bq/g). Both Fe 2+ and hydrogen can act as reducing species and could react with oxidizing radiolytic species. Fe 2+ concentrations of the order of 10 -5 M can decrease the rate of H 2 O 2 production. Low dissolution rates, 2 x 10 -8 to 2 x 10 -7 /yr, have been measured in the presence of metallic Fe with 5 and 10% 233 U-doped UO 2 in 0.01 to 1 M NaCl solutions. The tests with isotope dilution method showed precipitation phenomena of U to occur during dissolution process. The concentrations of dissolved U were extremely low (≤ 8.4 x 10 -11 M). No effects of -radiolysis could be seen. It is difficult to distinguish the effects of metallic Fe, Fe 2+ or hydrogen in these tests. Hydrogen could also act as a reducing agent

  18. Microbially influenced corrosion of copper nuclear fuel waste containers in a Canadian disposal vault

    International Nuclear Information System (INIS)

    King, F.

    1996-11-01

    An assessment of the potential for microbially influenced corrosion (MIC) of copper nuclear fuel waste containers in a Canadian disposal vault is presented. The assessment is based on a consideration of the microbial activity within a disposal vault, the reported cases of MIC of Cu alloys in the literature and the known corrosion behaviour of Cu. Because of the critical role of biofilms in the reported cases of MIC, their formation and properties are discussed in detail. Next, the literature on the MIC of Cu alloys is briefly reviewed. The various MIC mechanisms proposed are critically discussed and the implications for the corrosion of Cu containers considered. In the majority of literature cases, MIC depends on alternating aerated and deaerated environments, with accelerated corrosion being observed when fresh aerated water replaces stagnant water, e.g., the MIC of Cu-Ni heat exchangers in polluted seawater and the microbially influenced pitting of Cu water pipes. Finally, because of the predominance of corrosion by sulphate-reducing bacteria (SRB) in the MIC literature, the abiotic behaviour of Cu alloys in sulphide solutions is also reviewed. The effect of the evolving environment in a disposal vault on the extent and location of microbial activity is discussed. Biofilm formation on the container surface is considered unlikely throughout the container lifetime, but especially initially when the environmental conditions will be particularly aggressive. Microbial activity in areas of the vault away from the container is possible, however. Corrosion of the container could then occur if microbial metabolic by-products diffuse to the container surface. Sulphide, produced by the action of SRB are considered to be the most likely cause of container corrosion. It is concluded that the only likely form of MIC of Cu containers will result from sulphide produced by SRB diffusing to the container surface. A modelling procedure for predicting the extent of corrosion is

  19. Results of a recent crud/corrosion fuel risk assessment at a U.S. PWR

    International Nuclear Information System (INIS)

    Lamanna, Larry; Pop, Mike; Gregorich, Carola; Harne, Richard; Jones, John

    2012-09-01

    In order to avoid potential fuel reliability issues, specifically crud-related issues, it is necessary to achieve and maintain a crud safe environment. Therefore, the ability to confidently predict risks associated with crud deposition on fuel becomes critically important. AREVA is applying its cutting-edge PWR Fuel Crud (Primary System corrosion products)/Corrosion Tools, i.e. COBRA-FLX (subchannel-by-subchannel T/H tool) coupled with FDIC (crud deposition tool) to subsequently perform PWR Fuel Crud /Corrosion risk assessments for operating plants in the US. After describing the method, the result of one of these assessments is presented for an operating plant in the US that has experienced recent crud observations/concerns. Both Crud Induced Localized Corrosion (CILC) and Crud Induced Power Shift (CIPS) risk assessment methods, as applied to the upcoming cycle (Cycle N), were compared to the current/on-going cycle (Cycle N-1) and to the previous cycle (Cycle N-2). The results allowed the Utility to consider crud risk management changes associated with the upcoming cycle (Cycle-N). Benchmarking of the AREVA tools, using the plant-specific crud information gained from the crud sampling/characterization for the Unit will be presented. The CIPS analysis references boron loading and the amount of insoluble iron-nickel-borates predicted for Cycles N-2, N-1, and N. The results of the CILC evaluation reference FDIC-predicted crud thickness, cladding temperature under deposit, evolution of CILC bearing species and lithium concentration in the zirconium oxide layer. The approach taken by AREVA during the evaluation was to consider both 'risk' and 'margin' to fuel performance impact caused by crud deposits. The conclusion of the assessment, illustrated by the results presented in this paper, is that the example Plant has sufficient margin in worst case conditions for CIPS and CILC risk in Cycle N, based on Cycle N-1 and Cycle N-2 conditions and behavior

  20. Investigation on the corrosion resistance of PIM 316L stainless steel in PEM fuel cell simulated environment

    International Nuclear Information System (INIS)

    Oliveira, Mara Cristina Lopes de; Costa, Isolda; Antunes, Renato Altobelli

    2009-01-01

    Bipolar plates play main functions in PEM fuel cells, accounting for the most part of the weight and cost of these devices. Powder metallurgy may be an interesting manufacturing process of these components owing to the production of large scale, complex near-net shape parts. However, corrosion processes are a major concern due to the increase of the passive film thickness on the metal surface, lowering the power output of the fuel cell. In this work, the corrosion resistance of PIM AISI 316L stainless steel specimens was evaluated in 1M H 2 SO 4 + 2 ppm HF solution at room temperature during 30 days of immersion. The electrochemical measurements comprised potentiodynamic polarization and electrochemical impedance spectroscopy. The surface morphology of the specimens was observed before and after the corrosion tests through SEM images. The material presented low corrosion current density suggesting that it is suitable to operate in the PEM fuel cell environment. (author)

  1. Composition of the corrosion product oxide phases in the gaps of three LMFBR-type mixed oxide fuel pins

    International Nuclear Information System (INIS)

    Walker, C.T.

    1978-01-01

    Identification of the corrosion product oxide phases in the gaps of mixed oxide fuel pins clad with austenitic stainless steel is an important step in understanding the mechanisms of corrosion. Once the oxide is identified the reaction paths by which it is formed can be established. The corrosion product oxide will also be a reliable indication of the oxygen potential in the gap. A grey oxide phase which is recognised to contain the products of cladding corrosion is often observed in the gaps of mixed oxide fuel pins. Its main constituents are known to be chromium, caesium and oxygen, and recently the composition of the grey oxide phase in the gaps of three fuel pins was shown to be broadly similar, and it was proposed that the phase was essentially chromium oxide. (Auth.)

  2. Corrosion of Spent Nuclear Fuel: The Long-Term Assessment

    International Nuclear Information System (INIS)

    Ewing, Rodney C.

    2004-01-01

    Spent nuclear fuel, essentially U 2 , accounts for over 95% of the total radioactivity of all of the radioactive wastes in the United States that require disposal, disposition or remediation. The UO 2 in SNF is not stable under oxiding conditions and may also be altered under reducing conditions. The alteration of SNF results in the formation of new uranium phases that can cause the release or retardation of actinide and fission product radionuclides. Over the long term, and depending on the extent to which the secondary uranium phases incorporate fission products and actinides, these alteration phases become the near-field source term

  3. Corrosion of Spent Nuclear Fuel: The Long-Term Assessment

    Energy Technology Data Exchange (ETDEWEB)

    Rodney C. Ewing

    2004-10-07

    Spent nuclear fuel, essentially U{sub 2}, accounts for over 95% of the total radioactivity of all of the radioactive wastes in the United States that require disposal, disposition or remediation. The UO{sub 2} in SNF is not stable under oxiding conditions and may also be altered under reducing conditions. The alteration of SNF results in the formation of new uranium phases that can cause the release or retardation of actinide and fission product radionuclides. Over the long term, and depending on the extent to which the secondary uranium phases incorporate fission products and actinides, these alteration phases become the near-field source term.

  4. Corrosion resistance of ceramic materials in pyrochemical reprocessing atmosphere by using molten salt for spent nuclear oxide fuel. Corrosion research under chlorine gas condition

    International Nuclear Information System (INIS)

    Takeuchi, Masayuki; Hanada, Keiji; Koizumi, Tsutomu; Aose, Shinichi; Kato, Toshihiro

    2002-12-01

    Pyrochemical reprocessing using molten salts (RIAR process) has been recently developed for spent nuclear oxide fuel and discussed in feasibility study. It is required to improve the corrosion resistance of equipments such as electrolyzer because the process is operated in severe corrosion environment. In this study, the corrosion resistance of ceramic materials was discussed through the thermodynamic calculation and corrosion test. The corrosion test was basically carried out in alkali molten salt under chlorine gas condition. And further consideration about the effects of oxygen, carbon and main fission product's chlorides were evaluated in molten salt. The result of thermodynamic calculation shows most of ceramic oxides have good chemical stability on chlorine, oxygen and uranyl chloride, however the standard Gibb's free energies with carbon have negative value. On the other hand, eleven kinds of ceramic materials were examined by corrosion test, then silicon nitride, mullite and cordierite have a good corrosion resistance less than 0.1 mm/y. Cracks were not observed on the materials and flexural strength did not reduce remarkably after 480 hours test in molten salt with Cl 2 -O 2 bubbling. In conclusion, these three ceramic materials are most applicable materials for the pyrochemical reprocessing process with chlorine gas condition. (author)

  5. Corrosion cracking

    International Nuclear Information System (INIS)

    Goel, V.S.

    1985-01-01

    This book presents the papers given at a conference on alloy corrosion cracking. Topics considered at the conference included the effect of niobium addition on intergranular stress corrosion cracking, corrosion-fatigue cracking in fossil-fueled-boilers, fracture toughness, fracture modes, hydrogen-induced thresholds, electrochemical and hydrogen permeation studies, the effect of seawater on fatigue crack propagation of wells for offshore structures, the corrosion fatigue of carbon steels in seawater, and stress corrosion cracking and the mechanical strength of alloy 600

  6. Method of evaluation of stress corrosion cracking susceptibility of clad fuel tubes

    International Nuclear Information System (INIS)

    Takase, Iwao; Yoshida, Toshimi; Ikeda, Shinzo; Masaoka, Isao; Nakajima, Junjiro.

    1986-01-01

    Purpose: To determine, by an evaluation in out-pile test, the stress corrosion cracking susceptibility of clad fuel tubes in the reactor environment. Method: A plurality of electrodes are mounted in the circumferential direction on the entire surface of cladding tubes. Of the electrodes, electrodes at two adjacent places are used as measuring terminals and electrodes at another two places adjacent thereto are used as constant-current terminals. With a specific current flowing in the constant-current terminals, measurements are made of a potential difference between the terminals to be measured, and from a variation in the potential difference the depth of cracking of the cladding tube surface is presumed to determine the stress corrosion cracking susceptibility of the cladding tube. To check the entire surface of the cladding tube, the cladding tube is moved by each block in the circumferential direction by a contact changeover system, repeating the measurements of the potential difference. Contact type electrodes are secured with an insulator and held in uniform contact with the cladding tube by a spring. It is detachable by use of a locking system and movable as desired. Thus the stress corrosion cracking susceptibility can be determined without mounting the cladding tube through and also a fuel failure can be prevented. (Horiuchi, T.)

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

    Science.gov (United States)

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

    1994-07-01

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

  8. Design of containment system of nuclear fuel attacked by corrosion with leaking fission products

    International Nuclear Information System (INIS)

    Poblete Maturana, Tomas

    2015-01-01

    The following report presents the design of an innovative confinement system for the nuclear fuel attacked by corrosion, with leakage of fission products to be used in the RECH-1 nuclear experimental reactor of the Chilean Nuclear Energy Commission, is currently within the framework of the international nuclear waste management program developed by the member countries of the IAEA, including Chile. The main objective of this project is the development of a system that is capable of containing, in the smallest possible volume, the fission products that are released to the reactor coolant medium from the nuclear fuel that are attacked by corrosion. Among the tasks carried out for the development of the project are: the compilation of the necessary bibliography for the selection of the most suitable technology for the retention of the fission products, the calculation of the most important parameters to ensure that the system will operate within ranges that do not compromise the radiological safety, and the design of the hydraulic circuit of the system. The results obtained from the calculations showed that the fuel element confinement system is stable from a thermal point of view since the refrigerant does not under any circumstances reach the saturation temperature and, in addition, from a hydraulic point of view, since the rate at which the refrigerant flows through the hydraulic circuit is low enough so that the deformation of the fuel plates forming the nuclear fuel does not occur. The most appropriate technology for the extraction of fission products according to the literature consulted is by ion exchange. The calculations developed showed that with a very small volume of resins, it is possible to capture all of the non-volatile fission products of a nuclear fuel

  9. Investigation of thermally sensitised stainless steels as analogues for spent AGR fuel cladding to test a corrosion inhibitor for intergranular stress corrosion cracking

    Science.gov (United States)

    Whillock, Guy O. H.; Hands, Brian J.; Majchrowski, Tom P.; Hambley, David I.

    2018-01-01

    A small proportion of irradiated Advanced Gas-cooled Reactor (AGR) fuel cladding can be susceptible to intergranular stress corrosion cracking (IGSCC) when stored in pond water containing low chloride concentrations, but corrosion is known to be prevented by an inhibitor at the storage temperatures that have applied so far. It may be necessary in the future to increase the storage temperature by up to ∼20 °C and to demonstrate the impact of higher temperatures for safety case purposes. Accordingly, corrosion testing is needed to establish the effect of temperature increases on the efficacy of the inhibitor. This paper presents the results of studies carried out on thermally sensitised 304 and 20Cr-25Ni-Nb stainless steels, investigating their grain boundary compositions and their IGSCC behaviour over a range of test temperatures (30-60 °C) and chloride concentrations (0.3-10 mg/L). Monitoring of crack initiation and propagation is presented along with preliminary results as to the effect of the corrosion inhibitor. 304 stainless steel aged for 72 h at 600 °C provided a close match to the known pond storage corrosion behaviour of spent AGR fuel cladding.

  10. Approaches chosen to predict the effect of different forms of corrosion on copper in disposal conditions of spent fuel

    International Nuclear Information System (INIS)

    Bojinov, M.; Kinnunen, P.; Laitinen, T.; Maekelae, K.; Saario, T.; Hinttala, J.; Lilja, C.

    2003-01-01

    This paper presents deterministic approaches chosen to predict the effect of stress corrosion cracking, creep, pitting corrosion and general corrosion on the lifetime of copper in disposal conditions of spent fuel. The interest in copper in Finland and Sweden is because of the plans to use it as a corrosion shield for the cast iron canister containing high activity nuclear waste. Our approach is based on assessing whether the occurrence of different types of corrosion can be excluded in the disposal conditions. If this is not the case, we shall recommend the maximum corrosion penetration to be predicted on the basis of experimental results and to assess whether this is below an acceptable level. We have applied this approach to show that the occurrence of stress corrosion cracking in the presence of nitrite ions in the disposal conditions can very likely be excluded. In addition, we have shown that assessing the occurrence of localised corrosion due to bicarbonate ions and the occurrence of general corrosion in the disposal conditions requires further experimental work. (authors)

  11. Performance improvement in direct methanol fuel cells using a highly porous corrosion-resisting stainless steel flow field

    OpenAIRE

    Shudo, Toshio; Suzuki, Kensuke

    2008-01-01

    Power generation with direct methanol fuel cell (DMFC) systems require only simple equipment, and has the important advantages of using a liquid fuel with higher energy density and easier handling characteristics than hydrogen. However, the power output of DMFC is lower than hydrogen fuel cells. To improve the power output of DMFC it is very important to reduce diffusion polarization at higher current density conditions. This research used a corrosion-resisting type porous stainless steel dev...

  12. Temperature and humidity effects on the corrosion of aluminium-base reactor fuel cladding materials during dry storage

    International Nuclear Information System (INIS)

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

    2004-01-01

    The effect of temperature and relative humidity on the high temperature (up to 200 deg. C) corrosion of aluminum cladding alloys was investigated for dry storage of spent nuclear fuels. A dependency on alloy type and temperature was determined for saturated water vapor conditions. Models were developed to allow prediction of cladding behaviour of 1100, 5052, and 6061 aluminum alloys for up to 50+ years at 100% relative humidity. Calculations show that for a closed system, corrosion stops after all moisture and oxygen is used up during corrosion reactions with aluminum alloys. (author)

  13. Study of Influence of an Annealing on Corrosion Stability of Pipes-shells for Fuel of Zr1Nb Alloy

    International Nuclear Information System (INIS)

    Petel'guzov, I.A.; Rodak, A.G.; Pasenov, F.A.; Ishchenko, N.I.

    2006-01-01

    Explored influence an annealing to the kinetics of corrosion and mechanical characteristics of pipe material for shells fuel elements made from the experimental zirconium alloy Zr1Nb calcium-thermal way of production, in the comparison with the staff alloy E110 electrolytic way of production. Determined parameters of kinetics of corrosion depending on temperature and duration annealing before testing. Conducted also mechanical testing the alloys on the ring samples. Determined ranges of temperatures, within which corrosion characteristics save values, close to source, and connecting temperatures, under which is observed reduction research; investigating features

  14. Effect of Concentration and Temperature of Ethanol in Fuel Blends on Microbial and Stress Corrosion Cracking of High-Strength Steel

    Science.gov (United States)

    2012-02-01

    Localized environments in fuel grade ethanol (FGE) transportation systems, where conditions are suitable for growth, may allow for microbiologically influenced corrosion (MIC) of steel components. Interstate pipeline transportation of ethanol fuels i...

  15. Improving the corrosion resistance of proton exchange membrane fuel cell carbon supports by pentafluorophenyl surface functionalization

    Science.gov (United States)

    Forouzandeh, Farisa; Li, Xiaoan; Banham, Dustin W.; Feng, Fangxia; Joseph Kakanat, Abraham; Ye, Siyu; Birss, Viola

    2018-02-01

    In this study, the effect of surface functionalization on the electrochemical corrosion resistance of a high surface area, mesoporous colloid imprinted carbon powder (CIC), as well as microporous Vulcan carbon (VC, serving as the benchmark), was demonstrated, primarily for PEM fuel cell applications. CIC-22, which is highly hydrophilic and was synthesized with 22 nm silica colloid templates, and as-received, mildly hydrophobic, VC powders, were functionalized with 2,3,4,5,6-pentafluorophenyl (-PhF5) surface groups using a straightforward diazonium reduction reaction. These carbons were then subjected to corrosion testing, involving a potential cycling-step sequence in room temperature 0.5 M H2SO4. Using cyclic voltammetry and charge/time analysis, the double layer and pseudo-capacitive gravimetric charges of the carbons, prior to and after the application of these potential steps, were tracked in order to obtain information about surface area changes and the extent of carbon oxidation, respectively. It is shown that the corrosion resistance was improved by ca. 50-80% by surface functionalization, likely due to a combination of surface passivation (loss of carbon active sites) and increased surface hydrophobicity.

  16. Predicting the effects of microbial activity on the corrosion of copper nuclear fuel waste disposal containers

    International Nuclear Information System (INIS)

    King, F.; Stroes-Gascoyne, S.

    1996-08-01

    Microbially influenced corrosion (MIC) of copper nuclear fuel waste containers may occur in a disposal vault located 500-1000 m underground in the granitic rock of the Canadian Shield. The extent and diversity of microbial activity in the vault is expected to be limited initially because of the aggressive conditions produced by γ-radiation, elevated temperatures and desiccation of the clay-based buffer in which the containers will be embedded. Experimental results on the heat- and radiation-sensitivity of the natural microbiota in buffer material are presented. The data suggest that the low water activity in the buffer material will severely limit the growth of microbes near the container. The most likely form of MIC involves sulphate-reducing bacteria (SRB). Electrochemical experiments using a clay-covered copper electrode have shown that sulphide ions produced by SRB could diffuse through buffer material and induce corrosion of the container. A method to predict the long-term corrosion behaviour is presented. (author)

  17. Spent nuclear fuel. A review of properties of possible relevance to corrosion processes

    Energy Technology Data Exchange (ETDEWEB)

    Forsyth, R. [Caledon Consult AB, Nykoeping (Sweden)

    1995-04-01

    The report reviews the properties of spent fuel which are considered to be of most importance in determining the corrosion behaviour in groundwaters. Pellet cracking and fragment size distribution are discussed, together with the available results of specific surface area measurements on spent fuel. With respect to the importance of fuel microstructure, emphasis is placed on recent work on the so called structural rim effect, which consists of the formation of a zone of high porosity, and the polygonization of fuel grains to form many sub-grains, at the pellet rim, and appears to be initiated when the average pellet burnup exceeds a threshold of about 40 MWd/kgU. Due to neutron spectrum effects, the pellet rim is also associated with the buildup of plutonium and other actinides, which results in an enhanced local burnup and specific activity of both beta-gamma and alpha radiation, thus representing a greater potential for radiolysis effects in ingressed groundwater. The report presents and discusses the results of quantitative determination of the radial profiles of burnup and alpha activity on spent fuel with average burnups from 21.2 to 49 MWd/kgU. In addition to the radial variation of fission product and actinide inventories due to the effects mentioned above, migration, redistribution and release of some fission products can occur during reactor irradiation and the report concludes with a short review of these processes.

  18. Progress on matrix SiC processing and properties for fully ceramic microencapsulated fuel form

    International Nuclear Information System (INIS)

    Terrani, K.A.; Kiggans, J.O.; Silva, C.M.; Shih, C.; Katoh, Y.; Snead, L.L.

    2015-01-01

    The consolidation mechanism and resulting properties of the silicon carbide (SiC) matrix of fully ceramic microencapsulated (FCM) fuel form are discussed. The matrix is produced via the nano-infiltration transient eutectic-forming (NITE) process. Coefficient of thermal expansion, thermal conductivity, and strength characteristics of this SiC matrix have been characterized in the unirradiated state. An ad hoc methodology for estimation of thermal conductivity of the neutron-irradiated NITE–SiC matrix is also provided to aid fuel performance modeling efforts specific to this concept. Finally, specific processing methods developed for production of an optimal and reliable fuel form using this process are summarized. These various sections collectively report the progress made to date on production of optimal FCM fuel form to enable its application in light water and advanced reactors

  19. Stress corrosion cracking life estimation of hold-down spring screw for nuclear fuel assembly

    International Nuclear Information System (INIS)

    Koh, S.K.

    2005-01-01

    Hold-down spring screw fractures due to primary water stress corrosion cracking were observed in nuclear fuel assemblies. The screw fastens hold-down springs that are required to maintain the nuclear fuel assembly in contact with upper core plate and permit thermal and irradiation-induced length changes. In order to investigate the primary causes of the screw fractures, the finite element stress analysis and fracture mechanics analysis were performed on the hold-down spring assembly. The elastic-plastic finite element analysis showed that the local stresses at the critical regions of head-shank fillet and thread root significantly exceeded the yield strength of the screw material, resulting in local plastic deformation. Preloading on the screw applied for tightening had beneficial effects on the screw strength by reducing the stress level at the critical regions, compared to the screw without preload. Calculated deflections and strains at the hold-down springs using the finite element analysis were in very close agreements with the experimentally measured deflections and strains. Primary water stress corrosion cracking (PWSCC) life of the Inconel 600 screw was predicted by integrating the Scott's model and resulted in a life of 1.42 years, which was fairly close to the field experience. Cracks were expected to originate at the threaded region of the screw and propagated to the opposite side of the spring, which was confirmed by the fractographic analysis of the fractured screws. (orig.)

  20. Correlation of waterside corrosion and cladding microstructure in high-burnup fuel and gadolinia rods

    International Nuclear Information System (INIS)

    Chung, H.M.

    1989-09-01

    Waterside corrosion of the Zircaloy cladding has been examined in high-burnup fuel rods from several BWRs and PWRs, as well as in 3 wt % gadolinia burnable poison rods obtained from a BWR. The corrosion behavior of the high-burnup rods was then correlated with results from a microstructural characterization of the cladding by optical, scanning-electron, and transmission-electron microscopy (OM, SEM, and TEM). OM and SEM examination of the BWR fuel cladding showed both uniform and nodular oxide layers 2 to 45 μm in thickness after burnups of 11 to 30 MWd/kgU. For one of the BWRs, which was operated at 307 degree C rather than the normal 288 degree C, a relatively thick (50 to 70 μm) uniform oxide, rather than nodular oxides, was observed after a burnup of 27 to 30 MWd/kgU. TEM characterization revealed a number of microstructural features that occurred in association with the intermetallic precipitates in the cladding metal, apparently as a result of irradiation-induced or -enhanced processes. The BWR rods that exhibited white nodular oxides contained large precipitates (300 to 700 nm in size) that were partially amorphized during service, indicating that a distribution of the large intermetallic precipitates is conductive to nodular oxidation. 23 refs., 9 figs

  1. Corrosion behaviour of Zircaloy 4 fuel cans for high burnup in EdF PWRs

    International Nuclear Information System (INIS)

    Blat, M.; Kerrec, O.; Bourgoin, J.; Vrignaud, E.; Amanrich, H.

    1994-01-01

    Uniform corrosion of fuel cladding could be a limitation for burn-up enhancement. First, the oxide thickness measured on fuel cladding for high burn-up has been compared to the prediction of the EDF code, CYRANO 2E. A comparative metallurgical characterization has been also performed on samples which were oxidized in pile and in autoclave. Then, laboratories studies have been launched for a better understanding of the corrosion mechanisms. A reflection was proposed on the two main theoretical concepts proposed for these mechanisms. Their kinetics could be controlled by transfers in liquid medium (electrolyte) or in solid medium (compact oxide). For the first topic, a nanoscopic characterization of the oxide is in progress, using Atomic Force Microscope. The first results are presented. In the second case, an electrochemical approach (impedance spectroscopy and voltametry) is developed in our laboratories. The obtained results could give some new keys in order to understand the influence of some parameters (alloys composition, coolant chemistry,...). (authors). 7 figs., 1 tab., 7 refs

  2. In-reactor fuel cladding external corrosion measurement process and results

    International Nuclear Information System (INIS)

    Thomazet, J.; Musante, Y.; Pigelet, J.

    1999-01-01

    Analysis of the zirconium alloy cladding behaviour calls for an on-site corrosion measurement device. In the 80's, a FISCHER probe was used and allowed oxide layer measurements to be taken along the outer generating lines of the peripheral fuel rods. In order to allow measurements on inner rods, a thin Eddy current probe called SABRE was developed by FRAMATOME. The SABRE is a blade equipped with two E.C coils is moved through the assembly rows. A spring allows the measurement coil to be clamped on each of the generating lines of the scanned rods. By inserting this blade on all four assembly faces, measurements can also be performed along several generating lines of the same rod. Standard rings are fitted on the device and allow on-line calibration for each measured row. Signal acquisition and processing are performed by LAGOS, a dedicated software program developed by FRAMATOME. The measurements are generally taken at the cycle outage, in the spent fuel pool. On average, data acquisition calls for one shift per assembly (eight hours): this corresponds to more than 2500 measurement points. These measurements are processed statistically by the utility program SAN REMO. All the results are collected in a database for subsequent behaviour analysis: examples of investigated parameters are the thermal/hydraulic conditions of the reactors, the irradiation history, the cladding material, the water chemistry This analysis can be made easier by comparing the behaviour measurement and prediction by means of the COROS-2 corrosion code. (author)

  3. INITIAL ANALYSIS OF TRANSIENT POWER TIME LAG DUE TO HETEROGENEITY WITHIN THE TREAT FUEL MATRIX.

    Energy Technology Data Exchange (ETDEWEB)

    D.M. Wachs; A.X. Zabriskie, W.R. Marcum

    2014-06-01

    The topic Nuclear Safety encompasses a broad spectrum of focal areas within the nuclear industry; one specific aspect centers on the performance and integrity of nuclear fuel during a reactivity insertion accident (RIA). This specific accident has proven to be fundamentally difficult to theoretically characterize due to the numerous empirically driven characteristics that quantify the fuel and reactor performance. The Transient Reactor Test (TREAT) facility was designed and operated to better understand fuel behavior under extreme (i.e. accident) conditions; it was shutdown in 1994. Recently, efforts have been underway to commission the TREAT facility to continue testing of advanced accident tolerant fuels (i.e. recently developed fuel concepts). To aid in the restart effort, new simulation tools are being used to investigate the behavior of nuclear fuels during facility’s transient events. This study focuses specifically on the characterizing modeled effects of fuel particles within the fuel matrix of the TREAT. The objective of this study was to (1) identify the impact of modeled heterogeneity within the fuel matrix during a transient event, and (2) demonstrate acceptable modeling processes for the purpose of TREAT safety analyses, specific to fuel matrix and particle size. Hypothetically, a fuel that is dominantly heterogeneous will demonstrate a clearly different temporal heating response to that of a modeled homogeneous fuel. This time difference is a result of the uniqueness of the thermal diffusivity within the fuel particle and fuel matrix. Using MOOSE/BISON to simulate the temperature time-lag effect of fuel particle diameter during a transient event, a comparison of the average graphite moderator temperature surrounding a spherical particle of fuel was made for both types of fuel simulations. This comparison showed that at a given time and with a specific fuel particle diameter, the fuel particle (heterogeneous) simulation and the homogeneous simulation

  4. Study of the corrosion of AA 6061 in spent fuel materials

    International Nuclear Information System (INIS)

    Rodriguez, Sebastian; Haddad, Roberto; Lanzani, Liliana A.

    2003-01-01

    Localized attack induced by dust or other particles deposited on alloy AA 6061 surface under water has been addressed as a matter of concern after completion of an IAEA Coordinated Research Program (CRP) on the corrosion of aluminum clad spent fuel during storage in water basins. This attack has been observed in all kinds of waters, although it is more pronounced in those of higher conductivity. In these cases a strong attack (similar to pitting corrosion) up to several hundred microns in depth and about a millimeter in length has been found beneath the aluminum hydroxide blister formed in those places where specks had seated on. As this problem could seriously affect the fission product containing capacity of stored spent fuel even in well maintained high quality water, it is important to learn about the involved mechanism of attack and find out about the influence of particle composition, in order to establish the convenience or disapproval of the use of materials and procedures in storage basins. With this objective, an experimental approach has been developed to study the mechanism of corrosion attack linked with the deposition of particles of different composition on aluminum surfaces; this include two kinds of iron flakes, concrete powder, and sand particles. Immersion tests of up to 40 days have been conducted in nuclear grade demineralized water and sodium citrate solutions of several conductivities. The position of sediments was marked and followed through the process and the final state of the aluminum surfaces was assessed by optical and electron microscopy and other microanalysis techniques. Other complementary activities carried on in relation with this work are: through characterization of intermetallic particles in AA 6061, and the study of the electrochemical behavior of precipitates in high purity water. Mg 2 Si particles perform very actively, dissolving even in high pure water at open circuit potential, leaving a small hole on the surface. Iron

  5. Comparison of Material Behavior of Matrix Graphite for HTGR Fuel Elements upon Irradiation: A literature Survey

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Young-Woo; Yeo, Seunghwan; Cho, Moon Sung [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    The fuel elements for the HTGRs (i.e., spherical fuel element in pebble-bed type core design and fuel compact in prismatic core design) consists of coated fuel particles dispersed and bonded in a closely packed array within a carbonaceous matrix. This matrix is generally made by mixing fully graphitized natural and needle- or pitchcoke originated powders admixed with a binder material (pitch or phenolic resin), The resulting resinated graphite powder mixture, when compacted, may influence a number of material properties as well as its behavior under neutron irradiation during reactor operation. In the fabrication routes of these two different fuel element forms, different consolidation methods are employed; a quasi-isostatic pressing method is generally adopted to make pebbles while fuel compacts are fabricated by uni-axial pressing mode. The result showed that the hardness values obtained from the two directions showed an anisotropic behavior: The values obtained from the perpendicular section showed much higher micro hardness (176.6±10.5MPa in average) than from the parallel section ((125.6±MPa in average). This anisotropic behavior was concluded to be related to the microstructure of the matrix graphite. This may imply that the uni-axial pressing method to make compacts influence the microstructure of the matrix and hence the material properties of the matrix graphite.

  6. An investigation on the fuel/matrix reaction behaviors of U-Mo/Al dispersion fuels prepared with centrifugal atomization

    International Nuclear Information System (INIS)

    Kim, Ki-Hwan; Kwon, Hee-Jun; Lee, Jeong-Soo; Ryu, Ho-Jin; Park, Jong-Man; Kim, Chang-Kyu

    2000-01-01

    The out-pile examinations of atomized U-Mo dispersion fuels after reaction between U-Mo and the Al matrix at elevated temperature were carried out in order to investigate heat generation, reaction kinetics and thermal compatibility. The heat generation of the dispersion fuels during the thermal cycle from room temperature to 700 deg. C was measured employing differential scanning calorimetry (DSC). Growth rates of the reaction layer were fitted theoretically using spherical diffusion model. The effect of volume fraction and powdering method on the thermal compatibility of atomized UMo dispersion fuels was also examined. (author)

  7. Viability of inert matrix fuel in reducing plutonium amounts in reactors

    International Nuclear Information System (INIS)

    2006-08-01

    Reactors worldwide have produced more than 2000 tonnes of plutonium, contained in spent fuel or as separated forms through reprocessing. Disposition of fissile materials has become a primary concern of nuclear non-proliferation efforts. There is a significant interest in IAEA Member States to develop proliferation resistant nuclear fuel cycles for incineration of plutonium such as inert matrix fuels (IMFs). The present report summarises R and D work on inert matrix fuel for plutonium and (to a lesser extent) minor actinide stock-pile reduction, and discusses the possible strategies to include inert matrix fuel approaches to the nuclear fuel cycle. The publication reviews the status of potential IMF candidates and describes several identified candidate materials for both fast and thermal reactors: MgO, ZrO2, SiC, Zr alloy, SiAl, ZrN; some of these have undergone test irradiations and post-irradiation examination. Also discussed are modelling of IMF fuel performance and safety analysis. System studies have identified strategies for both implementation of IMF fuel as homogeneous or heterogeneous phases, as assemblies or core loadings and in existing reactors in the shorter term, as well as in new reactors in the longer term

  8. EPRI-NASA Cooperative Project on Stress Corrosion Cracking of Zircaloys. [nuclear fuel failures

    Science.gov (United States)

    Cubicciotti, D.; Jones, R. L.

    1978-01-01

    Examinations of the inside surface of irradiated fuel cladding from two reactors show the Zircaloy cladding is exposed to a number of aggressive substances, among them iodine, cadmium, and iron-contaminated cesium. Iodine-induced stress corrosion cracking (SCC) of well characterized samples of Zircaloy sheet and tubing was studied. Results indicate that a threshold stress must be exceeded for iodine SCC to occur. The existence of a threshold stress indicates that crack formation probably is the key step in iodine SCC. Investigation of the crack formation process showed that the cracks responsible for SCC failure nucleated at locations in the metal surface that contained higher than average concentrations of alloying elements and impurities. A four-stage model of iodine SCC is proposed based on the experimental results and the relevance of the observations to pellet cladding interaction failures is discussed.

  9. Corrosion of cermet cores of fuel plates for nuclear research reactor

    International Nuclear Information System (INIS)

    Durazzo, M.; Ramanathan, L.V.

    1984-01-01

    Materials Testing Reactor (MTR) type fuel plates containing U 3 O 8 -Al cores and clad with Al are used in various research reactor. Preliminary investigations, where in the cladding of samples was drilled to simulate conditions of rupture due to pitting attack, revealed that considerable quantities of H 2 was evolved upon exposure of the core to water. The corrosion of cermets cores of different densities was characterized as a function of H 2 evolution that revealed 3 stages. A first stage consisting of an incubation period followed by initiation of H 2 evolution, a second stage with a constant rate of H 2 evolution and a third stage with a low rate of H 2 evolution. All 3 stages were found to vary as a function of cermet density and water temperature. (Author) [pt

  10. Laboratory Investigations of the High Temperature Corrosion of Various Materials in Simulated oxy-fuel and Conventional Coal Firing

    International Nuclear Information System (INIS)

    Folkeson, N.; Pettersson, J.; Svensson, J.E.; Hjornhede, A.; Montgomery, M.; Bjurman, M.

    2009-01-01

    Laboratory exposures in horizontal tube furnaces were conducted to test various materials for corrosion resistance in simulated oxy-fuel firing and conventional coal firing environments. Two different exposures were done at 630 C for 672 hours. The reaction atmosphere, consisting of CO 2 , H 2 O, O 2 , N 2 and SO 2 , was mixed to resemble that of oxy-fuel firing in the first exposure and that of conventional coal firing in the second exposure (N 2 was added during the second exposure only). Four different materials were tested in the first exposure; Sanicro 63, Alloy 800HT, 304L and 304HCu. In the second exposure four different materials were tested; 304L, Alloy 800HT, Kanthal APMT and NiCrAl. Apart from cleaned sample coupons, some samples pre-exposed in a test rig under oxy-fuel conditions with lignite as fuel and some pre-exposed with bituminous coal as fuel were investigated in the first exposure. In the second exposure some samples were pre-exposed in a rig under conventional firing conditions with lignite as fuel. The corrosion attack on the investigated samples was analysed by gravimetry, x-ray diffraction (XRD) and scanning electron microscopy (SEM) with energy dispersive x-ray (EDX). The SEM/EDX analysis was made on both the sample envelope and metallographic cross sections of the samples. The results show that there is small difference in the corrosion attack between the two environments. There was also little difference in oxide morphology and composition between cleaned samples and pre-exposed samples of the same material. The austenitic chromia former 304HCu suffered the most extensive corrosion attack in the oxy-fuel environment. In the conventional air firing environment 304L showed the highest mass gain. Chromia formers with higher chromium concentrations performed better, especially the super austenitic Alloy 800HT, with its high chromium concentration, formed a thin and protective corundum type oxide. The nickel based Sanicro 63 showed very low

  11. Coated powder for electrolyte matrix for carbonate fuel cell

    International Nuclear Information System (INIS)

    Iacovangelo, C.D.; Browall, K.W.

    1985-01-01

    A plurality of electrolyte carbonate-coated ceramic particle which does not differ significantly in size from that of the ceramic particle and wherein no significant portion of the ceramic particle is exposed is fabricated into a porous tape comprised of said coated-ceramic particles bonded together by the coating for use in a molten carbonate fuel cell

  12. Multilayer graphene for long-term corrosion protection of stainless steel bipolar plates for polymer electrolyte membrane fuel cell

    DEFF Research Database (Denmark)

    Stoot, Adam Carsten; Camilli, Luca; Spiegelhauer, Susie Ann

    2015-01-01

    Abstract Motivated by similar investigations recently published (Pu et al., 2015), we report a comparative corrosion study of three sets of samples relevant as bipolar plates for polymer electrolyte fuel cells: stainless steel, stainless steel with a nickel seed layer (Ni/SS) and stainless steel...... with Ni seed layer coated by a multi-layered graphene thin film (G/Ni/SS). The graphene film, synthesized by chemical vapour deposition (CVD), has a moderate amount of defects according to Raman spectroscopy. Short/medium-term corrosion test shows no significant advantage of using G/Ni/SS rather than Ni...

  13. Calculus of radiolytic products generation in water due to alpha radiation. Determination of the spent nuclear fuels matrix alteration rate Determination of velocity of spent fuel matrix

    International Nuclear Information System (INIS)

    Quinones, J.; Serrano, J.; Diaz Arocas, P.; Rodriguez Almazan, J. L.; Bruno, J.; Cera, E.; Merino, J.; Esteban, J. A.; Martinez-Esparza, A.

    2000-01-01

    The generation of radiolytic products as a result of alpha radiation in the surface of the spent fuel is a key process in order to understand how the it becomes degraded in repository conditions. The present work has established a radiolytic model based on a set of reactions involving fuel oxidation-dissolution and radiolytic products recombination. It also includes the decrease of the dose rates as the main alpha emitters decay away. Four cases, with varying parameters of the system, have been assessed. The results show a decrease in both the concentration of the radiolytic products in the gap water and the degradation of the fuel matrix. It has been estimated that in the period of the evaluation (10''6 years) up to 52% of the pellet is altered in the conservative cases, whereas only 11% is altered in the realistic cases. No significant differences were observed when the carbonates reactions were included in the system. (Author)

  14. The Optimum Plutonium Inert Matrix Fuel Form for Reactor-Based Plutonium Disposition

    International Nuclear Information System (INIS)

    Tulenko, J.S.; Wang, J.; Acosta, C.

    2004-01-01

    The University of Florida has underway an ongoing research program to validate the economic, operational and performance benefits of developing an inert matrix fuel (IMF) for the disposition of the U.S. weapons plutonium (Pu) and for the recycle of reprocessed Pu. The current fuel form of choice for Pu disposition for the Department of Energy is as a mixed oxide (MOX) (PuO2/UO2). We will show analyses that demonstrate that a Silicon Carbide (SiC) IMF offers improved performance capabilities as a fuel form for Pu recycle and disposition. The reason that UF is reviewing various materials to serve as an inert matrix fuel is that an IMF fuel form can offer greatly reduced Pu and transuranic isotope (TRU) production and also improved thermal performance characteristics. Our studies showed that the Pu content is reduced by an order of magnitude while centerline fuel temperatures are reduced approximately 380 degrees centigrade compared to MOX. These reduced temperatures result in reduced stored heat and thermal stresses in the pellet. The reduced stored heat reduces the consequences of the loss of coolant accident, while the reduced temperatures and thermal stresses yield greatly improved fuel performance. Silicon Carbide is not new to the nuclear industry, being a basic fuel material in gas cooled reactors

  15. Advanced in-situ characterisation of corrosion properties of LWR fuel cladding materials

    International Nuclear Information System (INIS)

    Arilahti, E.; Bojinov, M.; Beverskog, B.

    1999-01-01

    The trend towards higher fuel burnups imposes a demand for better corrosion and hydriding resistance of cladding materials. Development of new and improved cladding materials is a long process. There is a lack of fast and reliable in-situ techniques to investigate zirconium alloys in simulated or in-core LWR coolant conditions. This paper describes a Thin Layer Electrode (TLE) arrangement suitable for in-situ characterization of oxide films formed on fuel cladding materials. This arrangement enables us to carry out: Versatile Thin Layer Electrochemical measurements, including: (i) Thin Layer Electrochemical impedance Spectroscopic (TLEIS) measurements to characterize the oxidation kinetics and mechanisms of metals and the properties of their oxide films in aqueous environments. These measurements can also be performed in low conductivity electrolytes. (ii) Thin-Layer Wall-Jet (TLWJ) measurements, which give the possibility to detect soluble reaction products and to evaluate the influence of novel water chemistry additions on their release. Solid Contact measurements: (i) Contact Electric Resistance (CER) measurements to investigate the electronic properties of surface films on the basis of d.c. resistance measurements. (i) Contact Electric impedance (CEI) measurements to study the electronic properties of surface films using a.c. perturbation. All the above listed measurements can be performed using one single measurement device developed at VTT. This device can be conveniently inserted into an autoclave. Its geometry is currently being optimized in cooperation with the OECD Halden Reactor Project. In addition, the applicability of the device for in-core measurements has been investigated in a joint feasibility study performed by VTT and JRC Petten. Results of some autoclave studies of the effect of LiOH concentration on the stability of fuel cladding oxide films are presented in this paper. (author)

  16. 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.

  17. Evaluation of steam corrosion and water quenching behavior of zirconium-silicide coated LWR fuel claddings

    Science.gov (United States)

    Yeom, Hwasung; Lockhart, Cody; Mariani, Robert; Xu, Peng; Corradini, Michael; Sridharan, Kumar

    2018-02-01

    This study investigates steam corrosion of bulk ZrSi2, pure Si, and zirconium-silicide coatings as well as water quenching behavior of ZrSi2 coatings to evaluate its feasibility as a potential accident-tolerant fuel cladding coating material in light water nuclear reactor. The ZrSi2 coating and Zr2Si-ZrSi2 coating were deposited on Zircaloy-4 flats, SiC flats, and cylindrical Zircaloy-4 rodlets using magnetron sputter deposition. Bulk ZrSi2 and pure Si samples showed weight loss after the corrosion test in pure steam at 400 °C and 10.3 MPa for 72 h. Silicon depletion on the ZrSi2 surface during the steam test was related to the surface recession observed in the silicon samples. ZrSi2 coating (∼3.9 μm) pre-oxidized in 700 °C air prevented substrate oxidation but thin porous ZrO2 formed on the coating. The only condition which achieved complete silicon immobilization in the oxide scale in aqueous environments was the formation of ZrSiO4 via ZrSi2 coating oxidation in 1400 °C air. In addition, ZrSi2 coatings were beneficial in enhancing quenching heat transfer - the minimum film boiling temperature increased by 6-8% in the three different environmental conditions tested. During repeated thermal cycles (water quenching from 700 °C to 85 °C for 20 s) performed as a part of quench tests, no spallation and cracking was observed and the coating prevented oxidation of the underlying Zircaloy-4 substrate.

  18. The Inhibitive Potential of Arecaceae Extract on the Corrosivity of Aluminium based Matrix Composite and Medium Carbon Steel in Different Media

    Directory of Open Access Journals (Sweden)

    Paul Aondona IHOM

    2013-06-01

    Full Text Available In this research study, the inhibitive potential of Arecaceae extract on the corrosivity of aluminium based matrix composite and medium carbon steel in different media have been undertaken. The composite and the carbon steel were exposed in 0.5M NaOH and 0.5M HCl solutions as the control solutions. The Arecaceae extract from the fruit juice was added in volume percentages of 5%v/v and 10%v/v. The exposure time covered 12 days, and the monitored temperature range during the exposure time was 25-39°C. The result indicated that the most corrosive media were those without the Arecaceae inhibitor; where corrosion rate ranged from 0.106mpy to 0.409mpy. The medium carbon steel did not show any sign of corrosion in the 0.5M NaOH solution. Arecaceae inhibitor addition of 5%v/v gave optimum corrosion rate reduction; the corrosion rate was in the range of 0.018mpy to 0.199mpy. Although the lowest range of 0.013mpy to 0.192mpy was obtained with 10%v/v of Arecaceae, however on the overall performance the 5%v/v of Arecaceae was better. The inhibition of corrosion by Arecaceae was as a result of adsorption and phase layers on the surface of the metals.

  19. Corrosion of high temperature resisting alloys exposed to heavy fuel ash; Corrosion de aleaciones resistentes a altas temperaturas expuestas a ceniza de combustoleo pesado

    Energy Technology Data Exchange (ETDEWEB)

    Wong Moreno, Adriana del Carmen

    1998-03-01

    The objective of the performed research was to study the degradation process by high temperature corrosion of alloys exposed to heavy fuel oil ashes through a comparative experimental evaluation of its performance that allowed to establish the mechanisms involved in the phenomenon. The experimentation carried out involved the determination of the resistance to the corrosion of 14 alloys of different type (low and medium alloy steels, ferritic and austenitic stainless steels, nickel base alloys and a FeCrAl alloy of type ODS) exposed to high temperatures (580 Celsius degrees - 900 Celsius degrees) in 15 ash deposits with different corrosive potential, which were collected in the high temperature zone of boilers of thermoelectric power stations. The later studies to the corrosion tests consisted of the analysis by sweeping electron microscopy supported by microanalysis of the corroded probes, with the purpose of determining the effect of Na, V and S on the corrosivity of the ash deposits and the effect of the main alloying elements on the corrosion resistance of the alloys. Such effects are widely documented to support the proposed mechanisms of degradation that are occurring. The global analysis of the generated results has allowed to propose a model to explain the global mechanism of corrosion of alloys exposed to the high temperatures of ash deposits. The proposed model, complements the processed one by Wilson, widely accepted for fused vanadates, as far as on one hand, it considers the effect of the sodium sulfate presence (in addition to the vanadium compounds) in the deposits, and on the other hand, it extends it to temperatures higher than the point of fusion of constituent vanadium compounds of the deposits. Both aspects involve considering the roll that the process of diffusion of species has on the degradation and the capacity of protection of the alloy. The research performed allowed to confirm what the Wilson model had established for deposits with high

  20. CORROSION RESISTANCE OF ORGANOMETALLIC COATING APLICATED IN FUEL TANKS USING ELECTROCHEMICAL IMPEDANCE SPECTROSCOPY IN BIOFUEL – PART I

    Directory of Open Access Journals (Sweden)

    Milene Adriane Luciano

    2014-10-01

    Full Text Available Nowadays, the industry has opted for more sustainable production processes, and the planet has also opted for new energy sources. From this perspective, automotive tanks with organometallic coatings as well as a partial substitution of fossil fuels by biofuels have been developed. These organometallic coated tanks have a zinc layer, deposited by a galvanizing process, formed between the steel and the organometallic coating. This work aims to characterize the organometallic coating used in metal automotive tanks and evaluate their corrosion resistance in contact with hydrated ethyl alcohol fuel (AEHC. For this purpose, the resistance of all layers formed between Zinc and EEP steel and also the tin coated steel, which has been used for over thirty years, were evaluated. The technique chosen was the Electrochemical Impedance Spectroscopy. The results indicated an increase on the corrosion resistance when organometallic coatings are used in AEHC medium. In addition to that, these coatings allow an estimated 25% reduction in tanks production costs.

  1. Fabrication experience of aluminium clad aluminium matrix dispersion fuels at BARC

    International Nuclear Information System (INIS)

    Ganguly, C.; Hedge, P.W.; Prasad, G.J.

    1995-01-01

    Aluminium clad, aluminium matrix plate type dispersion fuels have been fabricated in BARC in recent years as part of fuel development programme for small non-power research reactors. The present paper describes the flowsheet developed for fabrication of Al-UAl x , Al-U 3 Si 2 and Al-U 3 O 8 fuels at BARC. The Al-20% U alloy fuel for KAMINI neutron radiography reactor was prepared by 'melting and casting' route, followed by picture framing and roll-bonding. For higher 'U' density fuels namely, Al-UAl x , Al-U 3 O 8 and Al-U 3 Si 2 the 'powder metallurgy' route was followed for preparation of fuel meat. The novel features in fabrication route were: addition of Zr for stabilizing UAl 3 phase in Al-20% U alloy; x-ray radiography and microdensitometric scanning of radiographs for location of fuel outline inside fuel element and for confirming homogeneous distribution of fissile atoms; immersion ultrasonic testing for confirming good bonding between mating Al surface of the fuel plate. (author)

  2. Investigation of in-pile formed corrosion films on zircaloy fuel-rod claddings by impedance spectroscopy and galvanostatic anodization

    International Nuclear Information System (INIS)

    Gebhardt, O.

    1993-01-01

    Hot-cell investigations have been executed to study the corrosion behaviour of irradiated Zircaloy fuel-rod claddings by impedance spectroscopy and galvanostatic anodization. The thickness of the compact oxide at the metal/oxide interface and the thickness of the minimum barrier oxide have been determined at different positions along the claddings. As shown by analysis, both quantities first increase and then decrease with increasing thickness of the total oxide. (author) 6 figs., 33 refs

  3. Reactor fuel rod

    International Nuclear Information System (INIS)

    Inui, Mitsuhiro; Mori, Kazuma.

    1990-01-01

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

  4. Corrosion/95 conference papers

    International Nuclear Information System (INIS)

    Anon.

    1995-01-01

    The papers in this conference represent the latest technological advances in corrosion control and prevention. The following subject areas are covered: cathodic protection in natural waters; materials for fossil fuel combustion and conversion systems; modern problems in atmospheric corrosion; innovative ideas for controlling the decaying infrastructure; deposits and their effects on corrosion in industry; volatile high temperature and non aqueous corrosion inhibitors; corrosion of light-weight and precoated metals for automotive application; refining industry corrosion; corrosion in pulp and paper industry; arctic/cold weather corrosion; materials selection for waste incinerators and associated equipment; corrosion measurement technology; environmental cracking of materials; advancing technology in the coating industry; corrosion in gas treating; green inhibition; recent advances in corrosion control of rail equipment; velocity effects and erosion corrosion in oil and gas production; marine corrosion; corrosion of materials in nuclear systems; underground corrosion control; corrosion in potable and industrial water systems in buildings and its impact on environmental compliance; deposit related boiler tube failures; boiler systems monitoring and control; recent developments and experiences in reactive metals; microbiologically influenced corrosion; corrosion and corrosion control for steel reinforced concrete; international symposium on the use of 12 and 13 Cr stainless steels in oil and gas production environments; subsea corrosion /erosion monitoring in production facilities; fiberglass reinforced pipe and tubulars in oilfield service; corrosion control technology in power transmission and distribution; mechanisms and methods of scale and deposit control; closing the loop -- results oriented cooling system monitoring and control; and minimization of aqueous discharge

  5. Corrosion evaluation of metallic HLW/spent fuel disposal containers - review

    International Nuclear Information System (INIS)

    Kursten, B.; Smailos, E.; Azkarate, I.; Werme, L.; Smart, N.R.; Marx, G.; Cunado, M.A.; Santarini, G.

    2004-01-01

    Over the years a lot of investigations have been performed to choose suitable container materials and to characterize their long-term corrosion behaviour in contact with their potential disposal environments, i.e. salt, clay, and granite. Carbon steels, stainless steels, nickel-based alloys, titanium-based alloys, and copper have been widely investigated as potential container materials depending on the studied host rock formation. The results obtained in salt environments indicate that the passively corroding Ti99.8-Pd is the primary choice for the thin-walled corrosion-resistant concept, since its general corrosion rate is negligible and it is highly resistant to localized corrosion and stress corrosion cracking (SCC) in salt brines. The TStE 355 carbon steel is the first candidate for the corrosion-allowance concept because it is resistant to pitting corrosion and SCC and its general corrosion rates are sufficiently low to provide corrosion allowance acceptable for thick-walled containers. Stainless steels, Ni-based alloys, and Ti-based alloys are the most important candidate container materials in clay for the thin-walled concept, while carbon steel is considered the main choice for the thick-walled corrosion-allowance concept. Studies performed in granite seem to indicate that copper containers provide an excellent corrosion barrier with an estimated lifetime exceeding 100,000 years. The TStE 355 carbon steel is also a valid option for a thick-walled container concept in granite. In this paper, some relevant corrosion data of carbon steel and stainless steel in cementitious environments are given in addition because large amounts of concrete will be used as structural materials in most of the envisaged repository design concepts. This paper also provides recommendations for future studies. (authors)

  6. Corrosion Behaviour of Mg Alloys in Various Basic Media: Application of Waste Encapsulation of Fuel Decanning from UNGG Nuclear Reactor

    Science.gov (United States)

    Lambertin, David; Frizon, Fabien; Blachere, Adrien; Bart, Florence

    The dismantling of UNGG nuclear reactor generates a large volume of fuel decanning. These materials are based on Mg-Zr alloy. The dismantling strategy could be to encapsulate these wastes into an ordinary Portland cement (OPC) or geopolymer (aluminosilicate material) in a form suitable for storage. Studies have been performed on Mg or Mg-Al alloy in basic media but no data are available on Mg-Zr behaviour. The influence of representative pore solution of both OPC and geopolymer with Mg-Zr alloy has been studied on corrosion behaviour. Electrochemical methods have been used to determine the corrosion densities at room temperature. Results show that the corrosion densities of Mg-Zr alloy in OPC solution is one order of magnitude more important than in a geopolymer solution environment and the effect of an inhibiting agent has been undertaken with Mg-Zr alloy. Evaluation of corrosion hydrogen production during the encapsulation of Mg-Zr alloy in both OPC and geopolymer has also been done.

  7. Corrosion behaviour of austenitic stainless steel as a function of methanol concentration for direct methanol fuel cell bipolar plate

    Science.gov (United States)

    Wang, Lixia; Kang, Bin; Gao, Na; Du, Xiao; Jia, Linan; Sun, Juncai

    2014-05-01

    The corrosion behaviour of an AISI 304 stainless steel (304 SS) is investigated in aqueous acid methanol solutions (0.5 M H2SO4 + 2 ppm HF + x M CH3OH, x = 0, 1, 5, 10 and 20) at 50 °C to simulate the varied anodic operating conditions of direct methanol fuel cells. Electrochemical measurements including potentiodynamic polarisation, potentiostatic polarisation and electrochemical impedance spectroscopy tests, are employed to analyse the corrosion behaviour. The results reveal that the corrosion resistance of 304 SS is enhanced in solutions with higher methanol content. Scanning electron microscopy and inductively coupled plasma atomic emission spectrometry data indicate that the surface corrosion on 304 SS is alleviated when the methanol concentration is increased. According to the X-ray photoelectron spectroscopy and Mott-Schottky analyses, the passive films formed on the 304 SS after potentiostatic tests in all the test solutions are composed of a duplex electronic structure with an external n-type semiconductor layer and an internal p-type semiconductor layer. Further analyses of the surface conductivity conducted by measuring the interfacial contact resistance between the 304 SS and carbon paper reveal that the passive film formed in the solution with higher methanol content exhibits lower conductivity.

  8. Study of the aqueous corrosion mechanisms and kinetics of the AlFeNi aluminium based alloy used for the fuel cladding in the Jules Horowitz research reactor

    International Nuclear Information System (INIS)

    Wintergerst, M.

    2009-05-01

    For the Jules Horowitz new material-testing reactor (JHR), an aluminium base alloy, called AlFeNi, will be used for the cladding of the fuel plates. This alloy (Al - 1% Fe - 1% Ni - 1 % Mg), which is already used as fuel cladding, was developed for its good corrosion resistance in water at high temperatures. However, few studies dealing with the alteration process in water and the relationships with irradiation effects have been performed on this alloy. The conception of the JHR fuel requires a better knowledge of the corrosion mechanisms. Corrosion tests were performed in autoclaves at 70 C, 165 C and 250 C on AlFeNi plates representative of the fuel cladding. Several techniques were used to characterize the corrosion scale: SEM, TEM, EPMA, XRD, Raman spectroscopy. Our observations show that the corrosion scale is made of two main layers: a dense amorphous scale close to the metal and a porous crystalline scale in contact with the water. More than the morphology, the chemical compositions of both layers are different. This duplex structure results from a mixed growth mechanism: an anionic growth to develop the inner oxide and a cationic diffusion followed by a dissolution-precipitation process to form the outer one. Dynamic experiments at 70 C and corrosion kinetics measurements have demonstrated that the oxide growth process is controlled by a diffusion step associated to a dissolution/precipitation process. A corrosion mechanism of the AlFeNi alloy in aqueous media has been proposed. Then post-irradiation exams performed on irradiated fuel plates were used to investigate the effects of the irradiation on the corrosion behaviour in the reactor core. (author)

  9. Investigations on the corrosion resistance of metallic bipolar plates (BPP) in proton exchange membrane fuel cells (PEMFC) - understanding the effects of material, coating and manufacturing

    Science.gov (United States)

    Dur, Ender

    Polymer Electrolyte Membrane Fuel Cell (PEMFC) systems are promising technology for contributing to meet the deficiency of world`s clean and sustainable energy requirements in the near future. Metallic bipolar plate (BPP) as one of the most significant components of PEMFC device accounts for the largest part of the fuel cell`s stack. Corrosion for metallic bipolar plates is a critical issue, which influences the performance and durability of PEMFC. Corrosion causes adverse impacts on the PEMFC`s performance jeopardizing commercialization. This research is aimed at determining the corrosion resistance of metallic BPPs, particularly stainless steels, used in PEMFC from different aspects. Material selection, coating selection, manufacturing process development and cost considerations need to be addressed in terms of the corrosion behavior to justify the use of stainless steels as a BPP material in PEMFC and to make them commercially feasible in industrial applications. In this study, Ti, Ni, SS304, SS316L, and SS 430 blanks, and BPPs comprised of SS304 and SS316L were examined in terms of the corrosion behavior. SS316L plates were coated to investigate the effect of coatings on the corrosion resistance performance. Stamping and hydroforming as manufacturing processes, and three different coatings (TiN, CrN, ZrN) applied via the Physical Vapor Deposition (PVD) method in three different thicknesses were selected to observe the effects of manufacturing processes, coating types and coating thicknesses on the corrosion resistance of BPP, respectively. Uncoated-coated blank and formed BPP were subjected to two different corrosion tests: potentiostatic and potentiodynamic. Some of the substantial results: 1- Manufacturing processes have an adverse impact on the corrosion resistance. 2- Hydroformed plates have slightly higher corrosion resistance than stamped samples. 3- BPPs with higher channel size showed better corrosion resistance. 4- Since none of the uncoated samples

  10. Characterization and fuel cell performance analysis of polyvinylalcohol-mordenite mixed-matrix membranes for direct methanol fuel cell use

    Energy Technology Data Exchange (ETDEWEB)

    Uctug, Fehmi Goerkem, E-mail: gorkem.uctug@bahcesehir.edu.t [University of Manchester, School of Chemical Engineering and Analytical Science, M60 1QD (United Kingdom); Holmes, Stuart M. [University of Manchester, School of Chemical Engineering and Analytical Science, M60 1QD (United Kingdom)

    2011-10-01

    Highlights: > We investigated the availability of PVA-mordenite membranes for DMFC use. > We measured the methanol permeability of PVA-mordenite membranes via pervaporation. > We did the fuel cell testing of these membranes, which had not been done before. > We showed that PVA-mordenite membranes have poorer DMFC performance than Nafion. > Membrane performance can be improved by increasing the proton conductivity of PVA. - Abstract: Polyvinylalcohol-mordenite (PVA-MOR) mixed matrix membranes were synthesized for direct methanol fuel cell (DMFC) use. For the structural and the morphological characterization, Scanning Electron Microscopy and Thermal Gravimetric Analysis methods were used. Zeolite distribution within the polymer matrix was found to be homogeneous. An impedance spectroscope was used to measure the proton conductivity. In order to obtain information about methanol permeation characteristics, swelling tests and a series of pervaporation experiments were carried out. 60-40 wt% PVA-MOR membranes were found to give the optimum transport properties. Proton conductivity of these membranes was found to be slightly lower than that of Nafion117{sup TM} whereas their methanol permeability was at least two orders of magnitude lower than Nafion117{sup TM}. DMFC performance of the PVA-MOR membranes was also measured. The inferior DMFC performance of PVA-MOR membranes was linked to drying in the fuel cell medium and the consequent proton conductivity loss. Their performance was improved by adding a dilute solution of sulfuric acid into the feed methanol solution. Future studies on the improvement of the proton conductivity of PVA-MOR membranes, especially via sulfonation of the polymer matrix, can overcome the low-performance problem associated with insufficient proton conductivity.

  11. Corrosion resistance of tungsten carbide based cermet coatings deposited by High Velocity Oxy-Fuel spray process

    Energy Technology Data Exchange (ETDEWEB)

    Aw, Poh Koon [Singapore Institute of Manufacturing Technology (Singapore)], E-mail: pkaw@SIMTech.a-star.edu.sg; Tan, Annie Lai Kuan [Singapore Institute of Manufacturing Technology (Singapore); Tan, Tai Phong [Hamilton Sundstrand Pacific Aerospace Pte Ltd (Singapore); Qiu, Jianhai [School of Materials Science and Engineering, Nanyang Technological University (Singapore)

    2008-06-30

    WC-17Ni and WC-17Co coatings were deposited on mild steel and stainless steel substrates by High Velocity Oxy-Fuel (HVOF) spray process. WC-17Ni and WC-17Co coatings were obtained by the spray process and the porosity of these coatings was measured. Polarization and electrochemical impedance spectroscopy (EIS) were performed on both uncoated substrates and coated samples immersed in 3% NaCl solution. WC-17Ni coating with a lower porosity, serve as a better barrier and effectively prevented corrosion attack when it was deposited on mild steel substrate. The nickel binder in the WC-17Ni coating was found to have a better corrosion resistance than the cobalt binder in the WC-17Co coating.

  12. High temperature corrosion of metallic interconnects in solid oxide fuel cells

    Directory of Open Access Journals (Sweden)

    Bastidas, D. M.

    2006-12-01

    Full Text Available Research and development has made it possible to use metallic interconnects in solid oxide fuel cells (SOFC instead of ceramic materials. The use of metallic interconnects was formerly hindered by the high operating temperature, which made the interconnect degrade too much and too fast to be an efficient alternative. When the operating temperature was lowered, the use of metallic interconnects proved to be favourable since they are easier and cheaper to produce than ceramic interconnects. However, metallic interconnects continue to be degraded despite the lowered temperature, and their corrosion products contribute to electrical degradation in the fuel cell. Coatings of nickel, chromium, aluminium, zinc, manganese, yttrium or lanthanum between the interconnect and the electrodes reduce this degradation during operation

    El uso de interconectores metálicos en pilas de combustible de óxido sólido (SOFC en sustitución de materiales cerámicos ha sido posible gracias a la investigación y desarrollo de nuevos materiales metálicos. Inicialmente, el uso de interconectores metálicos fue limitado, debido a la elevada temperatura de trabajo, ocasionando de forma rápida la degradación del material, lo que impedía que fuesen una alternativa. A medida que la temperatura de trabajo de las SOFC descendió, el uso de interconectores metálicos demostró ser una buena alternativa, dado que son más fáciles de fabricar y más baratos que los interconectores cerámicos. Sin embargo, los interconectores metálicos continúan degradándose a pesar de descender la temperatura a la que operan las SOFC y, asimismo, los productos de corrosión favorecen las pérdidas eléctricas de la pila de combustible. Recubrimientos de níquel, cromo, aluminio, zinc, manganeso, itrio y lantano entre el interconector y los electrodos reduce dichas pérdidas eléctricas.

  13. Stability of SiC-matrix microencapsulated fuel constituents at relevant LWR conditions

    Science.gov (United States)

    Snead, L. L.; Terrani, K. A.; Katoh, Y.; Silva, C.; Leonard, K. J.; Perez-Bergquist, A. G.

    2014-05-01

    This paper addresses certain key feasibility issues facing the application of SiC-matrix microencapsulated fuels for light water reactor application. Issues addressed are the irradiation stability of the SiC-based nano-powder ceramic matrix under LWR-relevant irradiation conditions, the presence or extent of reaction of the SiC matrix with zirconium-based cladding, the stability of the inner and outer pyrolytic graphite layers of the TRISO coating system at this uncharacteristically low irradiation temperature, and the state of the particle-matrix interface following irradiation which could possibly affect thermal transport. In the process of determining these feasibility issues microstructural evolution and change in dimension and thermal conductivity was studied. As a general finding the SiC matrix was found to be quite stable with behavior similar to that of CVD SiC. In magnitude the irradiation-induced swelling of the matrix material was slightly higher and irradiation-degraded thermal conductivity was slightly lower as compared to CVD SiC. No significant reaction of this SiC-based nano-powder ceramic matrix material with Zircaloy was observed. Irradiation of the sample in the 320-360 °C range to a maximum dose of 7.7 × 1025 n/m2 (E > 0.1 MeV) did not have significant negative impact on the constituent layers of the TRISO coating system. At the highest dose studied, layer structure and interface integrity remained essentially unchanged with good apparent thermal transport through the microsphere to the surrounding matrix.

  14. Stability of SiC-matrix microencapsulated fuel constituents at relevant LWR conditions

    International Nuclear Information System (INIS)

    Snead, L.L.; Terrani, K.A.; Katoh, Y.; Silva, C.; Leonard, K.J.; Perez-Bergquist, A.G.

    2014-01-01

    This paper addresses certain key feasibility issues facing the application of SiC-matrix microencapsulated fuels for light water reactor application. Issues addressed are the irradiation stability of the SiC-based nano-powder ceramic matrix under LWR-relevant irradiation conditions, the presence or extent of reaction of the SiC matrix with zirconium-based cladding, the stability of the inner and outer pyrolytic graphite layers of the TRISO coating system at this uncharacteristically low irradiation temperature, and the state of the particle–matrix interface following irradiation which could possibly affect thermal transport. In the process of determining these feasibility issues microstructural evolution and change in dimension and thermal conductivity was studied. As a general finding the SiC matrix was found to be quite stable with behavior similar to that of CVD SiC. In magnitude the irradiation-induced swelling of the matrix material was slightly higher and irradiation-degraded thermal conductivity was slightly lower as compared to CVD SiC. No significant reaction of this SiC-based nano-powder ceramic matrix material with Zircaloy was observed. Irradiation of the sample in the 320–360 °C range to a maximum dose of 7.7 × 10 25 n/m 2 (E > 0.1 MeV) did not have significant negative impact on the constituent layers of the TRISO coating system. At the highest dose studied, layer structure and interface integrity remained essentially unchanged with good apparent thermal transport through the microsphere to the surrounding matrix

  15. Fuel Quality Impact in a Historical Perspective: A Review of 25 Years of EU-Funded Research on Fuel Characterization, Ash and Deposit Formation, and Corrosion

    DEFF Research Database (Denmark)

    Jappe Frandsen, Flemming; Fendt, Sebastian; Spliethoff, Hartmut

    2016-01-01

    Heating and cooling are responsible for approximately half of EU’s final energy demand, while biomass is currently responsible for more than 90% of all renewable heat. The goal is to increase share of biomass-based technologies, in the European heat market, from 11% in 2007 to about 25% in 2020...... and by intelligent plant design. Broaden the feedstocks for pulverized fuel (PF) and fluidized bed (FB) powerplants, using pre-treatment methods with focus on the reduction of harmful,inorganic elements: Cl, S and the alkali metals. Prevent power plant damage due to high-temperature Cl-induced corrosion. Reduce...

  16. Thermomechanical behavior of fuel particles in a matrix during reactor power excursions

    International Nuclear Information System (INIS)

    Brittan, R.O.; Smith, R.S.

    1977-01-01

    This work determines the largest particle size that can be used in fabricating fuel material without exceeding temperature or stress criteria during transient operation. To do this temperature distribution histories must be determined for various particle sizes and volume fractions using typical power densities histories of transient reactor operation. From these, the critical stresses are calculated. The model chosen to accomplish this is a spherical fuel particle in a spherical matrix shell. Heat flow and temperature continuity conditions are imposed at the interface, and a zero temperature gradient is specified at the outer radius of the matrix shell. The particle power density is assumed to be uniform radially. Provisions are made for uniform power density in the matrix to model gamma heating and power density in interface layers to allow for radiant and fission fragment heating. A computer code was prepared to solve the model performance, yielding the temperature and stress distribution histories. Material property variation with temperature is employed, along with a close mockup of the power density history during self-limiting reactor transients. To date, four fuel systems have been investigated: 1) UC.ZrC particles in graphite; 2) UO 2 particles in graphite; 3) UO 2 particles in chromium 4) UO 2 particles in stainless steel. The study indicates that the maximum allowable particle diameter varies as the square root of the initial transient period and of the particle volume fraction. The critical thermophysical parameter is the thermal diffusivity of the particle, since in all cases studied it is many times smaller than that of the matrix. That of the UC.ZrC solid solution particle is 5 or more times larger than that of the UO 2 particle. It was found that the particles of system 1) above could be about 4 times larger than that of the other sy

  17. Influence of hydrazine primary water chemistry on corrosion of fuel cladding and primary circuit components

    International Nuclear Information System (INIS)

    Iourmanov, V.; Pashevich, V.; Bogancs, J.; Tilky, P.; Schunk, J.; Pinter, T.

    1999-01-01

    Earlier at Paks 1-4 NPP standard ammonia chemistry was in use. The following station performance indicators were improved when hydrazine primary water chemistry was introduced: occupational radiation exposures of personnel; gamma-radiation dose rates near primary system components during refuelling and maintenance outages. The reduction of radiation exposures and radiation fields were achieved without significant expenses. Recent results of experimental studies allowed to explain the mechanism of hydrazine dosing influence on: corrosion rate of structure materials in primary coolant; behaviour of soluble and insoluble corrosion products including long-life corrosion-induced radionuclides in primary system during steady-state and transient operation modes; radiolytic generation of oxidising radiolytic products in core and its corrosion activity in primary system; radiation situation during refuelling and maintenance outages; foreign material degradation and removal (including corrosion active oxidant species) from primary system during abnormal events. Operational experience and experimental data have shown that hydrazine primary water chemistry allows to reduce corrosion wear and thereby makes it possible to extend the life-time of plant components in primary system. (author)

  18. Zirconium carbonitride pellets by internal sol gel and spark plasma sintering as inert matrix fuel material

    Science.gov (United States)

    Hedberg, Marcus; Cologna, Marco; Cambriani, Andrea; Somers, Joseph; Ekberg, Christian

    2016-10-01

    Inert matrix fuel is a fuel type where the fissile material is blended with a solid diluent material. In this work zirconium carbonitride microspheres have been produced by internal sol gel technique, followed by carbothermal reduction. Material nitride purities in the produced materials ranged from Zr(N0.45C0.55) to Zr(N0.74C0.26) as determined by X-ray diffraction and application of Vegard's law. The zirconium carbonitride microspheres have been pelletized by spark plasma sintering (SPS) and by conventional cold pressing and sintering. In all SPS experiments cohesive pellets were formed. Maximum final density reached by SPS at 1700 °C was 87% theoretical density (TD) compared to 53% TD in conventional sintering at 1700 °C. Pore sizes in all the produced pellets were in the μm scale and no density gradients could be observed by computer tomography.

  19. In situ ceramic layer growth on coated fuel particles dispersed in a zirconium metal matrix

    Science.gov (United States)

    Terrani, K. A.; Silva, C. M.; Kiggans, J. O.; Cai, Z.; Shin, D.; Snead, L. L.

    2013-06-01

    The extent and nature of the chemical interaction between the outermost coating layer of coated fuel particles embedded in zirconium metal during fabrication of metal matrix microencapsulated fuels were examined. Various particles with outermost coating layers of pyrocarbon, SiC, and ZrC have been investigated in this study. ZrC-Zr interaction was the least substantial, while the PyC-Zr reaction can be exploited to produce a ZrC layer at the interface in an in situ manner. The thickness of the ZrC layer in the latter case can be controlled by adjusting the time and temperature during processing. The kinetics of ZrC layer growth is significantly faster from what is predicted using literature carbon diffusivity data in ZrC. SiC-Zr interaction is more complex and results in formation of various chemical phases in a layered aggregate morphology at the interface.

  20. Localised corrosion processes of austenitic stainless steel bipolar plates for polymer electrolyte membrane fuel cells

    Science.gov (United States)

    Mele, Claudio; Bozzini, Benedetto

    This research addresses the problem of localised corrosion of stainless steel PEMFC bipolar plates. The susceptibility to pitting and crevice corrosion of austenitic AISI 304 stainless steel has been investigated both by post-mortem microscopic analysis of the end-plates of a laboratory single-cell and by studies of electrochemically corroded stainless steels, in the presence of specially-designed crevice-formers simulating the operating conditions of a PEMFC. This work is based on optical and scanning-electron microscopies as well as potentiostatic and potentiodynamic measurements. The crevice-formers we considered were: Teflon, graphite and AISI 304. The samples, coupled to the crevice-formers have been tested in aqueous solutions containing Cl -, SO 4 2- and F -. From the E-log i plot, the values of corrosion, pitting, crevice and protection potential have been obtained and perfect and imperfect passivity conditions have been identified.

  1. Influence of anti-corrosion additive on the performance, emission and engine component wear characteristics of an IDI diesel engine fueled with palm biodiesel

    International Nuclear Information System (INIS)

    Ashraful, A.M.; Masjuki, H.H.; Kalam, M.A.; Rashedul, H.K.; Sajjad, H.; Abedin, M.J.

    2014-01-01

    Highlights: • Maximum engine performance was obtained at 2000 rpm for all fuel blends. • IRGALUBE 349 additive is enhances diesel engine performance. • Reduction of CO and NOx considerably using anti-corrosion additive except HC. • Engine wear decreases with using blended fuels with anti-corrosion additive. - Abstract: This study evaluates the effect of anti-corrosion additives such as 8% and 16% (vol.%) palm olein oil (PO) with ordinary diesel (OD) fuel on engine operation, emission behavior, engine part wear, and lubrication characteristics. This experiment was conducted on 4-cylinder and 4-stroke IDI diesel engine at different engine speed ranging from 1200 to 2800 RPM with 30% throttle setting under full load condition. The properties of the palm olein oil blends meet the ASTM D6751 and EN 14214 standards. At 2000 rpm, the experimental results revealed that the POD8A (0.2% Additive + 8% PO + 92% OD) and POD16A (0.2% Additive + 16% PO + 84% OD) blended fuels produced 0.5% and 0.51% higher brake power as well as 1.45% and 1.25% higher torque than same blends without additive, respectively. In comparison with ODF, the brake specific fuel consumption (BSFC) was found 1.8% and 3.1% higher for POD8A and POD16A blends, respectively. Anti-corrosion additive is found more effectual in enhancing the engine performance as such additive helps in timely ignition for complete burn in the combustion chamber. The results from engine emission indicated that POD8A and POD16A blended fuel reduced CO emissions by 11% and 6.6% and NOx emission by 2.5% and 1.09%, respectively in compared with OD fuel. Although HC emissions for all blended fuel and OD fuel increased at higher engine speed, the average HC emissions of all blended fuel were not higher than OD fuel. The application of anti-corrosion additives in POD blends reduced ferrous (Fe) wear debris concentration (WBC) by 17.3%. The reductions in WBC were about 16.1%, 10.8%, and 19.3%, 17.6% for copper (Cu) and aluminum

  2. Development of models and online diagnostic monitors of the high-temperature corrosion of refractories in oxy/fuel glass furnaces : final project report.

    Energy Technology Data Exchange (ETDEWEB)

    Griffiths, Stewart K.; Gupta, Amul (Monofrax Inc., Falconer, NY); Walsh, Peter M.; Rice, Steven F.; Velez, Mariano (University of Missouri, Rolla, MO); Allendorf, Mark D.; Pecoraro, George A. (PPG Industries, Inc., Pittsburgh, PA); Nilson, Robert H.; Wolfe, H. Edward (ANH Refractories, Pittsburgh, PA); Yang, Nancy Y. C.; Bugeat, Benjamin () American Air Liquide, Countryside, IL); Spear, Karl E. (Pennsylvania State University, University Park, PA); Marin, Ovidiu () American Air Liquide, Countryside, IL); Ghani, M. Usman (American Air Liquide, Countryside, IL)

    2005-02-01

    This report summarizes the results of a five-year effort to understand the mechanisms and develop models that predict the corrosion of refractories in oxygen-fuel glass-melting furnaces. Thermodynamic data for the Si-O-(Na or K) and Al-O-(Na or K) systems are reported, allowing equilibrium calculations to be performed to evaluate corrosion of silica- and alumina-based refractories under typical furnace operating conditions. A detailed analysis of processes contributing to corrosion is also presented. Using this analysis, a model of the corrosion process was developed and used to predict corrosion rates in an actual industrial glass furnace. The rate-limiting process is most likely the transport of NaOH(gas) through the mass-transport boundary layer from the furnace atmosphere to the crown surface. Corrosion rates predicted on this basis are in better agreement with observation than those produced by any other mechanism, although the absolute values are highly sensitive to the crown temperature and the NaOH(gas) concentration at equilibrium and at the edge of the boundary layer. Finally, the project explored the development of excimer laser induced fragmentation (ELIF) fluorescence spectroscopy for the detection of gas-phase alkali hydroxides (e.g., NaOH) that are predicted to be the key species causing accelerated corrosion in these furnaces. The development of ELIF and the construction of field-portable instrumentation for glass furnace applications are reported and the method is shown to be effective in industrial settings.

  3. ''C-ring'' stress corrosion cracking scoping experiment for Zircaloy spent fuel cladding

    International Nuclear Information System (INIS)

    Smith, H.D.

    1986-03-01

    This document describes the purpose and execution of the stress corrosion cracking scoping experiment using ''C-ring'' cladding specimens. The design and operation of the ''C-ring'' stressing apparatus is described and discussed. The experimental procedures and post-experiment sample evaluation are described

  4. Field test corrosion experiments in Denmark with biomass fuels Part I Straw firing

    DEFF Research Database (Denmark)

    Montgomery, Melanie; Karlsson, A; Larsen, OH

    2002-01-01

    to investigate corrosion: a) the exposure of metal rings on water/air cooled probes, b) the exposure of test tubes in a test superheater, and c) the exposure of test tubes in existing superheaters. Thus both austenitic steels and ferritic steels were exposed in the steam temperature range of 450-600°C...

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

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  6. Characterization of the Corrosion of a P-130X Graphite Fiber Reinforced 6063 Aluminum Metal Matrix Composite

    Science.gov (United States)

    1989-12-01

    alloy’s surface [Ref. 51. 6 ALUMINIUM AND ALUMINIUM ALLOYS -2 -f 0 2 3 4 5 6 7 8 9 10 It 12 13 14 IS 16 048iAA|0+3--0_ - 0.41 - - SI 6 -T A GLI 2HT AO 02...1982. 3. Bonewitz, R. A., "An Electrochemical Evaluation of 1100, 5052 , and 6063 Aluminum Alloys for Desalination," Corrosion, Vol. 29, pp. 215-222, June...1 A. CORROSION OF ALUMINUM ALLOYS

  7. Nuclear reactor fuel structure containing uranium alloy wires embedded in a metallic matrix plate

    Science.gov (United States)

    Travelli, A.

    1985-10-25

    A flat or curved plate structure, to be used as fuel in a nuclear reactor, comprises elongated fissionable wires or strips embedded in a metallic continuous non-fissionable matrix plate. The wires or strips are made predominantly of a malleable uranium alloy, such as uranium silicide, uranium gallide or uranium germanide. The matrix plate is made predominantly of aluminum or an aluminum alloy. The wires or strips are located in a single row at the midsurface of the plate, parallel with one another and with the length dimension of the plate. The wires or strips are separated from each other, and from the surface of the plate, by sufficient thicknesses of matrix material, to provide structural integrity and effective fission product retention, under neutron irradiation. This construction makes it safely feasible to provide a high uranium density, so that the uranium enrichment with uranium 235 may be reduced below about 20%, to deter the reprocessing of the uranium for use in nuclear weapons.

  8. SCANNING ELECTRON MICROSCOPY ANALYSIS OF FUEL/MATRIX INTERACTION LAYERS IN HIGHLY-IRRADIATED U-Mo DISPERSION FUEL PLATES WITH Al AND Al–Si ALLOY MATRICES

    Directory of Open Access Journals (Sweden)

    DENNIS D. KEISER, JR.

    2014-04-01

    Full Text Available In order to investigate how the microstructure of fuel/matrix-interaction (FMI layers change during irradiation, different U–7Mo dispersion fuel plates have been irradiated to high fission density and then characterized using scanning electron microscopy (SEM. Specifially, samples from irradiated U–7Mo dispersion fuel elements with pure Al, Al-2Si and AA4043 (∼4.5 wt.%Si matrices were SEM characterized using polished samples and samples that were prepared with a focused ion beam (FIB. Features not observable for the polished samples could be captured in SEM images taken of the FIB samples. For the Al matrix sample, a relatively large FMI layer develops, with enrichment of Xe at the FMI layer/Al matrix interface and evidence of debonding. Overall, a significant penetration of Si from the FMI layer into the U–7Mo fuel was observed for samples with Si in the Al matrix, which resulted in a change of the size (larger and shape (round of the fission gas bubbles. Additionally, solid fission product phases were observed to nucleate and grow within these bubbles. These changes in the localized regions of the microstructure of the U–7Mo may contribute to changes observed in the macroscopic swelling of fuel plates with Al-Si matrices.

  9. The potential for stress corrosion cracking of copper containers in a Canadian nuclear fuel waste disposal vault

    International Nuclear Information System (INIS)

    King, F.

    1996-09-01

    The potential for stress corrosion cracking (SCC) of copper nuclear fuel waste containers in a conceptual Canadian disposal vault has been assessed through a review of the literature and comparison of those environmental factors that cause SCC with the expected disposal environment. Stress-corrosion cracking appears to be an unlikely failure mode for Cu containers in a Canadian disposal vault because of a combination of environmental factors. Most importantly, there is only a relatively short period during which the containers will be undergoing strain when cracking should be possible at all, and then cracking is not expected because of the absence of known SCC agents, such as NH 3 , NO 2 - or organic acids. In addition, other environmental factors will mitigate SCC, namely, the presence of C1 - and its effect on film properties and the limited supply of oxidants. These arguments, to greater or lesser extent, apply to the three major mechanisms proposed for SCC of Cu alloys in aqueous solutions: film-rupture/anodic dissolution, tarnish rupture and film-induced cleavage. Detailed reviews of the SCC literature are presented as Appendices. The literature on the SCC of Cu (>99 wt.% Cu) is reviewed, including studies carried out in a number of countries under nuclear waste disposal conditions. Because of similarities with the behaviour of Cu, the more extensive literature on the SCC of α-brass in ammonia solutions is also reviewed. (author). 140 refs., 3 tabs., 25 figs

  10. Influence of the silicon content on the core corrosion properties of dispersion type fuel plates

    International Nuclear Information System (INIS)

    Calvo, C.; Saenz de Tejada, L. M.; Diaz Diaz, J.

    1969-01-01

    A new process to produce aluminium base dispersion type fuel plates has been developed at the Spanish JEN (Junta de Energia Nuclear). The dispersed fuel material is obtained by an aluminothermic process to render a stoichiometric cermet of UAI 3 and AI 2 O 3 according to the reaction. (Author)

  11. Hydrogen Sulphide Corrosion of Carbon and Stainless Steel Alloys Immersed in Mixtures of Renewable Fuel Sources and Tested Under Co-processing Conditions

    Directory of Open Access Journals (Sweden)

    Gergely András

    2016-10-01

    Full Text Available In accordance with modern regulations and directives, the use of renewable biomass materials as precursors for the production of fuels for transportation purposes is to be strictly followed. Even though, there are problems related to processing, storage and handling in wide range of subsequent uses, since there must be a limit to the ratio of biofuels mixed with mineral raw materials. As a key factor with regards to these biomass sources pose a great risk of causing multiple forms of corrosion both to metallic and non-metallic structural materials. To assess the degree of corrosion risk to a variety of engineering alloys like low-carbon and stainless steels widely used as structural metals, this work is dedicated to investigating corrosion rates of economically reasonable engineering steel alloys in mixtures of raw gas oil and renewable biomass fuel sources under typical co-processing conditions. To model a desulphurising refining process, corrosion tests were carried out with raw mineral gasoline and its mixture with used cooking oil and animal waste lard in relative quantities of 10% (g/g. Co-processing was simulated by batch-reactor laboratory experiments. Experiments were performed at temperatures between 200 and 300ºC and a pressure in the gas phase of 90 bar containing 2% (m3/m3 hydrogen sulphide. The time span of individual tests were varied between 1 and 21 days so that we can conclude about changes in the reaction rates against time exposure of and extrapolate for longer periods of exposure. Initial and integral corrosion rates were defined by a weight loss method on standard size of coupons of all sorts of steel alloys. Corrosion rates of carbon steels indicated a linear increase with temperature and little variation with composition of the biomass fuel sources. Apparent activation energies over the first 24-hour period remained moderate, varying between 35.5 and 50.3 kJ mol−1. Scales developed on carbon steels at higher

  12. Corrosion resistance of canisters for final disposal of spent nuclear fuel

    International Nuclear Information System (INIS)

    Mattsson, E.

    1979-01-01

    A group of Swedish scientists has evaluated from the corrosion point of view three alternative canister types for final disposal of waste from nuclear reactors in boreholes in rock 500 m below ground. Titanium canisters with a wall-thickness of 6 mm and 100 mm thick lead lining have been estimated to have a life of at least thousands of years, and probably tens of thousands of years. Copper canisters with 200-mm-thick walls would last for hundreds of thousands of years. The third type, α-alumina sintered under isostatic pressure, is a very promising canister material

  13. Review of hot corrosion of thermal barrier coatings of gas turbine

    Directory of Open Access Journals (Sweden)

    LIU Yongbao

    2017-03-01

    Full Text Available The review was done in order to make clear the problem of the hot corrosion of the Thermal Barrier Coatings(TBCsduring gas turbine serving. This paper summarizes the factors resulting from the hot corrosion of TBCs during turbine service and classifies methods for enhancing the corrosive resistance of TBCs. A prospective methodology for improving corrosion resistance is also formulated. The main types of corrosion coating include phase reaction, oxidizing of the bond coating, salt-fog corrosion, CMAS corrosion and fuel impurity corrosion. So far, methods for improving the corrosion resistance of TBCs include developing new coating materials, anticorrosive treatment on the surface of TBCs, modifying the stacking configuration and improving the cleansing functions of the gas turbines. In the future, developing new materials with excellent performance will still be the main direction for boosting the improvement of the hot corrosion resistance of TBCs. Simultaneously, improving the tacking configuration and nanotechnology of TBC coatings are potential approaches for improving corrosion resistance. With the development of a Ceramic Matrix Composite (CMC, the focus of the hot corrosion of TBCs may turn to that of Environmental Barrier Coatings (EBCs.

  14. Electron microscopy characterization of an as-fabricated research reactor fuel plate comprised of U-7Mo particles dispersed in an Al-2Si alloy matrix

    International Nuclear Information System (INIS)

    Keiser, Dennis D.; Gan, J.; Jue, J.F.; Miller, B.D.; Clark, C.R.

    2010-01-01

    To understand the microstructural development of nuclear fuel plates during irradiation, it is imperative to know the microstructure of a fuel plate after all the fabrication steps have been completed and before it is inserted into the reactor. To this end, a U-7 wt.% Mo alloy research reactor dispersion fuel plate with Al-2 wt.% Si matrix was destructively examined using scanning and transmission electron microscopy to characterize the developed microstructure after fabrication. Of particular interest for this study was how the Si that was added to the fuel matrix partitioned between the various fuel plate phases during fabrication. Si was added to the matrix so that the microstructure that developed during fuel fabrication would exhibit good irradiation behavior. SEM analysis was used to identify the representative microstructure, the compositions of the various phases, and the partitioning behavior of the fuel and matrix constituents. TEM analysis was employed to definitively identify the phases in the U-7Mo alloy and the phases that formed due to diffusional interactions between the fuel particles and matrix during fuel plate fabrication. The TEM results are the first reported for an as-fabricated U-7 wt.% Mo dispersion fuel plate with an Al alloy matrix. SEM results showed that a significant portion of the original γ-(U-Mo) fuel particles had transformed to a lamellar microstructure, comprised of α-U and either γ or γ' phases, and the fuel/matrix interaction layers were enriched in Si. TEM analysis identified an ordered fcc (U-Mo)(Al-Si) 3 type of phase, which formed at the decomposed U-7Mo/matrix interface and extended into the lamellar microstructure. Some regions of the U-7Mo particles retained the single-phase γ-(U-Mo). Small precipitate phases were observed in the fuel meat matrix that contained Fe, Al, and Si. The Si that is added to the matrix of a U-Mo dispersion fuel plate to improve irradiation performance appears to result in the creation of a

  15. Corrosion/96 conference papers

    International Nuclear Information System (INIS)

    Anon.

    1996-01-01

    Topics covered by this conference include: cathodic protection in natural waters; cleaning and repassivation of building HVAC systems; worldwide opportunities in flue gas desulfurization; advancements in materials technology for use in oil and gas service; fossil fuel combustion and conversion; technology of corrosion inhibitors; computers in corrosion control--modeling and information processing; recent experiences and advances of austenitic alloys; managing corrosion with plastics; corrosion measurement technology; corrosion inhibitors for concrete; refining industry; advances in corrosion control for rail and tank trailer equipment; CO 2 corrosion--mechanisms and control; microbiologically influenced corrosion; corrosion in nuclear systems; role of corrosion in boiler failures; effects of water reuse on monitoring and control technology in cooling water applications; methods and mechanisms of scale and deposit control; corrosion detection in petroleum production lines; underground corrosion control; environmental cracking--relating laboratory results and field behavior; corrosion control in reinforced concrete structures; corrosion and its control in aerospace and military hardware; injection and process addition facilities; progress reports on the results of reinspection of deaerators inspected or repaired per RP0590 criteria; near 100% volume solids coating technology and application methods; materials performance in high temperature environments containing halides; impact of toxicity studies on use of corrosion/scale inhibitors; mineral scale deposit control in oilfield related operations; corrosion in gas treating; marine corrosion; cold climate corrosion; corrosion in the pulp and paper industry; gaseous chlorine alternatives in cooling water systems; practical applications of ozone in recirculating cooling water systems; and water reuse in industry. Over 400 papers from this conference have been processed separately for inclusion on the data base

  16. Development of UO2/PuO2 dispersed in uranium matrix CERMET fuel system for fast reactors

    International Nuclear Information System (INIS)

    Sinha, V.P.; Hegde, P.V.; Prasad, G.J.; Pal, S.; Mishra, G.P.

    2012-01-01

    CERMET fuel with either PuO 2 or enriched UO 2 dispersed in uranium metal matrix has a strong potential of becoming a fuel for the liquid metal cooled fast breeder reactors (LMR’s). In fact it may act as a bridge between the advantages and disadvantages associated with the two extremes of fuel systems (i.e. ceramic fuel and metallic fuel) for fast reactors. At Bhabha Atomic Research Centre (BARC), R and D efforts are on to develop this CERMET fuel by powder metallurgy route. This paper describes the development of flow sheet for preparation of UO 2 dispersed in uranium metal matrix pellets for three different compositions i.e. U–20 wt%UO 2 , U–25 wt%UO 2 and U–30 wt%UO 2 . It was found that the sintered pellets were having excellent integrity and their linear mass was higher than that of carbide fuel pellets used in Fast Breeder Test Reactor programme (FBTR) in India. The pellets were characterized by X-ray diffraction (XRD) technique for phase analysis and lattice parameter determination. The optical microstructures were developed and reported for all the three different U–UO 2 compositions.

  17. Current and future research on corrosion and thermalhydraulic issues of HLM cooled reactors and on LMR fuels for fast reactor systems

    International Nuclear Information System (INIS)

    Knebel, J.U.; Konings, R.J.M.

    2002-01-01

    Heavy liquid metals (HLM) such as lead (Pb) or lead-bismuth eutectic (Pb-Bi) are currently investigated world-wide as coolant for nuclear power reactors and for accelerator driven systems (ADS). Besides the advantages of HLM as coolant and spallation material, e.g. high boiling point, low reactivity with water and air and a high neutron yield, some technological issues, such as high corrosion effects in contact with steels and thermalhydraulic characteristics, need further experimental investigations and physical model improvements and validations. The paper describes some typical HLM cooled reactor designs, which are currently considered, and outlines the technological challenges related to corrosion, thermalhydraulic and fuel issues. In the first part of the presentation, the status of presently operated or planned test facilities related to corrosion and thermalhydraulic questions will be discussed. First approaches to solve the corrosion problem will be given. The approach to understand and model thermalhydraulic issues such as heat transfer, turbulence, two-phase flow and instrumentation will be outlined. In the second part of the presentation, an overview will be given of the advanced fuel types that are being considered for future liquid metal reactor (LMR) systems. Advantages and disadvantages will be discussed in relation to fabrication technology and fuel cycle considerations. For the latter, special attention will be given to the partitioning and transmutation potential. Metal, oxide and nitride fuel materials will be discussed in different fuel forms and packings. For both parts of the presentation, an overview of existing co-operations and networks will be given and the needs for future research work will be identified. (authors)

  18. Chemical process for treatment of corrosive radioactive waste from the fabrication of nuclear fuel

    International Nuclear Information System (INIS)

    Salimin, Z.

    1996-01-01

    Experiments were carried out to decrease fluors concentration in corrosive radioactive liquid waste containing 19430 ppm of fluors by chemical method i.e. combination of coagulation, flocculation and sedimentation. Calcium chloride, aluminium sulfate, resin WWS 116, and mix of aluminium sulfate and resin WWS 116 were used as chemical coagulant on variation of concentration. The precipitates of fluors salt was separated from the solution, and than fluors concentration in solution is analyzed. The best result of optimum Fluors concentration in solution was achieved at 4.88 ppm by using of 1.3 mole/l calcium chloride coagulant at pH 8, the separation factor is 99.97%. (author)

  19. Modeling of the behavior under fuel dispersed irradiation of U-Mo with aluminum matrix from the thermal point of view and its interrelationship with the interdiffusion phase fuel / matrix

    International Nuclear Information System (INIS)

    Moscarda, Maria V.; Taboada, Horacio H.; Rest, J.

    2009-01-01

    Results from postirradiation examinations of U-Mo / Al dispersion fuels plates denotes a strong interrelation and feedback between the fuel-matrix interaction and the fuel temperature, bringing undesired consequences on the total swelling and behavior under irradiation. The present work approaches this problem, modeling the profile of temperatures moment by moment to be able to evaluate the increase of this interaction. The Fast Dart program is used, optimized version of program Dart, developed by Dr. J. Rest in collaboration with Dr. H. Taboada. A subroutine of thermal calculation was implemented in this code, which allowed to calculate the evolution of the interaction between the fuel and the matrix. The results of simulations are compared with the results of postirradiation examinations realized by the Reduced Enrichment for Research and Test Reactors International Program. In particular, a good adjustment in the calculation of the depth of interdiffusion U-Mo/Al is observed, demonstrating a right estimation of the profile of temperatures on the fuel plate. It is considered necessary the inclusion of a model that describes the phases that form in the zone of interaction, denoting its thermal dependency and effects due to the radiation damage. (author)

  20. An Assessment of Alternative Diesel Fuels: Microbiological Contamination and Corrosion Under Storage Conditions

    Science.gov (United States)

    2010-08-01

    example, Clados- parium (Hormoconis) resinae grew in 80 mg water per 1 of kerosene and after 4 weeks incubation, the concentration of water increased...common isolate related to aircraft fuel and MIC is the fungus Hormoeonis resinae (Churchill 1963; Hendey 1964; Videla et al. 1993). de Mele et al. (1979...Videla (1996) demonstrated acid-etched traces of fungal mycelia on aluminum surfaces colonized by //. resinae . de Meybaum and de Schiapparelli

  1. Processes of elimination of activated corrosion products. Chemical decontamination - fuel cleaning

    International Nuclear Information System (INIS)

    Viala, C.; Brun, C.; Neuhaus, R.; Richier, S.; Bachet, M.

    2007-01-01

    The abatement of the individual and collective dose of a PWR imposes to control the source term through different processes implemented during the plant exploitation. When the limits of these different optimization processes are reached, the abatement of dose rates requires the implementation of curative processes. The objective is thus to eliminate the contaminated oxides and deposits present on surfaces free of radiation flux, and eventually on surfaces under radiation flux and on the fuel itself. The chemical decontamination of equipments and systems is the main and universal remedy implemented at different levels. On the other hand, the ultrasonic cleaning of fuel assemblies is a promising process. This paper aims at illustrating these different techniques using concrete examples of application in France and abroad (decontamination during steam generator replacement, decontamination of primary pump scroll in hot workshop, decontamination of loop sections, ultrasonic cleaning of fuel). The description of these different operations stresses on their efficiency in terms of dosimetric gain, duration of implementation, generation of wastes, and recontamination following their implementation. (J.S.)

  2. Production of ZrC Matrix for Use in Gas Fast Reactor Composite Fuels

    International Nuclear Information System (INIS)

    Vasudevamurthy, Gokul; Knight, Travis W.; Roberts, Elwyn; Adams, Thad

    2007-01-01

    Zirconium carbide is being considered as a candidate for inert matrix material in composite nuclear fuel for Gas fast reactors due to its favorable characteristics. ZrC can be produced by the direct reaction of pure zirconium and graphite powders. Such a reaction is exothermic in nature. The reaction is self sustaining once initial ignition has been achieved. The heat released during the reaction is high enough to complete the reaction and achieve partial sintering without any external pressure applied. External heat source is required to achieve ignition of the reactants and maintain the temperature close to the adiabatic temperature to achieve higher levels of sintering. External pressure is also a driving force for sintering. In the experiments described, cylindrical compacts of ZrC were produced by direct combustion reaction. External induction heating combined with varying amounts of external applied pressure was employed to achieve varying degrees of density/porosity. The effect of reactant particle size on the product characteristics was also studied. The samples were characterized for density/porosity, composition and microstructure. (authors)

  3. Synthesize and characterization of a novel anticorrosive cobalt ferrite nanoparticles dispersed in silica matrix (CoFe{sub 2}O{sub 4}-SiO{sub 2}) to improve the corrosion protection performance of epoxy coating

    Energy Technology Data Exchange (ETDEWEB)

    Gharagozlou, M., E-mail: Gharagozlou@icrc.ac.ir [Department of Nanomaterials and Nanocoatings, Institute for Color Science and Technology, P.O. Box 16765-654, Tehran (Iran, Islamic Republic of); Ramezanzadeh, B., E-mail: Rramezanzadeh-bh@icrc.ac.ir [Department of Surface Coatings and Corrosion, Institute for Color Science and Technology (ICST), PO 16765-654, Tehran (Iran, Islamic Republic of); Baradaran, Z. [Department of Nanomaterials and Nanocoatings, Institute for Color Science and Technology, P.O. Box 16765-654, Tehran (Iran, Islamic Republic of)

    2016-07-30

    Highlights: • An anticorrosive cobalt ferrite nanopigment dispersed in silica matrix was synthesized. • The nanopigment showed proper inhibition performance in solution study. • The nanopigment significantly improved the corrosion resistance of the epoxy coating. - Abstract: This study aimed at studying the effect of an anticorrosive nickel ferrite nanoparticle dispersed in silica matrix (NiFe{sub 2}O{sub 4}-SiO{sub 2}) on the corrosion protection properties of steel substrate. NiFe{sub 2}O{sub 4} and NiFe{sub 2}O{sub 4}-SiO{sub 2} nanopigments were synthesized and then characterized by X-ray diffraction (XRD), Fourier transform infrared spectroscopy (FT-IR) and transmission electron microscope (TEM). Then, 1 wt.% of nanopigments was dispersed in an epoxy coating and the resultant nanocomposites were applied on the steel substrates. The corrosion inhibition effects of nanopigments were tested by an electrochemical impedance spectroscopy (EIS) and salt spray test. Results revealed that dispersing nickel ferrite nanoparticles in a silica matrix (NiFe{sub 2}O{sub 4}-SiO{sub 2}) resulted in the enhancement of the nanopigment dispersion in the epoxy coating matrix. Inclusion of 1 wt.% of NiFe{sub 2}O{sub 4}-SiO{sub 2} nanopigment into the epoxy coating enhanced its corrosion protection properties before and after scratching.

  4. Effect of surface treatment on the interfacial contact resistance and corrosion resistance of Fe–Ni–Cr alloy as a bipolar plate for polymer electrolyte membrane fuel cells

    International Nuclear Information System (INIS)

    Yang, Meijun; Zhang, Dongming

    2014-01-01

    The bipolar plate is an important component of the PEMFC (polymer electrolyte membrane fuel cell) because it supplies the pathway of electron flow between each unit cell. Fe–Ni–Cr alloy is considered as a good candidate material for bipolar plate, but it is limited to use as a bipolar plate due to its high ICR (interfacial contact resistance) and corrosion problem. In order to explore a cost-effective method on surface modification, various chemical and electrochemical treatments are performed on Fe–Ni–Cr alloy to acquire the effect of the surface modification on the ICR and corrosion behavior. The ICR and corrosion resistance of Fe–Ni–Cr alloy can be effectively controlled by the chemical treatment of immersion in the mixed acid solution with 10 vol% HNO 3 , 2 vol% HCl and 1 vol% HF for 10 min at 65 °C and then was placed in 30 vol% HNO 3 solution for 5 min. The chemical treatment is more effective on reducing ICR and improving corrosion resistance than that of electrochemical methods (be carried out in the 2 mol/L H 2 SO 4 solution with the electrical potential from −0.4 V to 0.6 V) for Fe–Ni–Cr alloy as a bipolar plate for polymer electrolyte membrane fuel cells. - Highlights: • The procedure of the surface treatments on Fe–Ni–Cr alloy as bipolar plate was described in detail. • Effects of various surface treatments on the interfacial contact resistivity and corrosion behavior were discussed. • The mechanism of the surface modification was particularly analyzed

  5. Neutronic analysis of U-free inert matrix and thoria fuels for plutonium disposition in pressurised water reactors

    Science.gov (United States)

    Lombardi, C.; Mazzola, A.; Padovani, E.; Ricotti, M. E.

    1999-08-01

    Inserting reactor-grade (RG) or weapons-grade (WG) plutonium in uranium-free matrices and burning it in light water reactors (LWRs) is an option gaining a wider consensus in the nuclear community. The main results of our neutronic studies performed in the last few years on this subject are reported. Our attention was mainly concentrated on two kinds of matrices: inert matrix in the form of calcia-stabilised zirconia, and thoria. Both materials are likely to exhibit excellent behaviour under irradiation (already demonstrated for thoria fuels) and high chemical stability. Direct disposal of spent fuel should be made feasible and attractive. A preliminary neutronic analysis was performed on these U-free fuels, imposing the constraint of maintaining the same assembly design and cycle length of a standard enriched-uranium fuel. In particular inert matrix fuel (IMF) showed a high plutonium burning capability, but associated with unacceptable feedback coefficients. Therefore, a whole IMF core results unfeasible, and only a partial core loading is possible. The solution then studied consists in replacing ≈21% of the pins of a standard enriched-U subassembly with IMF pins. Detailed assembly and core calculations were performed. A crucial aspect is the choice of a suitable burnable poison, which has to dampen the power peaks in the different fuel pin types without life penalisation. Among the considered poisons, a thin boron coating on the IMF pellets resulted the only effective one. Preliminary IMF pin cell calculations and the detailed ones gave similar results in terms of burnt plutonium fractions: 90% of fissile and 73% of total plutonium is burnt when RG plutonium is used. The main drawbacks of this fuel are the limited core loading capability and the lack of in-pile technological validation. In the case of Pu-Th fuels, pin cell calculations showed that increasing the plutonia content, decreasing the thoria content, and decreasing the pellet diameter are all possible

  6. Sulfur removal from fuel using zeolites/polyimide mixed matrix membrane adsorbents

    International Nuclear Information System (INIS)

    Lin, Ligang; Wang, Andong; Dong, Meimei; Zhang, Yuzhong; He, Benqiao; Li, Hong

    2012-01-01

    Graphical abstract: Membrane adsorption process is proposed for sulfur removal. Three-dimensional network structure is key to fulfill adsorption function of MMMs, which adsorption/desorption behavior is markedly related with binding force with sulfur molecules. Highlights: ► Membrane adsorption process is proposed for sulfur removal. ► Three-dimensional network structure of MMMs is key to fulfill adsorption function. ► Adsorption/desorption behavior is markedly related with binding force. - Abstract: A novel membrane adsorption process was proposed for the sulfur removal from fuels. The mixed matrix membranes (MMMs) adsorbents composed of polyimide (PI) and various Y zeolites were prepared. By the detailed characterization of FT-IR, morphology, thermal and mechanical properties of MMMs adsorbents, combining the adsorption and desorption behavior research, the process–structure–function relationship was discussed. Field-emission scanning electron microscope (FESEM) images show that the functional particles are incorporated into the three-dimensional network structure. MMMs adsorbents with 40% of zeolites content possess better physical properties, which was confirmed by mechanical strength and thermo stability analysis. Influence factors including post-treatment, content of incorporated zeolites, adsorption time, temperature, initial sulfur concentration as well as sulfur species on the adsorption performance of MMMs adsorbents have been evaluated. At 4 wt.% zeolites content, adsorption capacity for NaY/PI, AgY/PI and CeY/PI MMMs adsorbents come to 2.0, 7.5 and 7.9 mg S/g, respectively. And the regeneration results suggest that the corresponding spent membranes can recover about 98%, 90% and 70% of the desulfurization capacity, respectively. The distinct adsorption and desorption behavior of MMMs adsorbents with various functional zeolites was markedly related with their various binding force and binding mode with sulfur compounds.

  7. Online fuel failure detection and damage severity analysis for thorium based AHWR fuel matrix - an empirical analysis

    International Nuclear Information System (INIS)

    Rajalakshmi, R.; Robin, Roshini; Umashankari, K.; Rama Rao, A.; Vijayan, P.K.

    2015-01-01

    A clad failure, results in the escape of fission products from the fuel to coolant. Continued operation of the reactor with the presence of failed fuel would cause excessive radioactive contamination of the Main Heat Transport (MHT) system and its associated components. Therefore online detection and precise location of failed fuel in the core is necessary for the safe and healthy operation of the reactor and to reduce the man-rem exposure. For the development of online system for iodine and gaseous fission product monitoring for AHWR, an empirical analysis was carried out to compute these fission product release rate data for thorium based AHWR mix-oxide fuel consisting of (Th- 233 U)O 2 and (Th-Pu)O 2 . The release rate and activity concentration rates in the coolant were calculated for various types of fuel failures and a feasibility study was carried out for online gaseous fission product and iodine monitoring using HPGe detector and high resolution gamma ray spectrometer system. Further, this paper also discusses the different methodologies for identifying severity of fuel damage. (author)

  8. Stress corrosion (Astm G30-90 standard) in 08x18H10T stainless steel of nuclear fuel storage pool in WWER reactors

    International Nuclear Information System (INIS)

    Herrera, V.; Zamora R, L.

    1997-01-01

    At the water storage of the irradiated nuclear fuel has been an important factor in its management. The actual pools have its walls covered with inoxidable steel and heat exchangers to dissipate the residual heat from fuel. It is essential to control the water purity to eliminate those conditions which aid to the corrosion process in fuel and at related components. The steel used in this research was obtained from an austenitic inoxidizable steel standardized with titanium 08x18H10T (Type 321) similar to one of the two steel coatings used to cover walls and the pools floor. the test consisted in the specimen deformation through an U ply according to the Astm G30-90 standard. The exposition of the deformed specimen it was realized in simulated conditions to the chemical regime used in pools. (Author)

  9. Superheater corrosion in a boiler fired with refuse-derived fuel

    Energy Technology Data Exchange (ETDEWEB)

    Blough, J.L.; Stanko, G.J. [Foster Wheeler Development Corp., Livingston, NJ (United States); Bakker, W.T. [Electric Power Research Inst., Palo Alto, CA (United States); Steinbeck, T. [United Power Association, Elk River, MN (United States)

    1995-12-31

    The environment in the superheater of a boiler firing refuse-derived fuel (RDF) is very aggressive. The high wastage rate for the standard T-22 material necessitated a materials testing program. Simples of Types 304H, HR3C, T-22 chromized, 825 and 625 were assembled into tubular test sections and welded into the superheater tubing. After 1,180 hours the test sections were evaluated and the wastage rates calculated for each material. The chlorides contained in the RDF are believed to be the primary corrodent. The chlorine may be interacting with the metal samples as HCl, a low-melting-point eutectic or a combination of these. Of the six materials tested, Alloy 625 exhibited the best resistance--substantially better than the next-best Type 304. Alloy 825 and HR3C corroded approximately 1.5 times the rate of Type 304. The chromized layer on T-22 showed no resistance to the environment and was consumed in large areas.

  10. Corrosion behaviour of groundnut shell ash and silicon carbide hybrid reinforced Al-Mg-Si alloy matrix composites in 3.5% NaCl and 0.3M H2SO4 solutions

    Directory of Open Access Journals (Sweden)

    Kenneth Kanayo ALANEME

    2015-05-01

    Full Text Available The corrosion behaviour of Al-Mg-Si alloy based composites reinforced with groundnut shell ash (GSA and silicon carbide (SiC was investigated. The aim is to assess the corrosion properties of Al-Mg-Si alloy based hybrid reinforced composites developed using different mix ratios of GSA (a cheaply processed agro waste derivative which served as partial replacement for SiC and SiC as reinforcing materials. GSA and SiC mixed in weight ratios 0:1, 1:3, 1:1, 3:1, and 1:0 were utilized to prepare 6 and 10 wt% of the reinforcing phase with Al‐Mg‐Si alloy as matrix using two‐step stir casting method. Mass loss and corrosion rate measurement was used to study the corrosion behaviour of the produced composites in 3.5% NaCl and 0.3M H2SO4 solutions. The results show that the Al-Mg-Si alloy based composites containing 6 and 10 wt% GSA and SiC in varied weight ratios were resistant to corrosion in 3.5% NaCl solution. The composites were however more susceptible to corrosion in 0.3M H2SO4 solution (in comparison with the 3.5% NaCl solution. It was noted that the Al-Mg-Si/6 wt% GSA-SiC hybrid composite grades containing GSA and SiC in weight ratio 1:3 and 3:1 respectively exhibited superior corrosion resistance in the 0.3M H2SO4 solution compared to other composites produced for this series. In the case of the Al-Mg-Si/10 wt% GSA-SiC hybrid composite grades, the corrosion resistance was relatively superior for the composites containing a greater weight ratio of GSA (75% and 100% in 0.3M H2SO4 solution.

  11. Study of the behavior of the irradiated fuel in conditions of storage and atmosphere of Hz by means of electrochemical techniques; Estudio del comportamiento del combustible irradiado en condiciones de almacen y atmosfera de H{sub 2} mediante tecnicas electroquimicas

    Energy Technology Data Exchange (ETDEWEB)

    Nieto, J.; Iglesias, E.; Rodriguez, N.; Palomo, C.; Quinones, J.

    2011-07-01

    In studies of alteration (oxidation/dissolution) of the matrix of U0{sub 2} in groundwater, electrochemistry is a technique that allows information about corrosive processes of nuclear fuel in such conditions.

  12. Corrosion and corrosion control

    International Nuclear Information System (INIS)

    Khanna, A.S.; Totlani, M.K.

    1995-01-01

    Corrosion has always been associated with structures, plants, installations and equipment exposed to aggressive environments. It effects economy, safety and product reliability. Monitoring of component corrosion has thus become an essential requirement for the plant health and safety. Protection methods such as appropriate coatings, cathodic protection and use of inhibitors have become essential design parameters. High temperature corrosion, especially hot corrosion, is still a difficult concept to accommodate in corrosion allowance; there is a lack of harmonized system of performance testing of materials at high temperatures. In order to discuss and deliberate on these aspects, National Association for Corrosion Engineers International organised a National Conference on Corrosion and its Control in Bombay during November 28-30, 1995. This volume contains papers presented at the symposium. Paper relevant to INIS is indexed separately. refs., figs., tabs

  13. Fuel assemblies

    International Nuclear Information System (INIS)

    Sadaoka, Noriyuki.

    1986-01-01

    Purpose: To maintain a satisfactory integrity by preventing the increase of corrosion at the outer surface of a fuel can near the point of contact between the fuel can and the spacer due to the use of fuel pellets incorporated with burnable poisons. Constitution: Since reactor coolants are at high temperature and high pressure, zirconium and water are brought into reaction to proceed oxidation at the outer surface of a fuel can to form uniform oxidation layers. However, abrasion corrosion is additionally formed at the contact portion between the spacer and the fuel can, by which the corrosion is increased by about 25 %. For preventing such nodular corrosion, fuel pellets not incorporated with burnable poisons are charged at a portion of the fuel rod where the spacer is supported and fuel pellets incorporated with burnable poisons are charged at the positions other than about to thereby suppress the amount of the corrosion at the portion where the corrosion of the fuel can is most liable to be increased to thereby improve the fuel integrity. That is, radiolysis of coolants due to gamma-rays produced from gadolinium is lowered to reduce the oxygen concentration near the outer surface thereby preventing the corrosion. (Kawakami, Y.)

  14. Nuclear fuels for material test reactors

    International Nuclear Information System (INIS)

    Ramanathan, L.V.; Durazzo, M.; Freitas, C.T. de

    1982-01-01

    Experimental results related do the development of nuclear fuels for reactors cooled and moderated by water have been presented cylindrical and plate type fuels have been described in which the core consists of U compouns dispersed in an Al matrix and is clad with aluminium. Fabrication details involving rollmilling, swaging or hot pressing have been described. Corrosion and irradiation test results are also discussed. The performance of the different types of fuels indicates that it is possible to locally fabricate fuel plates with U 3 O 8 +Al cores (20% enriched U) for use in operating Brazilian research reactors. (Author) [pt

  15. Inert matrix fuel performance during the first two irradiation cycles in a test reactor: comparison with modelling results

    Science.gov (United States)

    Hellwig, Ch.; Kasemeyer, U.

    2003-06-01

    In the inert matrix fuel (IMF) type investigated at Paul Scherrer Institut, plutonium is dissolved in the yttrium stabilised zirconium oxide (YSZ), a highly radiation resistant cubic phase with additions of erbium as burnable poison for reactivity control. A first irradiation experiment of YSZ based IMF is ongoing in the OECD Material Test Reactor in Halden together with mixed oxide fuel. The results of the first two cycles for IMF to a burnup of some 105 kW d cm -3 are presented and the modelling results in comparison with the experimental results are shown. A first approximation for a simple swelling model for this YSZ based IMF can be given. Possible fission gas release mechanisms are briefly discussed. The implications of the modelling results are discussed.

  16. The assessment of the long-term evolution of the spent nuclear fuel matrix by kinetic, thermodynamic and spectroscopic studies of uranium minerals

    International Nuclear Information System (INIS)

    Bruno, J.; Casas, I.; Cera, E.; Ewing, R.C.; Finch, R.J.

    1995-01-01

    The long term behavior of spent nuclear fuel is discussed in the light of recent thermodynamic and kinetic data on mineralogical analogues related to the key phases in the oxidative alteration of uraninite. The implications for the safety assessment of a repository of the established oxidative alteration sequence of the spent fuel matrix are illustrated with Pagoda calculations. The application to the kinetic and thermodynamic data to source term calculations indicates that the appearance and duration of the U(VI) oxyhydroxide transient is critical for the stability of the fuel matrix

  17. Corrosion studies with high burnup light water reactor fuel. Release of nuclides into simulated groundwater during accumulated contact time of up to two years

    Energy Technology Data Exchange (ETDEWEB)

    Zwicky, Hans-Urs (Zwicky Consulting GmbH, Remigen (Switzerland)); Low, Jeanett; Ekeroth, Ella (Studsvik Nuclear AB, Nykoeping (Sweden))

    2011-03-15

    In the framework of comprehensive research work supporting the development of a Swedish concept for the disposal of highly radioactive waste and spent fuel, Studsvik has performed a significant number of spent fuel corrosion studies under a variety of different conditions. These experiments, performed between 1990 and 2002, covered a burnup range from 27 to 49 MWd/kgU, which was typical for fuel to be disposed at that time. As part of this work, the so called Series 11 tests were performed under oxidising conditions in synthetic groundwater with fuel samples from a rod irradiated in the Ringhals 1 Boiling Water Reactor (BWR). In the meantime, Swedish utilities tend to increase the discharge burnup of fuel operated in their reactors. This means that knowledge of spent fuel corrosion performance has to be extended to higher burnup as well. Therefore, a series of experiments has been started at Studsvik, aiming at extending the data base acquired in the Series 11 corrosion tests to higher burnup fuel. Fuel burnup leads to complex and significant changes in the composition and properties of the fuel. The transformed microstructure, which is referred to as the high burnup structure or rim structure in the outer region of the fuel, consists of small grains of submicron size and a high concentration of pores of typical diameter 1 to 2 mum. This structure forms in UO{sub 2} fuel at a local burnup above 50 MWd/kgU, as long as the temperature is below 1,000-1,100 deg C. The high burnup at the pellet periphery is the consequence of plutonium build-up by neutron capture in 238U followed by fission of the formed plutonium. The amount of fission products in the fuel increases more or less linearly with burnup, in contrast to alpha emitting actinides that increase above average. As burnup across a spent fuel pellet is not uniform, but increases towards the periphery, the radiation field is also larger at the pellet surface. At the same time, it is easier for water to access the

  18. Elements of comparison between different inert matrix fuels towards plutonium use and safety coefficients

    International Nuclear Information System (INIS)

    Baldi, St.; Porta, J.

    2000-08-01

    This work deals with the CERMET fuels, chosen for their good behaviour under irradiation and their high thermal conductivity. The kinetic coefficients have been particularly studied. Comparisons have been made with other solutions using other composite fuels in particular the solid solutions and the ROX solution. The core control requiring an heterogeneous assembly, we propose an assembly whose characteristics are compared with those of the APA reference. (O.M.)

  19. Nanoindentation measurements of the mechanical properties of zirconium matrix and hydrides in unirradiated pre-hydrided nuclear fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Rico, A., E-mail: alvaro.rico@urjc.es [DIMME, Departamento de Tecnología Mecánica, Universidad Rey Juan Carlos, c/Tulipán s/n, E-28933 Móstoles, Madrid (Spain); Martin-Rengel, M.A., E-mail: mamartin@mater.upm.es [Departamento de Ciencia de los Materiales, UPM, E.T.S.I. Caminos, Canales y Puertos, Profesor Aranguren SN, E-28040 Madrid (Spain); Ruiz-Hervias, J., E-mail: jesus.ruiz@upm.es [Departamento de Ciencia de los Materiales, UPM, E.T.S.I. Caminos, Canales y Puertos, Profesor Aranguren SN, E-28040 Madrid (Spain); Rodriguez, J. [DIMME, Departamento de Tecnología Mecánica, Universidad Rey Juan Carlos, c/Tulipán s/n, E-28933 Móstoles, Madrid (Spain); Gomez-Sanchez, F.J., E-mail: javier.gomez@amsimulation.com [Advanced Material Simulation, S.L, Madrid (Spain)

    2014-09-15

    It is well known that the mechanical properties of the nuclear fuel cladding may be affected by the presence of hydrides. The average mechanical properties of hydrided cladding have been extensively investigated from a macroscopic point of view. In addition, the mechanical and fracture properties of bulk hydride samples fabricated from zirconium plates have also been reported. In this paper, Young’s modulus, hardness and yield stress are measured for each phase, namely zirconium hydrides and matrix, of pre-hydrided nuclear fuel cladding. To this end, nanoindentation tests were performed on ZIRLO samples in as-received state, on a hydride blister and in samples with 150 and 1200 ppm of hydrogen homogeneously distributed along the hoop direction of the cladding. The results show that the measured mechanical properties of the zirconium hydrides and ZIRLO matrix (Young’s modulus, hardness and yield stress) are rather similar. From the experimental data, the hydride volume fraction in the cladding samples with 150 and 1200 ppm was estimated and the average mechanical properties were calculated by means of the rule of mixtures. These values were compared with those obtained from ring compression tests. Good agreement between the results obtained by both methods was found.

  20. Nanoindentation measurements of the mechanical properties of zirconium matrix and hydrides in unirradiated pre-hydrided nuclear fuel cladding

    International Nuclear Information System (INIS)

    Rico, A.; Martin-Rengel, M.A.; Ruiz-Hervias, J.; Rodriguez, J.; Gomez-Sanchez, F.J.

    2014-01-01

    It is well known that the mechanical properties of the nuclear fuel cladding may be affected by the presence of hydrides. The average mechanical properties of hydrided cladding have been extensively investigated from a macroscopic point of view. In addition, the mechanical and fracture properties of bulk hydride samples fabricated from zirconium plates have also been reported. In this paper, Young’s modulus, hardness and yield stress are measured for each phase, namely zirconium hydrides and matrix, of pre-hydrided nuclear fuel cladding. To this end, nanoindentation tests were performed on ZIRLO samples in as-received state, on a hydride blister and in samples with 150 and 1200 ppm of hydrogen homogeneously distributed along the hoop direction of the cladding. The results show that the measured mechanical properties of the zirconium hydrides and ZIRLO matrix (Young’s modulus, hardness and yield stress) are rather similar. From the experimental data, the hydride volume fraction in the cladding samples with 150 and 1200 ppm was estimated and the average mechanical properties were calculated by means of the rule of mixtures. These values were compared with those obtained from ring compression tests. Good agreement between the results obtained by both methods was found

  1. Study of corrosion kinetics of fuel element tubes from calcium-thermal zirconium alloy Zr1Nb in water at 350 degree C and in vapour at 400 and 500 degree C

    International Nuclear Information System (INIS)

    Petel'guzov, I.A.

    2002-01-01

    In the report brought results of corrosion process studies in water medium of pipe samples for fuel element shells from Zr1Nb alloy (earlier KTZ-110),made from the calcium-thermal zirconium alloys developed in the Ukraine of technology and,for the comparison,samples of pipes from the staff alloy E110, applicable in fuel elements acting reactors of type WWER. Tests were conducted under the working temperature of fuel shells in the reactor (350 degree C) in during of 14000 hours and under increased temperatures (400 degree C) within a time acordinly 4000 hours. Samples from the alloy Zr1Nb had more high contents of oxygen (before 0,12%...0,16%), than staff alloy Eh110 (0,08%O). Studies have shown sufficiently high corrosion stability of experimental alloy Zr1Nb, close to stability of alloy E110.Discovered signs of corrosion 'breakway' or 'transition' on kinetic corrosion curves of Zr1Nb alloys and E110 alloy, characterisating zircaloy type of alloy. Considered mechanism of influence of oxygen on the corrosion process of zirconium alloys with the additive a niobium

  2. Study of the uniform corrosion of an aluminium alloy used for the fuel cladding of the Jules Horowitz experimental reactor; Etude de la corrosion uniforme d'un alliage d'aluminium utilise comme gainage du combustible nucleaire du reacteur experimental Jules Horowitz

    Energy Technology Data Exchange (ETDEWEB)

    Wintergerst, M. [CEA Saclay, Dept. des Materiaux pour le Nucleaire (DEN/DANS/DMN/SEMI), 91 - Gif-sur-Yvette (France)

    2008-07-01

    For the Jules Horowitz new material testing reactor, an aluminium base alloy, AlFeNi, will be used for the cladding of the fuel plates. Taking into account the thermal properties of the alloy and of its oxide, the corrosion of the fuel cans presents many problems. The aim of this thesis is to provide a growing kinetic of the oxide layer at the surface of the AlFeNi fuel can in order to predict the life time of fuel element. Thus the mechanism of degradation of the cladding will be describe in order to integrate the different parameters of the operating reactor. (A.L.B.)

  3. The influence of surface microstructure and chemical composition on corrosion behaviour in fuel-grade bio-ethanol of low-alloy steel modified by plasma nitro-carburizing and post-oxidizing

    Science.gov (United States)

    Boniatti, Rosiana; Bandeira, Aline L.; Crespi, Ângela E.; Aguzzoli, Cesar; Baumvol, Israel J. R.; Figueroa, Carlos A.

    2013-09-01

    The interaction of bio-ethanol on steel surfaces modified by plasma-assisted diffusion technologies is studied for the first time. The influence of surface microstructure and chemical composition on corrosion behaviour of AISI 4140 low-alloy steel in fuel-grade bio-ethanol was investigated. The steel surfaces were modified by plasma nitro-carburizing followed plasma oxidizing. X-ray diffraction, scanning electron microscopy, optical microscopy, X-ray dispersive spectroscopy, and glow-discharge optical emission spectroscopy were used to characterize the modified surface before and after immersion tests in bio-ethanol up to 77 days. The main corrosion mechanism is pit formation. The pit density and pit size were measured in order to quantify the corrosion resistance which was found to depend more strongly on microstructure and morphology of the oxide layer than on its thickness. The best corrosion protection was observed for samples post-oxidized at 480 °C and 90 min.

  4. Analysis of Natural Graphite, Synthetic Graphite, and Thermosetting Resin Candidates for Use in Fuel Compact Matrix

    Energy Technology Data Exchange (ETDEWEB)

    Trammell, Michael P [ORNL; Pappano, Peter J [ORNL

    2011-09-01

    The AGR-1 and AGR-2 compacting process involved overcoating TRISO particles and compacting them in a steel die. The overcoating step is the process of applying matrix to the OPyC layer of TRISO particles in a rotating drum in order to build up an overcoat layer of desired thickness. The matrix used in overcoating is a mixture of natural graphite, synthetic graphite, and thermosetting resin in the ratio, by weight, of 64:16:20. A wet mixing process was used for AGR-1 and AGR-2, in that the graphites and resin were mixed in the presence of ethyl alcohol. The goal of the wet mixing process was to 'resinate' the graphite particles, or coat each individual graphite particle with a thin layer of resin. This matrix production process was similar to the German, Chinese, Japanese, and South African methods, which also use various amount of solvent during mixing. See Appendix 1 for information on these countries matrix production techniques. The resin used for AGR-1 and AGR-2 was provided by Hexion, specifically Hexion grade Durite SC1008. Durite SC1008 is a solvated (liquid) resole phenolic resin. A resole resin does not typically have a hardening agent added. The major constituent of SC1008 is phenol, with minor amounts of formaldehyde. Durite SC1008 is high viscosity, so additional ethyl alcohol was added during matrix production in order to reduce its viscosity and enhance graphite particle resination. The current compacting scale up plan departs from a wet mixing process. The matrix production method specified in the scale up plan is a co-grinding jet mill process where powdered phenolic resin and graphite are all fed into a jet mill at the same time. Because of the change in matrix production style, SC1008 cannot be used in the jet milling process because it is a liquid. The jet milling/mixing process requires that a suite of solid or powdered resins be investigated. The synthetic graphite used in AGR-1 and AGR-2 was provided by SGL Carbon, grade KRB2000. KRB

  5. Influence of environment on the alteration of the UO2 matrix of spent fuel in storage condition

    International Nuclear Information System (INIS)

    Gaulard, C.

    2012-01-01

    Within the framework of the geological disposal of spent nuclear fuel, research on the long term behavior of spent fuel is undertaken and in particular the study of mechanisms of UO 2 oxidation and dissolution in water-saturated host rock. Under the law program on the sustainable management of radioactive materials and waste of June 28, 2006, France was chose as the reference solution the retreatment of spent fuel and disposal in deep geological repository of vitrified final waste. Nevertheless, studies on a direct disposal of spent fuel will continue for safety. The disposal concept provides for conditioning spent fuel in a steel container whose seal is guaranteed for a period specified in the order of 10,000 years. It is also reasonable to assume that the groundwater comes into contact with the fuel after the deterioration of container and lead to the UO 2 matrix degradation and the release of radionuclides. The oxidation/dissolution of UO 2 has been studied by means electrochemical methods coupled to XPS and ICP-MS measurements.A thermodynamic and bibliographic study of U(VI)/UO 2 (s) system allowed to show the effect of the physical and chemical conditions of the solution on the system, and to show the different mechanisms proposed to describe the oxidation and the dissolution of the uranium dioxide in different media (non-complexing, carbonate and clay). The study of the oxidation/dissolution of UO 2 in acidic and non-complexing media (0.1 mol/L NaCF 3 SO 3 , pH = 3), where UO 2 2+ /UO 2 (s) predominates and the formation of precipitates is limited or even avoided, showed a mechanism with two electrochemical steps and a model characteristic of UO 2 oxidation in acidic non-complexing media. Then, the study in neutral non-complexing media (0.05 mol/L NaCl, pH = 7.5) showed a mechanism with two electrochemical steps and one chemical step (EEC) in which both electrochemical steps are similar to those proposed in acidic media. Finally, a first approach of the UO 2

  6. Natural analogues to the spent fuel behaviour of radioactive wastes (MATRIX, FASES I y II projects); Analogos naturales de la liberacion y migracion del UO2 y elementos metalicos asociados (Proyecto MATRIX, FASES I y II)

    Energy Technology Data Exchange (ETDEWEB)

    Perez del Villa, L.; Campos, R.; Garralon, A.; Crespo, M. T.; Quejido, J. A.; Cozar, J. S.; Arcos, D.; Bruno, J.; Grive, M.; Domenech, C.; Duro, L.; Ruiz Sanchez-Prro, J.; Marin, F.; Izquierdo, A.; Cattetero, G.; Ortuno, F.; Floria, E.

    2005-07-01

    Uranium ore deposits have been extensively studied as natural analogues to the spent fuel behaviour of radioactive wastes. These investigations constitute an essential element of both national and international research programmes applied to the assessment of HLNW repositories and their interaction with the environment. The U ore deposit of Mina Fe (Ciudad Rodrigo, Salamanca) is hosted in highly fractured schistose rocks, a geological setting that has not been envisaged in the ENRESA option for nuclear waste disposal. However, the processes occurring at Mina Fe maintain some analogies with those occurring in a HLNW repository: The existence of large U concentrations as pitchblende (UO{sub 2}+x), which is chemically analogous to the main component of spent nuclear fuel, which has an oxidation degree of 2.25 < x < 2.66 as a result of radiolytic oxidation. The solubility behaviour of pitchblende as a result of interaction with groundwaters of varying chemical composition can be used to validate predictive models for spent fuel stability under severe alteration conditions. Some of the weathering products of pitchblende are similar to those that have been identified during the experimental oxidative dissolution of UO{sub 2}, Sim fuel, as well as natural uraninite and pitchblende. This is a subject that has been previously investigated in other research projects. Fe(III)-oxy hydroxides in the oxidised zone of the deposit could be similar to the spent fuel container corrosion products that could be formed under redox transition conditions. These corrosion products may act as radionuclide and trace metal scavengers. (Author)

  7. Irradiation and lithium presence influence on the crystallographic nature of zirconia in the framework of PWR zircaloy 4 fuel cladding corrosion study

    International Nuclear Information System (INIS)

    Gibert, C.

    1999-01-01

    The-increasing deterioration of the initially protective zirconia layer is one of the hypotheses which can explain the impairment with time of PWR fuel cladding corrosion. This deterioration could be worsened by irradiation or lithium presence in the oxidizing medium. The aim of this thesis was to underline the influence of those two parameters on zirconia crystallographic nature. We first studied the impact of ionic irradiation on pure, powdery, monoclinic zirconia and oxidation formed zirconia, mainly with X-ray diffraction and Raman microscopy. The high or low energy particles used (Kr n+- , Ar n+ ) respectively favored electronic or atomic defaults production. The crystallographic analyses showed that these irradiation have a significant effect on zirconia by inducing nucleation or growth of tetragonal phase. The extent depends on sample nature and particles energy. In all cases, phase transformation is correlated with crystalline parameters, grain size and especially micro-stress changes. The results are consistent with those obtained with 1 to 5 cycles PWR claddings. Therefore, the corrosion acceleration observed in reactor can partly be explained by the stress fields appearance under irradiation, which is particularly detrimental to zirconia layer cohesion. Last, we have underlined that the presence of considerable amounts of lithium in the oxidizing medium ((> 700 ppm) induces the disappearance of the tetragonal zirconia located at the metal/oxide interface and the appearance of a porosity of the dense under layer, which looses its protectiveness. (author)

  8. Evaluation of Corrosion of the Dummy “EE” Plate 19 in YA Type ATR Fuel Element During Reactor PALM Cycles

    Energy Technology Data Exchange (ETDEWEB)

    Brower, Jeffrey Owen [Idaho National Lab. (INL), Idaho Falls, ID (United States); Glazoff, Michael Vasily [Idaho National Lab. (INL), Idaho Falls, ID (United States); Eiden, Thomas John [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rezvoi, Aleksey Victor [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-08-01

    Advanced Test Reactor (ATR) Cycle 153B-1 was a 14-day, high-power, powered axial locator mechanism (PALM) operating cycle that completed on April 12, 2013. Cycle 153B-1 was a typical operating cycle for the ATR and did not result in any unusual plant transients. ATR was started up and shut down as scheduled. The PALM drive physically moves the selected experiments into and out of the core to simulate reactor startup and heat up, and shutdown and cooldown transients, while the reactor remains in steady state conditions. However, after the cycle was over, several thousand of the flow-assisted corrosion pits and “horseshoeing” defects were readily observable on the surface of the several YA-type fuel elements (these are “dummy” plates that contain no fuel). In order understand these corrosion phenomena a thermal-hydraulic model of coolant channel 20 on a YA-M fuel element was generated. The boundaries of the model were the aluminum EE plate of a YA-M fuel element and a beryllium reflector block with 13 horizontal saw cuts which represented regions of zero flow. The heat generated in fuel plates 1 through 18 was modeled to be passing through the aluminum EE plate. The coolant channel 20 width was set at 0.058 in. (58 mils). It was established that the horizontal saw cuts had a significant effect on the temperature of the coolant. The flow, which was expected to vary linearly with gradual heating of the coolant as it passed through the channel, was extremely turbulent. The temperature rise, which was expected to be a smooth “S” curve, was represented by a series temperature rise “humps,” which occurred at each horizontal saw cut in the beryllium reflector block. Each of the 13 saw cuts had a chamfered edge which resulted in the coolant flow being re-directed as a jet across the coolant channel into the surface of the EE plate, which explained the temperature rise and the observed sscalloping and possibly pitting degradation on the YA-M fuel elements. In

  9. Evaluation of Corrosion of the Dummy ''EE'' Plate 19 in YA Type ATR Fuel Element During Reactor PALM Cycles

    International Nuclear Information System (INIS)

    Brower, Jeffrey Owen; Glazoff, Michael Vasily; Eiden, Thomas John; Rezvoi, Aleksey Victor

    2016-01-01

    Advanced Test Reactor (ATR) Cycle 153B-1 was a 14-day, high-power, powered axial locator mechanism (PALM) operating cycle that completed on April 12, 2013. Cycle 153B-1 was a typical operating cycle for the ATR and did not result in any unusual plant transients. ATR was started up and shut down as scheduled. The PALM drive physically moves the selected experiments into and out of the core to simulate reactor startup and heat up, and shutdown and cooldown transients, while the reactor remains in steady state conditions. However, after the cycle was over, several thousand of the flow-assisted corrosion pits and ''horseshoeing'' defects were readily observable on the surface of the several YA-type fuel elements (these are ''dummy'' plates that contain no fuel). In order understand these corrosion phenomena a thermal-hydraulic model of coolant channel 20 on a YA-M fuel element was generated. The boundaries of the model were the aluminum EE plate of a YA-M fuel element and a beryllium reflector block with 13 horizontal saw cuts which represented regions of zero flow. The heat generated in fuel plates 1 through 18 was modeled to be passing through the aluminum EE plate. The coolant channel 20 width was set at 0.058 in. (58 mils). It was established that the horizontal saw cuts had a significant effect on the temperature of the coolant. The flow, which was expected to vary linearly with gradual heating of the coolant as it passed through the channel, was extremely turbulent. The temperature rise, which was expected to be a smooth ''S'' curve, was represented by a series temperature rise ''humps,'' which occurred at each horizontal saw cut in the beryllium reflector block. Each of the 13 saw cuts had a chamfered edge which resulted in the coolant flow being re-directed as a jet across the coolant channel into the surface of the EE plate, which explained the temperature rise and the observed

  10. Field test corrosion experiments in Denmark with biomass fuels Part II Co-firing of straw and coal

    DEFF Research Database (Denmark)

    Montgomery, Melanie; Larsen, OH

    2002-01-01

    In Denmark, straw is used for generating energy in power plants. However during straw combustion, potassium chloride and SO2 are released in the flue gas and through condensation and deposition processes they will result in the formation of superheater deposits rich in potassium chloride and pota......In Denmark, straw is used for generating energy in power plants. However during straw combustion, potassium chloride and SO2 are released in the flue gas and through condensation and deposition processes they will result in the formation of superheater deposits rich in potassium chloride...... and potassium sulphate. These components give rise to varying degrees of accelerated corrosion. This paper concerns co-firing of straw with coal to reduce the corrosion rate from straw to an acceptable level. A field investigation at Midtkraft Studstrup suspension-fired power plant in Denmark has been...... undertaken where coal has been co-fired with 10% straw and 20% straw (% energy basis) for up to approx. 3000 hours. Two types of exposure were undertaken to investigate corrosion: a) the exposure of metal rings on water/air cooled probes, and b) the exposure of a range of materials built into the existing...

  11. Evaluation of Annealing Treatments for Producing Si-Rich Fuel/Matrix Interaction Layers in Low-Enriched U-Mo Dispersion Fuel Plates Rolled at a Low Temperature

    International Nuclear Information System (INIS)

    Keiser, Dennis D. Jr.; Jue, Jan-Fong; Woolstenhulme, Nicolas E.

    2010-01-01

    During fabrication of U-7Mo dispersion fuels, exposure to relatively high temperatures affects the final microstructure of a fuel plate before it is inserted into a reactor. One impact of this high temperature exposure is a chemical interaction that can occur between dissimilar materials. For U-7Mo dispersion fuels, the U-7Mo particles will interact to some extent with the Al or Al alloy matrix to produce interaction products. It has been observed that the final irradiation behavior of a fuel plate can depend on the amount of interaction that occurs at the U-7Mo/matrix interface during fabrication, along with the type of phases that develop at this interface. For the case where a U-7Mo dispersion fuel has a Si-containing Al alloy matrix and is rolled at around 500 C, a Si-rich interaction product has been observed to form that can potentially have a positive impact on fuel performance during irradiation. This interaction product can exhibit stable irradiation behavior and it can act as a diffusion barrier to additional U-Mo/matrix interaction during irradiation. However, for U-7Mo dispersion fuels with softer claddings that are rolled at lower temperatures (e.g., near 425 C), a significant interaction layer has not been observed to form. As a result, the bulk of any interaction layer that develops in these fuels happens during irradiation, and the layer that forms may not exhibit as stable a behavior as one that is formed during fabrication. Therefore, it may be beneficial to add a heat treatment step during the fabrication of dispersion fuel plates with softer cladding alloys that will result in the formation of a uniform, Si-rich interaction layer that is a few microns thick around the U-Mo fuel particles. This type of layer would have characteristics like the one that has been observed in dispersion fuel plates with AA6061 cladding that are fabricated at 500 C, which may exhibit increased stability during irradiation. This report discusses the result of

  12. Evaluation of Annealing Treatments for Producing Si-Rich Fuel/Matrix Interaction Layers in Low-Enriched U-Mo Dispersion Fuel Plates Rolled at a Low Temperature

    Energy Technology Data Exchange (ETDEWEB)

    Dennis D. Keiser, Jr.; Jan-Fong Jue; Nicolas E. Woolstenhulme

    2010-06-01

    During fabrication of U-7Mo dispersion fuels, exposure to relatively high temperatures affects the final microstructure of a fuel plate before it is inserted into a reactor. One impact of this high temperature exposure is a chemical interaction that can occur between dissimilar materials. For U-7Mo dispersion fuels, the U-7Mo particles will interact to some extent with the Al or Al alloy matrix to produce interaction products. It has been observed that the final irradiation behavior of a fuel plate can depend on the amount of interaction that occurs at the U-7Mo/matrix interface during fabrication, along with the type of phases that develop at this interface. For the case where a U-7Mo dispersion fuel has a Si-containing Al alloy matrix and is rolled at around 500°C, a Si-rich interaction product has been observed to form that can potentially have a positive impact on fuel performance during irradiation. This interaction product can exhibit stable irradiation behavior and it can act as a diffusion barrier to additional U-Mo/matrix interaction during irradiation. However, for U-7Mo dispersion fuels with softer claddings that are rolled at lower temperatures (e.g., near 425°C), a significant interaction layer has not been observed to form. As a result, the bulk of any interaction layer that develops in these fuels happens during irradiation, and the layer that forms may not exhibit as stable a behavior as one that is formed during fabrication. Therefore, it may be beneficial to add a heat treatment step during the fabrication of dispersion fuel plates with softer cladding alloys that will result in the formation of a uniform, Si-rich interaction layer that is a few microns thick around the U-Mo fuel particles. This type of layer would have characteristics like the one that has been observed in dispersion fuel plates with AA6061 cladding that are fabricated at 500°C, which may exhibit increased stability during irradiation. This report discusses the result of

  13. Corrosion Behavior of Nickel Alloy (ASTM A 494 M) Reinforced with Fused SiO2 Chilled Metal Matrix Composites (MMCs) for Marine Applications

    Science.gov (United States)

    Hemanth, Joel, Dr.

    2017-08-01

    This paper presents the results obtained and the discussions made from a series of corrosion experiments involving Nickel alloy (ASTM A 494 M) reinforced with fused SiO2, size of the particles dispersed varies from 80-120 µm and amount of addition varies from 3 to 12 wt.% in steps of 3 wt.%. The resulting chilled MMCs are solidified under the influence of copper chill of 25 mm thickness to study the effect of corrosion behavior. Corrosion resistance was found to increase significantly with increase in SiO2 content in chilled MMCs. Nevertheless, even with high SiO2 content corrosion attack ie., pitting was found to be most severe during the initial stages of each test but it invariably decreased to a very low value in the later stages, due to the formation of an adherent protective layer on the MMCs developed.

  14. Preparation and properties of hybrid direct methanol fuel cell membranes by embedding organophosphorylated titania submicrospheres into a chitosan polymer matrix

    Energy Technology Data Exchange (ETDEWEB)

    Wu, Hong [Key Laboratory for Green Chemical Technology, School of Chemical Engineering and Technology, Tianjin University, 92 Weijin Road, Nankai District, Tianjin 300072 (China); Tianjin Key Laboratory of Membrane Science and Desalination Technology, Tianjin University, Tianjin 300072 (China); Hou, Weiqiang; Wang, Jingtao; Xiao, Lulu; Jiang, Zhongyi [Key Laboratory for Green Chemical Technology, School of Chemical Engineering and Technology, Tianjin University, 92 Weijin Road, Nankai District, Tianjin 300072 (China)

    2010-07-01

    Organophosphorylated titania submicrospheres (OPTi) are prepared and incorporated into a chitosan (CS) matrix to fabricate hybrid membranes with enhanced methanol resistance and proton conductivity for application in direct methanol fuel cells (DMFC). The pristine monodispersed titania submicrospheres (TiO{sub 2}) of controllable particle size are synthesized through a modified sol-gel method and then phosphorylated by amino trimethylene phosphonic acid (ATMP) via chemical adsorption, which is confirmed by XPS, FTIR and TGA. The morphology and thermal property of the hybrid membranes are explored by SEM and TGA. The ionic cross-linking between the -PO{sub 3}H{sub 2} groups on OPTi and the -NH{sub 2} groups on CS lead to better compatibility between the inorganic fillers and the polymer matrix, as well as a decreased fractional free volume (FFV), which is verified by positron annihilation lifetime spectroscopy (PALS). The effects of particle size and content on the methanol permeability, proton conductivity, swelling and FFV of the membranes are investigated. Compared to pure CS membrane, the hybrid membranes exhibit an increased proton conductivity to an acceptable level of 0.01 S cm{sup -1} for DMFC application and a reduced methanol permeability of 5 x 10{sup -7} cm{sup 2} s{sup -1} at a 2 M methanol feed. (author)

  15. Preparation and properties of hybrid direct methanol fuel cell membranes by embedding organophosphorylated titania submicrospheres into a chitosan polymer matrix

    Science.gov (United States)

    Wu, Hong; Hou, Weiqiang; Wang, Jingtao; Xiao, Lulu; Jiang, Zhongyi

    Organophosphorylated titania submicrospheres (OPTi) are prepared and incorporated into a chitosan (CS) matrix to fabricate hybrid membranes with enhanced methanol resistance and proton conductivity for application in direct methanol fuel cells (DMFC). The pristine monodispersed titania submicrospheres (TiO 2) of controllable particle size are synthesized through a modified sol-gel method and then phosphorylated by amino trimethylene phosphonic acid (ATMP) via chemical adsorption, which is confirmed by XPS, FTIR and TGA. The morphology and thermal property of the hybrid membranes are explored by SEM and TGA. The ionic cross-linking between the -PO 3H 2 groups on OPTi and the -NH 2 groups on CS lead to better compatibility between the inorganic fillers and the polymer matrix, as well as a decreased fractional free volume (FFV), which is verified by positron annihilation lifetime spectroscopy (PALS). The effects of particle size and content on the methanol permeability, proton conductivity, swelling and FFV of the membranes are investigated. Compared to pure CS membrane, the hybrid membranes exhibit an increased proton conductivity to an acceptable level of 0.01 S cm -1 for DMFC application and a reduced methanol permeability of 5 × 10 -7 cm 2 s -1 at a 2 M methanol feed.

  16. The Study of Corrosion and Wear Resistance of Copper Composite Coatings with Inclusions of Carbon Nanomaterials in the Copper Metal Matrix

    Directory of Open Access Journals (Sweden)

    Viktorija MEDELIENĖ

    2011-07-01

    Full Text Available This paper deals with the peculiarities of the behaviour of copper nanocomposite coatings with CNMs inclusions under the free corrosion conditions in the acidic medium. The parameters of corrosion current density (jcorr, anodic dissolution current density (ia and polarization resistance (Rp have been determined. In the acidic medium a stronger oxidation of nanostructured copper nanocomposites occurred. With longer immersion periods more corrosion products are formed, resulting in a increase in the polarization resistance (Rp of corrosion. Corrosion products cover the whole surface of the coatings and the corrosion rate (jcorr tends towards a steady value of 1.7×10-3 ¸ 2.1×10-3 A·cm-2 for all copper coatings studied: 1.7×10-3 A·cm-2 for both Cu and Cu-CNM1, 1.9×10-3 A·cm-2 - for Cu-CNM2 and 2.1×10-3 A·cm-2 - for Cu-CNM3 composite coatings. It has been established that nanocomposites possess a higher wear resistance as compared to that of pure copper. The damage of metal characterized as a depth scar (đ is lower. The roughness of the composites studied was found to be the essential factor affecting their wear resistance. Therefore, the wear resistance of nanocomposites is impaired when they are deposited on a hard steel substrate.http://dx.doi.org/10.5755/j01.ms.17.2.481

  17. Improved performance of U-Mo dispersion fuel by Si addition in Al matrix.

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Y S; Hofman, G L [Nuclear Engineering Division

    2011-06-01

    The purpose of this report is to collect in one publication and fit together work fragments presented in many conferences in the multi-year time span starting 2002 to the present dealing with the problem of large pore formation in U-Mo/Al dispersion fuel plates first observed in 2002. Hence, this report summarizes the excerpts from papers and reports on how we interpreted the relevant results from out-of-pile and in-pile tests and how this problem was dealt with. This report also provides a refined view to explain in detail and in a quantitative manner the underlying mechanism of the role of silicon in improving the irradiation performance of U-Mo/Al.

  18. Radionuclides and isotopes release of spent fuel matrix. Conceptual and mathematical models of wastes behaviour

    International Nuclear Information System (INIS)

    Cera, E.; Merino, J.; Bruno, J.

    2000-01-01

    We have developed a conceptual and numerical model to calculate release of selected radionuclides from spent fuel under repository condition. This has been done in the framework of the Enresa 2000 performance assessment exercise. The model has been developed based on kinetic mass balance equations in order to study the evolution of the spent fuel water interface as a function of time. Several processes have been kinetically modelled: congruent dissolution, radioactive decay, ingrowth and water turnover in the gap. The precipitation/redissolution of secondary solid phases has been taken into account from a thermodynamic point of view. Both approaches have been coupled and the resulting equations solved for a number of radionuclides in both, a conservative and realistic approach. The results show three distinct groups of radionuclides based on their release behaviour: a first group is composed of radioisotopes of highly insoluble elements (e. g., Pu, Am, Pd) whose concentration in the gap is mainly controlled by their solubility and therefore their evolution is identical in both cases. Secondly, a set of radionuclides from soluble elements under these conditions (e. g., I, Cs, Ra) show concentrations kinetically controlled, decreasing with time following the congruent dissolution trend. Their release concentrations are one order of magnitude larger in the conservative case than in the realistic case. Finally, a third group has been identified (e. g., Se, Th, Cm) where a mixed behaviour takes place: initially their solubility limiting phases control their concentration in the gap but the situation reverts to a kinetic control as the chemical conditions change and the secondary precipitates become totally dissolved. The fluxes of the different radionuclides are also given as an assessment of the source term in the performance assessment. (Author)

  19. Corrosion problems in light water nuclear reactors

    International Nuclear Information System (INIS)

    Berry, W.E.

    1984-01-01

    The corrosion problems encountered during the author's career are reviewed. Attention is given to the development of Zircaloys and attendant factors that affect corrosion; the caustic and chloride stress corrosion cracking (SCC) of austenitic stainless steel steam generator tubing; the qualification of Inconel Alloy 600 for steam generator tubing and the subsequent corrosion problem of secondary side wastage, caustic SCC, pitting, intergranular attack, denting, and primary side SCC; and SCC in weld and furnace sensitized stainless steel piping and internals in boiling water reactor primary coolants. Also mentioned are corrosion of metallic uranium alloy fuels; corrosion of aluminum and niobium candidate fuel element claddings; crevice corrosion and seizing of stainless steel journal-sleeve combinations; SCC of precipitation hardened and martensitic stainless steels; low temperature SCC of welded austenitic stainless steels by chloride, fluoride, and sulfur oxy-anions; and corrosion problems experienced by condensers

  20. To the corrosion of austenitic steels in sodium loops

    International Nuclear Information System (INIS)

    Schad, M.

    1978-03-01

    This report describes the comparison between experimental corrosion and calculated corrosion effects on austenitic steels exposed to liquid sodium. As basis for the calculations served a diffusion model. The comparison showed that the model is able to predict the corrosion effects. In addition the model was used to calculate the corrosion effect along an actual fuel rod. (orig.) [de

  1. Effect of WO3 nanoparticle loading on the microstructural, mechanical and corrosion resistance of Zn matrix/TiO2-WO3 nanocomposite coatings for marine application

    Science.gov (United States)

    Popoola, A. P. I.; Daniyan, A. A.; Umoru, L. E.; Fayomi, O. S. I.

    2017-03-01

    In this study, for marine application purposes, we evaluated the effect of process parameter and particle loading on the microstructure, mechanical reinforcement and corrosion resistance properties of a Zn-TiO2-WO3 nanocomposite produced via electrodeposition. We characterized the morphological properties of the composite coatings with a Scanning Electron Microscope (SEM) equipped with an Energy Dispersive Spectrometer (EDS). We carried out mechanical examination using a Dura Scan hardness tester and a CERT UMT-2 multi-functional tribological tester. We evaluated the corrosion properties by linear polarization in 3.5% NaCl. The results show that the coatings exhibited good stability and the quantitative particle loading greatly enhanced the structural and morphological properties, hardness behavior and corrosion resistance of the coatings. We observed the precipitation of this alloy on steel is greatly influenced by the composite characteristics.

  2. Effect of Burnable Absorbers on Inert Matrix Fuel Performance and Transuranic Burnup in a Low Power Density Light-Water Reactor

    Directory of Open Access Journals (Sweden)

    Geoff Recktenwald

    2013-04-01

    Full Text Available Zirconium dioxide has received particular attention as a fuel matrix because of its ability to form a solid solution with transuranic elements, natural radiation stability and desirable mechanical properties. However, zirconium dioxide has a lower coefficient of thermal conductivity than uranium dioxide and this presents an obstacle to the deployment of these fuels in commercial reactors. Here we show that axial doping of a zirconium dioxide based fuel with erbium reduces power peaking and fuel temperature. Full core simulations of a modified AP1000 core were done using MCNPX 2.7.0. The inert matrix fuel contained 15 w/o transuranics at its beginning of life and constituted 28% of the assemblies in the core. Axial doping reduced power peaking at startup by more than ~23% in the axial direction and reduced the peak to average power within the core from 1.80 to 1.44. The core was able to remain critical between refueling while running at a simulated 2000 MWth on an 18 month refueling cycle. The results show that the reactor would maintain negative core average reactivity and void coefficients during operation. This type of fuel cycle would reduce the overall production of transuranics in a pressurized water reactor by 86%.

  3. Magnesium alloys and graphite wastes encapsulated in cementitious materials: Reduction of galvanic corrosion using alkali hydroxide activated blast furnace slag

    International Nuclear Information System (INIS)

    Chartier, D.; Muzeau, B.; Stefan, L.; Sanchez-Canet, J.; Monguillon, C.

    2017-01-01

    Highlights: • Embedded in cement, magnesium is corroded by residual water present in porosity of the matrix. • Corrosion is enhanced by galvanic phenomenon when magnesium is in contact with graphite. • Galvanic corrosion of magnesium in contact with graphite debris is shown to be severe with ordinary Portland cement. • Galvanic corrosion is significantly lowered in high alkali medium such as sodium hydroxide. • Sodium hydroxide activated blast furnace slag is a convenient binder to embed magnesium. - Abstract: Magnesium alloys and graphite from spent nuclear fuel have been stored together in La Hague plant. The packaging of these wastes is under consideration. These wastes could be mixed in a grout composed of industrially available cement (Portland, calcium aluminate…). Within the alkaline pore solution of these matrixes, magnesium alloys are imperfectly protected by a layer of Brucite resulting in a slow corrosion releasing hydrogen. As the production of this gas must be considered for the storage safety, and the quality of wasteform, it is important to select a cement matrix capable of lowering the corrosion kinetics. Many types of calcium based cements have been tested and most of them have caused strong hydrogen production when magnesium alloys and graphite are conditioned together because of galvanic corrosion. Exceptions are binders based on alkali hydroxide activated ground granulated blast furnace slag (BFS) which are presented in this article.

  4. Magnesium alloys and graphite wastes encapsulated in cementitious materials: Reduction of galvanic corrosion using alkali hydroxide activated blast furnace slag

    Energy Technology Data Exchange (ETDEWEB)

    Chartier, D., E-mail: david.chartier@cea.fr [Commissariat à l' Energie Atomique et aux Energies Alternatives, CEA, DEN, DTCD, SPDE, F-30207 Bagnols-sur-Cèze (France); Muzeau, B. [DEN-Service d’Etude du Comportement des Radionucléides (SECR), CEA, Université Paris-Saclay, F-91191, Gif-sur-Yvette (France); Stefan, L. [AREVA NC/D& S - France/Technical Department, 1 place Jean Millier 92084 Paris La Défense (France); Sanchez-Canet, J. [Commissariat à l' Energie Atomique et aux Energies Alternatives, CEA, DEN, DTCD, SPDE, F-30207 Bagnols-sur-Cèze (France); Monguillon, C. [DEN-Service d’Etude du Comportement des Radionucléides (SECR), CEA, Université Paris-Saclay, F-91191, Gif-sur-Yvette (France)

    2017-03-15

    Highlights: • Embedded in cement, magnesium is corroded by residual water present in porosity of the matrix. • Corrosion is enhanced by galvanic phenomenon when magnesium is in contact with graphite. • Galvanic corrosion of magnesium in contact with graphite debris is shown to be severe with ordinary Portland cement. • Galvanic corrosion is significantly lowered in high alkali medium such as sodium hydroxide. • Sodium hydroxide activated blast furnace slag is a convenient binder to embed magnesium. - Abstract: Magnesium alloys and graphite from spent nuclear fuel have been stored together in La Hague plant. The packaging of these wastes is under consideration. These wastes could be mixed in a grout composed of industrially available cement (Portland, calcium aluminate…). Within the alkaline pore solution of these matrixes, magnesium alloys are imperfectly protected by a layer of Brucite resulting in a slow corrosion releasing hydrogen. As the production of this gas must be considered for the storage safety, and the quality of wasteform, it is important to select a cement matrix capable of lowering the corrosion kinetics. Many types of calcium based cements have been tested and most of them have caused strong hydrogen production when magnesium alloys and graphite are conditioned together because of galvanic corrosion. Exceptions are binders based on alkali hydroxide activated ground granulated blast furnace slag (BFS) which are presented in this article.

  5. A review of materials and corrosion issues regarding canisters for disposal of spent fuel and high-level waste in Opalinus clay

    International Nuclear Information System (INIS)

    Landolt, D.; Davenport, A.; Payer, J.; Shoesmith, D.

    2009-01-01

    The project 'Entsorgungsnachweis' presented by NAGRA to the Swiss Federal Government in December 2002 assessed the feasibility of disposal of spent fuel (SF), vitrified high level waste (HLW) from reprocessing and long-lived intermediate level waste in an Opalinus Clay repository site in Northern Switzerland. NAGRA proposed the use of carbon steel canisters for disposal of SF/HLW and it also put forward an alternative concept of copper canisters with cast iron insert. In its reply the Federal Government acknowledged that NAGRA had successfully demonstrated the technical feasibility of disposal of SF/HLW. However, some of its experts raised a number of questions related to the choice of steel as canister material. Among others, it was questioned whether hydrogen formed by corrosion of steel in contact with saturated bentonite might adversely affect the barrier function of the Opalinus clay. It was also recommended that alternative canister materials and/or design concepts should be evaluated. To deal with these concerns NAGRA convened an international group of experts, the Canister Materials Review Board (CMRB), who were to review the existing information on canister materials that could be suitable for the proposed repository environment. Based on present knowledge of materials science, the CMRB was to recommend to NAGRA the most suitable material(s) for meeting the performance requirements for SF/HLW canisters. Specifically, the CMRB was to consider corrosion, including hydrogen generation, and stress-assisted failure processes that could affect the integrity and projected life time of SF/HLW canisters or impede the functioning of geological barriers while keeping in mind the overall feasibility of manufacturing, sealing and inspecting the canisters. The CMRB was further asked to identify the needs and provide advice for further studies by NAGRA on the long term performance and safety of SF/HLW canisters in the Swiss repository concept. For the assessment of the

  6. Burn-up calculation of different thorium-based fuel matrixes in a thermal research reactor using MCNPX 2.6 code

    Directory of Open Access Journals (Sweden)

    Gholamzadeh Zohreh

    2014-12-01

    Full Text Available Decrease of the economically accessible uranium resources and the inherent proliferation resistance of thorium fuel motivate its application in nuclear power systems. Estimation of the nuclear reactor’s neutronic parameters during different operational situations is of key importance for the safe operation of nuclear reactors. In the present research, thorium oxide fuel burn-up calculations for a demonstrative model of a heavy water- -cooled reactor have been performed using MCNPX 2.6 code. Neutronic parameters for three different thorium fuel matrices loaded separately in the modelled thermal core have been investigated. 233U, 235U and 239Pu isotopes have been used as fissile element in the thorium oxide fuel, separately. Burn-up of three different fuels has been calculated at 1 MW constant power. 135X and 149Sm concentration variations have been studied in the modelled core during 165 days burn-up. Burn-up of thorium oxide enriched with 233U resulted in the least 149Sm and 135Xe productions and net fissile production of 233U after 165 days. The negative fuel, coolant and void reactivity of the used fuel assures safe operation of the modelled thermal core containing (233U-Th O2 matrix. Furthermore, utilisation of thorium breeder fuel demonstrates several advantages, such as good neutronic economy, 233U production and less production of long-lived α emitter high radiotoxic wastes in biological internal exposure point of view

  7. Corrosion behavior of Zr-x(Nb, Sn and Cu) binary alloys

    International Nuclear Information System (INIS)

    Kim, M. H.; Lee, M. H.; Park, S. Y.; Jung, Y. H.; We, M. Y.

    1999-01-01

    For the development of advanced zirconium alloys for nuclear fuel cladding, the corrosion behaviors of zirconium binary alloys were studied on the Zr-xNb, Zr-xSn, and Zr-xCu alloys. The corrosion test were performed in water at 360 deg C, steam at 400 deg C and LiOH at 360 deg C for 45 days. The corrosion behaviors of Zr-xNb was similar to that of Zr-xCu alloys. However, the corrosion behavior of Zr-xSn was different from Zr-xNb and Zr-xCu. The weight gain of Zr-xNb and Zr-xCu was increased with addition of alloying elements. When Sn is added to Zr matrix in range below the solubility limit, the corrosion resistance decrease with increasing Sn-content, while in the range over solubility limit, Sn has an adverse effect on the corrosion resistance. Especially, Zr-xSn alloys showed higher corrosion resistance than Zr-xNb and Zr-xCu alloys in LiOH solution

  8. The influence of pluronic P123 micelles on corrosion behaviour of steel in cement extract and bulk matrix properties of cement paste

    NARCIS (Netherlands)

    Koleva, D.A.; Denkova, A.G.; Hu, J.; van Breugel, K.

    2012-01-01

    The influence of Pluronic P123 (PEO20-PPO20-PEO70) micelles (of 10 nm size) on the corrosion behaviour of low carbon steel in cement extract (CE) was studied using electrochemical impedance spectroscopy (EIS) and potentio-dynamic polarisation (PDP). Additionally, mercury intrusion porosimetry (MIP)

  9. Penetration of corrosion products and corrosion-induced cracking in reinforced cementitious materials

    DEFF Research Database (Denmark)

    Michel, Alexander; Pease, Brad J.; Peterova, Adela

    2014-01-01

    -dependent concentrations of corrosion products averaged through the specimen thickness. Digital image correlation (DIC) was used to measure corrosion-induced deformations including deformations between steel and cementitious matrix as well as formation and propagation of corrosion-induced cracks. Based on experimental...... observations, a conceptual model was developed to describe the penetration of solid corrosion products into capillary pores of the cementitious matrix. Only capillary pores within a corrosion accommodating region (CAR), i.e. in close proximity of the steel reinforcement, were considered accessible...

  10. Aircraft Corrosion

    Science.gov (United States)

    1981-08-01

    transport anciens dont il a fallu remplacer les panneaux extrados en raison de is corrosion profonde des lisses. Ce type d’avion a etc mis en service h partir...as six years) where lavatories are placed above the lower section of the aft pressure dome bulkhead (fig.8)., corrosion Is found in the bulkhead web ...due to leakage of toilet fluids. One of the most severe corrosion ever found In this area was with the corrosion completely penetrated through the web

  11. Pitting corrosion

    International Nuclear Information System (INIS)

    Allgaier, W.

    1985-01-01

    Pitting corrosion is a kind of electrolytic corrosion by which the surface of a material is locally affected owing to inhomogeneities on the part of the material or medium. The paper deals briefly with questions relating to the importance, to parameters medium or materialwise, influence on production and construction, as well as the general conditions for pitting corrosion. In particular oxygen corrosion in unalloyed and low-alloy steel, and pitting corrosion in ferritic chromium-steel and austenitic chromium-nickel (molybdenum) steel is described. (DG) [de

  12. Role of Chloride in the Corrosion and Fracture Behavior of Micro-Alloyed Steel in E80 Simulated Fuel Grade Ethanol Environment

    Directory of Open Access Journals (Sweden)

    Olufunmilayo O. Joseph

    2016-06-01

    Full Text Available In this study, micro-alloyed steel (MAS material normally used in the production of auto parts has been immersed in an E80 simulated fuel grade ethanol (SFGE environment and its degradation mechanism in the presence of sodium chloride (NaCl was evaluated. Corrosion behavior was determined through mass loss tests and electrochemical measurements with respect to a reference test in the absence of NaCl. Fracture behavior was determined via J-integral tests with three-point bend specimens at an ambient temperature of 27 °C. The mass loss of MAS increased in E80 with NaCl up to a concentration of 32 mg/L; beyond that threshold, the effect of increasing chloride was insignificant. MAS did not demonstrate distinct passivation behavior, as well as pitting potential with anodic polarization, in the range of the ethanol-chloride ratio. Chloride caused pitting in MAS. The fracture resistance of MAS reduced in E80 with increasing chloride. Crack tip blunting decreased with increasing chloride, thus accounting for the reduction in fracture toughness.

  13. Role of Chloride in the Corrosion and Fracture Behavior of Micro-Alloyed Steel in E80 Simulated Fuel Grade Ethanol Environment.

    Science.gov (United States)

    Joseph, Olufunmilayo O; Loto, Cleophas A; Sivaprasad, Seetharaman; Ajayi, John A; Tarafder, Soumitra

    2016-06-16

    In this study, micro-alloyed steel (MAS) material normally used in the production of auto parts has been immersed in an E80 simulated fuel grade ethanol (SFGE) environment and its degradation mechanism in the presence of sodium chloride (NaCl) was evaluated. Corrosion behavior was determined through mass loss tests and electrochemical measurements with respect to a reference test in the absence of NaCl. Fracture behavior was determined via J-integral tests with three-point bend specimens at an ambient temperature of 27 °C. The mass loss of MAS increased in E80 with NaCl up to a concentration of 32 mg/L; beyond that threshold, the effect of increasing chloride was insignificant. MAS did not demonstrate distinct passivation behavior, as well as pitting potential with anodic polarization, in the range of the ethanol-chloride ratio. Chloride caused pitting in MAS. The fracture resistance of MAS reduced in E80 with increasing chloride. Crack tip blunting decreased with increasing chloride, thus accounting for the reduction in fracture toughness.

  14. Study of the stress corrosion cracking of zirconium alloys in iodine vapor and its relation with nuclear fuel element failures for pellet-cladding interaction

    International Nuclear Information System (INIS)

    Bojorge, Claudia D.

    2003-01-01

    Fission product induced Stress Corrosion Cracking (SCC) susceptibility of Zirconium alloy cladding is a relevant attribute to be considered when evaluating nuclear fuel performance in power reactors. It is normally introduced in the codes through a typical parameter, namely the threshold stress needed to start a brittle fissure in normal reactor operating conditions. Iodine induced SCC of Zircaloy-4 cladding was measured at 350 and 400 C degrees by the pressurised-tube technique, method which permits an accurate control of the tensile hoop stress applied on the material. Materials with two different thermo-mechanical treatments were used, essentially differing in the number of roll steps used to arrive to tube finale dimensions. Threshold stress values of about 180-200 MPa were obtained in both cases at 400 C degrees, whereas the results at 350 C degrees have been strongly affected by the different fabrication routes. From the microscopic failure analysis carried out using optical and electronic microscopy, it was concluded that improper treatment may produce surface defects which can act as SCC initiators, giving way to failure at much lower applied stresses than in the case of well finished tubes [es

  15. Plastics for corrosion inhibition

    CERN Document Server

    Goldade, Victor A; Makarevich, Anna V; Kestelman, Vladimir N

    2005-01-01

    The development of polymer composites containing inhibitors of metal corrosion is an important endeavour in modern materials science and technology. Corrosion inhibitors can be located in a polymer matrix in the solid, liquid or gaseous phase. This book details the thermodynamic principles for selecting these components, their compatibility and their effectiveness. The various mechanisms of metal protection – barrier, inhibiting and electromechanical – are considered, as are the conflicting requirements placed on the structure of the combined material. Two main classes of inhibited materials (structural and films/coatings) are described in detail. Examples are given of structural plastics used in friction units subjected to mechano-chemical wear and of polymer films/coatings for protecting metal objects against corrosion.

  16. Annealing Effect on Corrosion Behavior of the Beta-Quenched HANA Alloy

    International Nuclear Information System (INIS)

    Kim, Hyun Gil; Kim, Il Hyun; Choi, Byung Kwan; Park, Sang Yoon; Park, Jeong Yong; Jeong, Yong Hwan

    2009-01-01

    The advanced fuel cladding materials named as HANA cladding have been developed at KAERI for application of high burn-up and that cladding showed an improved performance in both in-pile and out-of-pile conditions. However, the cladding performance could be changed by the annealing conditions during the tube manufacturing process. Especially, the corrosion resistance is considerably sensitive to their microstructure which is determined by a manufacturing process in the high Nb-containing zirconium alloys. They reported that the corrosion properties of the Nb-containing Zr alloys were considerably affected by the microstructure conditions such as the Nb concentration in the matrix and the second phase types. Therefore, the corrosion behavior of HANA cladding having the high Nb could be considerably affected by the annealing time and temperatures. The purpose of this study is focused on the annealing effect of the beta quenched HANA alloy to obtain the optimum annealing conditions

  17. Effect of Carbide Dissolution on Chlorine Induced High Temperature Corrosion of HVOF and HVAF Sprayed Cr3C2-NiCrMoNb Coatings

    Science.gov (United States)

    Fantozzi, D.; Matikainen, V.; Uusitalo, M.; Koivuluoto, H.; Vuoristo, P.

    2018-01-01

    Highly corrosion- and wear-resistant thermally sprayed chromium carbide (Cr3C2)-based cermet coatings are nowadays a potential highly durable solution to allow traditional fluidized bed combustors (FBC) to be operated with ecological waste and biomass fuels. However, the heat input of thermal spray causes carbide dissolution in the metal binder. This results in the formation of carbon saturated metastable phases, which can affect the behavior of the materials during exposure. This study analyses the effect of carbide dissolution in the metal matrix of Cr3C2-50NiCrMoNb coatings and its effect on chlorine-induced high-temperature corrosion. Four coatings were thermally sprayed with HVAF and HVOF techniques in order to obtain microstructures with increasing amount of carbide dissolution in the metal matrix. The coatings were heat-treated in an inert argon atmosphere to induce secondary carbide precipitation. As-sprayed and heat-treated self-standing coatings were covered with KCl, and their corrosion resistance was investigated with thermogravimetric analysis (TGA) and ordinary high-temperature corrosion test at 550 °C for 4 and 72 h, respectively. High carbon dissolution in the metal matrix appeared to be detrimental against chlorine-induced high-temperature corrosion. The microstructural changes induced by the heat treatment hindered the corrosion onset in the coatings.

  18. Corrosion protection

    Science.gov (United States)

    Brown, Donald W.; Wagh, Arun S.

    2003-05-27

    There has been invented a chemically bonded phosphate corrosion protection material and process for application of the corrosion protection material for corrosion prevention. A slurry of iron oxide and phosphoric acid is used to contact a warm surface of iron, steel or other metal to be treated. In the presence of ferrous ions from the iron, steel or other metal, the slurry reacts to form iron phosphates which form grains chemically bonded onto the surface of the steel.

  19. Corrosion aspects in reprocessing technology

    International Nuclear Information System (INIS)

    Fauvet, P.; Pinard Legry, G.

    1992-01-01

    This paper presents two examples illustrating the importance of the physicochemical conditions existing at the metal-medium interface on the corrosion behaviour of materials utilized in spent fuel reprocessing plants: corrosion of a stainless steel in the presence of nitric acid condensates, which is much more severe than in the liquid bulk; behaviour of zirconium, which has an outstanding corrosion resistance in nitric acid, but may suffer depassivation in drastic conditions (not existing in reprocessing plants), with the consequence of a loss of the protective effect of the zirconia passive layer

  20. Study of corrosion and deposit on fuel cladding under heat flux. Adjustment of an original experimental device

    Energy Technology Data Exchange (ETDEWEB)

    Foucault, M.; Albinet, B.; Lucotte, J.P.; Thomazet, J. [Framatome ANP (France)

    2002-07-01

    I INTRODUCTION The cycle extension imposes an increase of the boric acid concentration of the primary environment at the beginning of the cycles. This evolution modifies the chemical and electrochemical conditions in the primary system. The main effect is to reduce the pH of the environment. The solubility of metal cations is dependent on the temperature and the pH of the solution. The solubility limit of iron is decreasing with the temperature in the range 300 - 360 C if the pH is lowered than a critical value which is near 5.8 [1]. The figure 1 gives for instance the evolution for iron solubility limit with temperature in this pH condition. Thus the metal release of the primary system can be the source of deposit on fuel cladding during cycles. In parallel, in the modern PWR's, the surface power density is increased, leading to higher surface temperature with risk of boiling. (authors)

  1. Stress corrosion (Astm G30-90 standard) in 08x18H10T stainless steel of nuclear fuel storage pool in WWER reactors; Corrosion bajo esfuerzo (Norma ASTM G30-90) en acero inoxidable 08x18H10T de piscinas de almacenamiento de combustible nuclear en reactores V.V.E.R

    Energy Technology Data Exchange (ETDEWEB)

    Herrera, V.; Zamora R, L. [Centro de Estudios Aplicados al Desarrollo Nuclear (Cuba)

    1997-07-01

    At the water storage of the irradiated nuclear fuel has been an important factor in its management. The actual pools have its walls covered with inoxidable steel and heat exchangers to dissipate the residual heat from fuel. It is essential to control the water purity to eliminate those conditions which aid to the corrosion process in fuel and at related components. The steel used in this research was obtained from an austenitic inoxidizable steel standardized with titanium 08x18H10T (Type 321) similar to one of the two steel coatings used to cover walls and the pools floor. the test consisted in the specimen deformation through an U ply according to the Astm G30-90 standard. The exposition of the deformed specimen it was realized in simulated conditions to the chemical regime used in pools. (Author)

  2. Effects of gadolinium content in the Gd-doped UO{sub 2} pellet on the corrosion current density

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jandee; Lee, Jeongmook; Yoon, Young-Sang; Kim, Jong-Goo; Ha, Yeong-Keong; Bae, Sang-Eun; Kim, Jong-Yun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Liu, Nazhen [Western University, London (Canada)

    2015-10-15

    Container failure is not expected, since model calculations predict a loss of wall thickness by general corrosion of only a few microns. However, it can be assumed that some containers will fail allowing groundwater to contact the fuel causing radionuclide release. Because the majority of the radio nuclides in used fuel are located within the oxide matrix, their release rate to the groundwater will be controlled by the fuel corrosion/dissolution rate. In determining fuel corrosion and radionuclide release rates, the reactivity of the UO{sub 2} matrix and how it is modified by prior irradiation, is important. The one of the important changes expected to influence the reactivity of the fuel is the rare earth doping of UO{sub 2} which will change the conductivity and structural properties of the UO{sub 2} matrix. In this work, UO{sub 2} and Gd doped UO{sub 2} pellets were prepared and investigated to study about influences of trivalent rare earth of Gd(III) doped UO{sub 2} on corrosion and dissolution rates. Gd doped UO{sub 2} pellets are prepared by compress the Gd{sub 2}O{sub 3} and UO{sub 2} mixed powder following sinter at high temperature in reducing mood. The reproducibility of pellets preparation has been verified by observing resistivity of pellets using 4-probe method. The effect of Gd doping (0 to 10 mol% of Gd) on the electrochemical behavior of UO{sub 2} has been investigated in slightly alkaline solutions with carbonate.

  3. Corrosion Engineering.

    Science.gov (United States)

    White, Charles V.

    A description is provided for a Corrosion and Corrosion Control course offered in the Continuing Engineering Education Program at the General Motors Institute (GMI). GMI is a small cooperative engineering school of approximately 2,000 students who alternate between six-week periods of academic study and six weeks of related work experience in…

  4. Corrosion and alteration of materials from the nuclear industry

    International Nuclear Information System (INIS)

    Beauvy, M.; Berthoud, G.; Defranceschi, M.; Ducros, G.; Feron, D.; Guerin, Y.; Latge, C.; Limoge, Y.; Madic, C.; Santarini, G.; Seiler, J.M.; Vernaz, E.; Richet, C.

    2010-01-01

    The control of the corrosion phenomenon is of prime importance for the nuclear industry. The efficiency and the safety of facilities can be affected by this phenomenon. The nuclear industry has to face corrosion for a large variety of materials submitted to various environments. Metallic corrosion operates in the hot and aqueous environment of water reactors which represent the most common reactor type in the world. Progresses made in the control of the corrosion of the different components of these reactors allow to improve their safety. Corrosion is present in the facilities of the back-end of the fuel cycle as well (corrosion in acid environment in fuel reprocessing plants, corrosion of waste containers in disposal and storage facilities, etc). The future nuclear systems will widen even more the range of materials to be studied and the situations in which they will be placed (corrosion by liquid metals or by helium impurities). Very often, corrosion looks like a patchwork of particular cases in its description. The encountered corrosion problems and their study are presented in this book according to chapters representing the main sectors of the nuclear industry and classified with respect to their phenomenology. This monograph illustrates the researches in progress and presents some results of particular importance obtained recently. Content: 1 - Introduction: context, stakes and goals; definition of corrosion; a complex science; corrosion in the nuclear industry; 2 - corrosion in water reactors - phenomenology, mechanisms, remedies: A - uniform corrosion: mechanisms, uniform corrosion of fuel cladding, in-situ measurement of generalized corrosion rate by electrochemical methods, uniform corrosion of nickel alloys, characterization of the passive layer and growth mechanisms, the PACTOLE code - an integrating tool, influence of water chemistry on corrosion and contamination, radiolysis impact on uniform corrosion; B - stress corrosion: stress corrosion cracking

  5. Growth of the interaction layer around fuel particles in dispersion fuel

    International Nuclear Information System (INIS)

    Olander, D.

    2009-01-01

    Corrosion of uranium particles in dispersion fuel by the aluminum matrix produces interaction layers (an intermetallic-compound corrosion product) around the shrinking fuel spheres. The rate of this process was modeled as series resistances due to Al diffusion through the interaction layer and reaction of aluminum with uranium in the fuel particle to produce UAl x . The overall kinetics are governed by the relative rates of these two steps, the slowest of which is reaction at the interface between Al in the interaction layer and U in the fuel particle. The substantial volume change as uranium is transferred from the fuel to the interaction layer was accounted for. The model was compared to literature data on in-reactor growth of the interaction layer and the Al/U gradient in this layer, the latter measured in ex-reactor experiments. The rate constant of the Al-U interface reaction and the diffusivity of Al in the interaction layer were obtained from this fitting procedure. The second feature of the corrosion process is the transfer of fission products from the fuel particle to the interaction layer due to the reaction. It is commonly assumed that the observed swelling of irradiated fuel elements of this type is due to release of fission gas in the interaction layer to form large bubbles. This hypothesis was tested by using the model to compute the quantity of fission gas available from this source and comparing the pressure of the resulting gas with the observed swelling of fuel plates. It was determined that the gas pressure so generated is too small to account for the observed delamination of the fuel

  6. Thermal Conductivity Measurement and Analysis of Fully Ceramic Microencapsulated fuel

    International Nuclear Information System (INIS)

    Lee, H. G.; Kim, D. J.; Park, J. Y.; Kim, W. J.; Lee, S. J.

    2015-01-01

    FCM nuclear fuel is composed of tristructural isotropic(TRISO) fuel particle and SiC ceramic matrix. SiC ceramic matrix play an essential part in protecting fission product. In the FCM fuel concept, fission product is doubly protected by TRISO coating layer and SiC ceramic matrix in comparison with the current commercial UO2 fuel system of LWR. In addition to a safety enhancement of FCM fuel, thermal conductivity of SiC ceramic matrix is better than that of UO2 fuel. Because the centerline temperature of FCM fuel is lower than that of the current UO2 fuel due to the difference of thermal conductivity of fuel, an operational release of fission products from the fuel can be reduced. SiC ceramic has attracted for nuclear fuel application due to its high thermal conductivity properties with good radiation tolerant properties, a low neutron absorption cross-section and a high corrosion resistance. Thermal conductivity of ceramic matrix composite depends on the thermal conductivity of each component and the morphology of reinforcement materials such as fibers and particles. There are many results about thermal conductivity of fiber-reinforced composite like as SiCf/SiC composite. Thermal conductivity of SiC ceramics and FCM pellets with the volume fraction of TRISO particles were measured and analyzed by analytical models. Polycrystalline SiC ceramics and FCM pellets with TRISO particles were fabricated by hot press sintering with sintering additives. Thermal conductivity of the FCM pellets with TRISO particles of 0 vol.%, 10 vol.%, 20 vol.%, 30 vol.% and 40 vol.% show 68.4, 52.3, 46.8, 43.0 and 34.5 W/mK, respectively. As the volume fraction of TRISO particles increased, the measured thermal conductivity values closely followed the prediction of Maxwell's equation

  7. Stress Corrosion Cracking of Pipeline Steels in Fuel Grade Ethanol and Blends - Study to Evaluate Alternate Standard Tests and Phenomenological Understanding of SCC

    Science.gov (United States)

    2011-10-30

    Main aim of this project was to evaluate alternate standard test methods for stress corrosion cracking (SCC) and compare them with the results from slow strain rate test (SSRT) results under equivalent environmental conditions. Other important aim of...

  8. Fuel assembly

    International Nuclear Information System (INIS)

    Chaki, Masao; Nishida, Koji; Karasawa, Hidetoshi; Kanazawa, Toru; Orii, Akihito; Nagayoshi, Takuji; Kashiwai, Shin-ichi; Masuhara, Yasuhiro

    1998-01-01

    The present invention concerns a fuel assembly, for a BWR type nuclear reactor, comprising fuel rods in 9 x 9 matrix. The inner width of the channel box is about 132mm and the length of the fuel rods which are not short fuel rods is about 4m. Two water rods having a circular cross section are arranged on a diagonal line in a portion of 3 x 3 matrix at the center of the fuel assembly, and two fuel rods are disposed at vacant spaces, and the number of fuel rods is 74. Eight fuel rods are determined as short fuel rods among 74 fuel rods. Assuming the fuel inventory in the short fuel rod as X(kg), and the fuel inventory in the fuel rods other than the short fuel rods as Y(kg), X and Y satisfy the relation: X + Y ≥ 173m, Y ≤ - 9.7X + 292, Y ≤ - 0.3X + 203 and X > 0. Then, even when the short fuel rods are used, the fuel inventory is increased and fuel economy can be improved. (I.N.)

  9. Calculus of radiolytic products generation in water due to alpha radiation. Determination of the spent nuclear fuels matrix alteration rate Determination of velocity of spent fuel matrix; Calculo de la generacion de productos radioliticos en agua por radiacion {alpha}. Determinacion de la velocidad de alteracion de la matriz del combustible nuclear gastado

    Energy Technology Data Exchange (ETDEWEB)

    Quinones, J.; Serrano, J.; Diaz Arocas, P.; Rodriguez Almazan, J. L. [Ciemat. Madrid (Spain); Bruno, J.; Cera, E.; Merino, J.; Esteban, J. A.; Martinez-Esparza, A. [Enresa. Madrid (Spain)

    2000-07-01

    The generation of radiolytic products as a result of alpha radiation in the surface of the spent fuel is a key process in order to understand how the it becomes degraded in repository conditions. The present work has established a radiolytic model based on a set of reactions involving fuel oxidation-dissolution and radiolytic products recombination. It also includes the decrease of the dose rates as the main alpha emitters decay away. Four cases, with varying parameters of the system, have been assessed. The results show a decrease in both the concentration of the radiolytic products in the gap water and the degradation of the fuel matrix. It has been estimated that in the period of the evaluation (10''6 years) up to 52% of the pellet is altered in the conservative cases, whereas only 11% is altered in the realistic cases. No significant differences were observed when the carbonates reactions were included in the system. (Author)

  10. Chemical Engineering Division Fuel Cycle Programs. Quarterly progress report, April-June 1978. [Advanced solvent extraction; accidents; pyrochemical; radwaste in metal matrix; waste migration

    Energy Technology Data Exchange (ETDEWEB)

    Steindler, M. J.; Ader, M.; Barletta, R. E.

    1979-12-01

    Fuel cycle studies reported include development of centrifugal contactors for Purex processes. Tricaprylmethyl-ammonium nitrate and di-n-amyl-n-amylphosphonate are being evaluated as Thorex extractants. Dispersion of uranium and plutonium by fires, and mechanisms for subdividing and dispersing liquids and solids were reviewed. In the pyrochemical and dry processing program, a facility for testing containment materials is under construction; a flowsheet for carbide fuel processing has been designed and studies of carbide reactions in bismuth are underway; salt transport processes are being studied; process-size refractory metal vessels are being fabricated; the feasibility of AIROX reprocessing is being determined; the solubility of UO/sub 2/, UO/sub 2/ + fission products, and PuO/sub 2/ in molten alkali metal nitrates, has been investigated; a flowsheet was developed for reprocessing actinide oxides in molten salts; preparation of Th-U carbide from the oxide is being studied; new flowsheets based on the Dow Aluminum Pyrometallurgical process for reprocessing of spent uranium metal fuel have been prepared; the chloride volitility processing of thorium-based fuels is being studied; the reprocessing of (Th,U)O/sub 2/ solid solution in KCl-LiCl-ThCl/sub 4/-Th is being studied; and a flowsheet for processing spent nuclear fuel in molten tin has been constructed. Leach rates of simulated encapsulated waste forms in a metal matrix were studied. Nine criteria for handling waste cladding hulls were established. Strontium and tin migration in glauconite columns was measured. Radioactive Sr in a stream of water moved through oolitic limestone as rapidly as water, but in a stream of water equilibrated with the limestone, Sr moved through the limestone one-tenth as fast. Migration of trace quantities of Cs and I through kaolinite was studied. 88 figures, 53 tables.

  11. Oxidation kinetics of simulated metallic spent fuel in air at 200∼300 .deg. C

    International Nuclear Information System (INIS)

    Joo, J. S.; Yoo, K. S.; Jo, I. J.; Kook, D. H.; Lee, E. P.; Lee, J. C.; Bang, K. S.; Kim, H. D.

    2003-01-01

    In order to evaluate the long term storage safety study of the metallic spent fuel, U-5Zr, U-5Ti, U-5Ni, U-5Nb, and U-5Hf simulated metallic uranium alloys, known as corrosion resistant alloys, were fabricated and oxidized in oxygen gas at 200 .deg. C ∼ 300 .deg. C. All simulated metallic uranium alloys were more corrosion resistant than pure uranium metal, and corrosion resistance increases Nb, Ni, Ti, Zr, Hf in that order. The oxidation rates of uranium alloys determined and activation energy was calculated for each alloy. The matrix microstructure of the test specimens were analyzed using OM, SEM, and EPMA. It was concluded that Nb was the best acceptable alloying elements for reducing corrosion of uranium metal, and Ni, Ti were also considered to suitable as candidate

  12. Extended life aluminide fuel. Final report

    International Nuclear Information System (INIS)

    Miller, L.G.; Beeston, J.M.

    1986-06-01

    As the price of fuel fabrication, shipment of both new and spent fuel, and fuel reprocessing continue to rise at a rapid rate, researchers look for alternate methods to keep reactor fuel costs within their limited funding. Extended fuel element lifetimes, without jeopardizing reactor safety, can reduce fuel costs by up to a factor of two. The Extended Life Aluminide (ELAF) program was started at the Idaho National Engineering Laboratory (INEL) as a joint project of the United States Department of Energy (DOE), the University of Missouri, and the Massachusetts Institute of Technology research reactors. Fuel plates of Advanced Test Reactor (ATR) type construction were fabricated at Atomics International and irradiated in the ATR at the INEL. Four fuel matrix compositions were tested (i.e., 50 vol% UAl/sub x/ cores for reference, and 40, 45 and 50 vol% UAl 2 cores). The 50 vol% UAl 2 cores contained up to 3 grams U-235 per cm 3 of core. Three plates of each composition were irradiated to peak burnup levels of 3 x 10 21 fission/cm 3 of core. The only observed damage was due to external corrosion at similar rates experienced by UAl/sub x/ fuel elements in test reactors

  13. Corrosion performance of Al-Si-Cu hypereutectic alloys in a synthetic condensed automotive solution

    Directory of Open Access Journals (Sweden)

    Hamilta de Oliveira Santos

    2005-06-01

    Full Text Available In this investigation the corrosion resistance of four Al-Si hypereutectic alloys in a solution typical of condensate from automotive fuel combustion products, and referred to here as synthetic condensed automotive solution, has been studied. Three commercial alloys that are used for cylinder liners, and a laboratory made alloy, were studied by electrochemical impedance spectroscopy and measurements were taken after increasing times of immersion in this solution. Comparison of the electrochemical response of the four alloys in the corrosive solution was carried out. Although the mechanisms by which the four alloys corroded were similar, the results indicated differences in corrosion resistances of these alloys, and these differences could be related to their microstructures. The laboratory prepared alloy showed increased susceptibility to pitting corrosion compared to the commercial alloys. The surfaces of the alloys were examined, before and after the corrosion test, by scanning electron microscopy and analyzed by energy dispersive spectroscopy. The results indicated preferential attack of the aluminium matrix phase in all the alloys. The alloy with higher copper content and prepared by spray forming was more susceptible to pitting compared to the other alloys. The EIS response at low frequencies indicated a diffusion-controlled process, probably that of oxygen to the alloy interface.

  14. Influence of the silicon content on the core corrosion properties of dispersion type fuel plates; Influencia del Contenido en silicio sobre la corrosion acuosa de los nucleos de placas combustibles

    Energy Technology Data Exchange (ETDEWEB)

    Calvo, C.; Saenz de Tejada, L. M.; Diaz Diaz, J.

    1969-07-01

    A new process to produce aluminium base dispersion type fuel plates has been developed at the Spanish JEN (Junta de Energia Nuclear). The dispersed fuel material is obtained by an aluminothermic process to render a stoichiometric cermet of UAI{sub 3} and AI{sub 2}O{sub 3} according to the reaction. (Author)

  15. Future and benefits of corrosion research

    International Nuclear Information System (INIS)

    Staehle, Roger W.

    2002-01-01

    The subject of corrosion is a design science. The subject of stress analysis is a design science as is the subject of heat transfer. When the subject of corrosion is considered in the framework design a clear framework of the priorities and objectives becomes apparent. Further, corrosion becomes a more explicit and important subject in the overall design, manufacturing, and operation phases of equipment: in this framework, the funding and support of corrosion work is necessary to the designers and users of equipment. The subject of corrosion is usually less important in the early stages of operation of equipment: in these early stages, the subjects. Corrosion becomes important to the longer term reliability and safety of equipment. Corrosion is often a principal determiner of design life. Corrosion is often more important after the manufacturing warranty is expired: therefore the subject is often more important to the user than to the manufacturer. In order that the subject of corrosion is considered and incorporated in the design as well as in user specifications, there must be a language and means of easily understood communication between the design-operation community and the corrosion community. For example, the designers do not understand the language of 'pitting potential': rather, they understand design life and permissible stress. Thus, corrosion must be put into terms that can be understood and utilized by designers and operators. Two methodologies have been developed for communicating effectively between the corrosion and the design communities: these are the 'Corrosion Based Design Approach' and the 'Location for Analysis Matrix.' These provide simple check off lists to designers for asking questions and assuring that credible answers have been obtained on issues that affect reliable and economic performance. Both of these subject are discussed in this presentation. The future of corrosion research is its effective linkage with design and operation of

  16. Evaluation of corrosion attack of chimney liners

    Directory of Open Access Journals (Sweden)

    Blahetová M.

    2016-06-01

    Full Text Available The case study of chimney liner corrosion addresses three specific cases of damage of chimney systems from of stainless steels. These systems were used for flue of gas arising from the combustion of brown coal in small automatic boilers, which are used for heating. Detailed analyzes implied that the cause of devastating corrosion of the steel AISI 316 and 304 steel (CSN 17349, 17241 was particularly high content of halides (chlorides and fluorides, which caused a severe pitting corrosion, which led up to the perforation of the liner material. Simultaneous reduction of the thickness of the used sheets was due to by the general corrosion, which was caused by the sulfur in the solid fuel. The condensation then led to acid environment and therefore the corrosion below the dew point of the sulfuric acid has occurred. All is documented by metallographic analysis and microanalysis of the corrosion products.

  17. Formation and Role of Gel Fractions in the Corrosion Layer of Zirconium Cladding as the First-stage Protection of the Nuclear Power Plant Fuel

    Czech Academy of Sciences Publication Activity Database

    Weishauptová, Zuzana; Vrtílková, V.; Bláhová, O.; Maixner, J.

    -, č. 16 (2007), s. 29-38 ISSN 1214-9691 R&D Projects: GA ČR GA106/04/0043 Institutional research plan: CEZ:AV0Z30460519 Keywords : zirconium alloys * corrosion layer * hydrated ZrO2 Subject RIV: CF - Physical ; Theoretical Chemistry

  18. The effect of mixing methods and polymer infiltration and pyrolysis (PIP) cycles on the densification of silicon carbide inert matrix fuel through a polymer precursor route

    International Nuclear Information System (INIS)

    Shih Chunghao; Tulenko, J.S.; Baney, R.H.

    2011-01-01

    Highlights: ► A pre-ceramic polymer precursor route was utilized to fabricate SiC based inert matrix fuel (IMF). ► The effect of mixing and polymer infiltration and pyrolysis (PIP) cycles on the fabrication of the SiC IMF was investigated. ► A theoretical density of ∼86% has been achieved after one PIP cycle. ► Pore size distributions before and after PIP cycles have been determined. ► SEM images showed a “springback” effect for 600 MPa pressed pellets. - Abstract: The effect of mixing methods on the fabrication of silicon carbide (SiC) inert matrix fuel through a polymer precursor route was investigated in order to break up the agglomerates of the SiC particles observed in earlier studies. It was found that a high energy shaker mill could effectively break up the agglomerates and thereby achieve a higher pellet density. Moreover, it was found that the pellet density depended less on the pressing pressure, when the particles are well mixed. SEM images showed cracks caused by the springback effect on pellets with a high cold pressing pressure of 600 MPa, but no signs of springback effect were observed for the 200 MPa pressed pellets. The polymer infiltration and pyrolysis (PIP) cycles were used to further increase the pellet density and close the open pores. The first PIP cycle was found to increase the theoretical density of the pellets from 81.2% to 86.0% and close ∼50% of the open pores. The pore size distribution showed that most of the remaining open pores had diameters smaller than 10 nm. The successive second PIP cycle is not as effective on either the density or the pore size distribution because the small pores present before the second PIP cycle made the infiltration process difficult.

  19. Corrosion inhibitors

    International Nuclear Information System (INIS)

    El Ashry, El Sayed H.; El Nemr, Ahmed; Esawy, Sami A.; Ragab, Safaa

    2006-01-01

    The corrosion inhibition efficiencies of some triazole, oxadiazole and thiadiazole derivatives for steel in presence of acidic medium have been studied by using AM1, PM3, MINDO/3 and MNDO semi-empirical SCF molecular orbital methods. Geometric structures, total negative charge on the molecule (TNC), highest occupied molecular energy level (E HOMO ), lowest unoccupied molecular energy level (E LUMO ), core-core repulsion (CCR), dipole moment (μ) and linear solvation energy terms, molecular volume (V i ) and dipolar-polarization (π *), were correlated to corrosion inhibition efficiency. Four equations were proposed to calculate corrosion inhibition efficiency. The agreement with the experimental data was found to be satisfactory; the standard deviations between the calculated and experimental results ranged between ±0.03 and ±4.18. The inhibition efficiency was closely related to orbital energies (E HOMO and E LUMO ) and μ. The correlation between quantum parameters and experimental inhibition efficiency has been validated by single point calculations for the semi-empirical AM1 structures using B3LYP/6-31G** as a higher level of theory. The proposed equations were applied to predict the corrosion inhibition efficiency of some related structures to select molecules of possible activity from a presumable library of compounds

  20. Underground pipeline corrosion

    CERN Document Server

    Orazem, Mark

    2014-01-01

    Underground pipelines transporting liquid petroleum products and natural gas are critical components of civil infrastructure, making corrosion prevention an essential part of asset-protection strategy. Underground Pipeline Corrosion provides a basic understanding of the problems associated with corrosion detection and mitigation, and of the state of the art in corrosion prevention. The topics covered in part one include: basic principles for corrosion in underground pipelines, AC-induced corrosion of underground pipelines, significance of corrosion in onshore oil and gas pipelines, n

  1. Simultaneous wastewater treatment and bioelectricity production in microbial fuel cells using cross-linked chitosan-graphene oxide mixed-matrix membranes.

    Science.gov (United States)

    Holder, Shima L; Lee, Ching-Hwa; Popuri, Srinivasa R

    2017-05-01

    Microbial fuel cells (MFCs) are emerging technology for wastewater treatment by chemical oxygen demand (COD) reduction and simultaneous bioelectricity production. Fabrication of an effective proton exchange membrane (PEM) is a vital component for MFC performance. In this work, green chitosan-based (CS) PEMs were fabricated with graphene oxide (GO) as filler material (CS-GO) and cross-linked with phosphoric acid (CS-GO-P(24)) or sulfuric acid (CS-GO-S(24)) to determine their effect on PEM properties. Interrogation of the physicochemical, thermal, and mechanical properties of the cross-linked CS-GO PEMs demonstrated that ionic cross-linking based on the incorporation of PO 4 3- groups in the CS-GO mixed-matrix composites, when compared with sulfuric acid cross-linking commonly used in proton exchange membrane fuel cell (PEMFC) studies, generated additional density of ionic cluster domains, rendered enhanced sorption properties, and augmented the thermal and mechanical stability of the composite structure. Consequently, bioelectricity performance analysis in MFC application showed that CS-GO-P(24) membrane produced 135% higher power density than the CS-GO-S(24) MFC system. Simultaneously, 89.52% COD removal of primary clarifier municipal wastewater was achieved in the MFC operated with the CS-GO-P(24) membrane.

  2. Modelling of zirconium alloys corrosion in LWRs

    International Nuclear Information System (INIS)

    Kritskij, V.G.; Berezina, I.G.; Kritskij, A.V.; Stjagkin, P.S.

    1999-01-01

    Chemical parameters, that exerted effect on Zr+1%Nb alloy corrosion and deserved consideration during reactor operation, were defined and a model was developed to describe the influence of physical and chemical parameters on zirconium alloys corrosion in nuclear power plants. The model is based on the correlation between the zirconium oxide solubility in high-temperature water under the influence of the chemical parameters and the measured values of fuel cladding corrosion under LWR conditions. The intensity of fuel cladding corrosion in the primary circuits depends on the coolant water quality, growth of iron oxide deposits and vaporization portion. Mathematically, the oxidation rate can be expressed as a sum of heat and radiation components. The temperature dependence on the oxidation rate can be described by the Arrenius equation. The radiation component of Zr uniform corrosion equation is a function of several factors such as neutron fluency, the temperature the metallurgical composition and et. We assume that the main factor is the changing of water chemistry and the H 2 O 2 concentration play the determinative role. Probably, the influence of H 2 O 2 is based on the formation of unstable compound ZrO 3 ·nH 2 O and Zr(OH) 4 with high solubility. The validity of the used formulae was confirmed by corrosion measurements on WWER and RBMK fuel cladding. The model can be applied for calculating the reliability of nuclear fuel operation. (author)

  3. Corrosion of carbon steel in oxidizing caustic solutions

    International Nuclear Information System (INIS)

    Divine, J.R.; Bowen, W.M.

    1984-01-01

    A series of tests have been completed on a range of proposed waste compositions at temperatures up to 100 0 C. These tests have sought data on uniform corrosion, pitting, and stress corrosion cracking. No indication of the latter two types of corrosion was observed within the test matrix. Corrosion rates after four months were generally below 25μm/y. By the end of twelve months all results, except for very concentrated mixtures, were below 13 μm/y. Prediction equations were generated from a model fitted to the data. The equations provide a rapid means of estimating the corrosion rate for waste compositions and temperatures within the test limits

  4. Understanding the Risk of Chloride Induced Stress Corrosion Cracking of Interim Storage Containers for the Dry Storage of Spent Nuclear Fuel: Evolution of Brine Chemistry on the Container Surface

    International Nuclear Information System (INIS)

    Enos, David; Bryan, Charles R.

    2015-01-01

    Although the susceptibility of austenitic stainless steels to chloride-induced stress corrosion cracking is well known, uncertainties exist in terms of the environmental conditions that exist on the surface of the storage containers. While a diversity of salts is present in atmospheric aerosols, many of these are not stable when placed onto a heated surface. Given that the surface temperature of any container storing spent nuclear fuel will be well above ambient, it is likely that salts deposited on its surface may decompose or degas. To characterize this effect, relevant single and multi-salt mixtures are being evaluated as a function of temperature and relative humidity to establish the rates of degassing, as well as the likely final salt and brine chemistries that will remain on the canister surface.

  5. Understanding the Risk of Chloride Induced Stress Corrosion Cracking of Interim Storage Containers for the Dry Storage of Spent Nuclear Fuel: Evolution of Brine Chemistry on the Container Surface.

    Energy Technology Data Exchange (ETDEWEB)

    Enos, David; Bryan, Charles R.

    2015-10-01

    Although the susceptibility of austenitic stainless steels to chloride-induced stress corrosion cracking is well known, uncertainties exist in terms of the environmental conditions that exist on the surface of the storage containers. While a diversity of salts is present in atmospheric aerosols, many of these are not stable when placed onto a heated surface. Given that the surface temperature of any container storing spent nuclear fuel will be well above ambient, it is likely that salts deposited on its surface may decompose or degas. To characterize this effect, relevant single and multi-salt mixtures are being evaluated as a function of temperature and relative humidity to establish the rates of degassing, as well as the likely final salt and brine chemistries that will remain on the canister surface.

  6. NOVEL CORROSION SENSOR FOR VISION 21 SYSTEMS

    Energy Technology Data Exchange (ETDEWEB)

    Heng Ban

    2004-12-01

    Advanced sensor technology is identified as a key component for advanced power systems for future energy plants that would have virtually no environmental impact. This project intends to develop a novel high temperature corrosion sensor and subsequent measurement system for advanced power systems. Fireside corrosion is the metal loss caused by chemical reactions on surfaces exposed to the combustion environment. Such corrosion is the leading mechanism for boiler tube failures and has emerged to be a significant concern for current and future energy plants due to the introduction of technologies targeting emissions reduction, efficiency improvement, or fuel/oxidant flexibility. Corrosion damage can lead to catastrophic equipment failure, explosions, and forced outages. Proper management of corrosion requires real-time indication of corrosion rate. However, short-term, on-line corrosion monitoring systems for fireside corrosion remain a technical challenge to date due to the extremely harsh combustion environment. The overall objective of this proposed project is to develop a technology for on-line corrosion monitoring based on a new concept. This report describes the initial results from the first-year effort of the three-year study that include laboratory development and experiment, and pilot combustor testing.

  7. Corrosion mechanisms for metal alloy waste forms: experiment and theory Level 4 Milestone M4FT-14LA0804024 Fuel Cycle Research & Development

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Xiang-Yang [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Taylor, Christopher D. [The Ohio State Univ., Columbus, OH (United States). Fontana Corrosion Center; Kim, Eunja [Univ. of Nevada, Las Vegas, NV (United States); Goff, George Scott [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Kolman, David Gary [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2014-07-31

    This document meets Level 4 Milestone: Corrosion mechanisms for metal alloy waste forms - experiment and theory. A multiphysics model is introduces that will provide the framework for the quantitative prediction of corrosion rates of metallic waste forms incorporating the fission product Tc. The model requires a knowledge of the properties of not only the metallic waste form, but also the passive oxide films that will be generated on the waste form, and the chemistry of the metal/oxide and oxide/environment interfaces. in collaboration with experimental work, the focus of this work is on obtaining these properties from fundamental atomistic models. herein we describe the overall multiphysics model, which is based on MacDonald's point-defect model for passivity. We then present the results of detailed electronic-structure calculations for the determination of the compatibility and properties of Tc when incorporated into intermetallic oxide phases. This work is relevant to the formation of multi-component oxides on metal surfaces that will incorporate Tc, and provide a kinetic barrier to corrosion (i.e. the release of Tc to the environment). Atomistic models that build upon the electronic structure calculations are then described using the modified embedded atom method to simulate metallic dissolution, and Buckingham potentials to perform classical molecular dynamics and statics simulations of the technetium (and, later, iron-technetium) oxide phases. Electrochemical methods were then applied to provide some benchmark information of the corrosion and electrochemical properties of Technetium metal. The results indicate that published information on Tc passivity is not complete and that further investigation is warranted.

  8. Irradiation and lithium presence influence on the crystallographic nature of zirconia in the framework of PWR zircaloy 4 fuel cladding corrosion study; Influence de l'irradiation et de la presence du lithium sur la nature cristallographique de la zircone dans le cadre de l'etude de la corrosion du zircaloy 4 en milieu reacteur a eau pressurisee

    Energy Technology Data Exchange (ETDEWEB)

    Gibert, C

    1999-07-01

    The-increasing deterioration of the initially protective zirconia layer is one of the hypotheses which can explain the impairment with time of PWR fuel cladding corrosion. This deterioration could be worsened by irradiation or lithium presence in the oxidizing medium. The aim of this thesis was to underline the influence of those two parameters on zirconia crystallographic nature. We first studied the impact of ionic irradiation on pure, powdery, monoclinic zirconia and oxidation formed zirconia, mainly with X-ray diffraction and Raman microscopy. The high or low energy particles used (Kr{sup n+-}, Ar{sup n+}) respectively favored electronic or atomic defaults production. The crystallographic analyses showed that these irradiation have a significant effect on zirconia by inducing nucleation or growth of tetragonal phase. The extent depends on sample nature and particles energy. In all cases, phase transformation is correlated with crystalline parameters, grain size and especially micro-stress changes. The results are consistent with those obtained with 1 to 5 cycles PWR claddings. Therefore, the corrosion acceleration observed in reactor can partly be explained by the stress fields appearance under irradiation, which is particularly detrimental to zirconia layer cohesion. Last, we have underlined that the presence of considerable amounts of lithium in the oxidizing medium ((> 700 ppm) induces the disappearance of the tetragonal zirconia located at the metal/oxide interface and the appearance of a porosity of the dense under layer, which looses its protectiveness. (author)

  9. Neural network for prediction of superheater fireside corrosion

    Energy Technology Data Exchange (ETDEWEB)

    Makkonen, P. [Foster Wheeler Energia Oy, Karhula R and D Center, Karhula (Finland)

    1998-12-31

    Superheater corrosion causes vast annual losses to the power companies. If the corrosion could be reliably predicted, new power plants could be designed accordingly, and knowledge of fuel selection and determination of process conditions could be utilized to minimize superheater corrosion. If relations between inputs and the output are poorly known, conventional models depending on corrosion theories will fail. A prediction model based on a neural network is capable of learning from errors and improving its performance as the amount of data increases. The neural network developed during this study predicts superheater corrosion with 80 % accuracy at early stage of the project. (orig.) 10 refs.

  10. Estimating the lifetimes of titanium containers for nuclear fuel waste: A damage function for the crevice corrosion of grade-2 titanium

    International Nuclear Information System (INIS)

    Shoesmith, D.W.; Ikeda, B.M.; Quinn, M.J.; Kolar, M.

    1995-06-01

    The assumptions upon which the lifetime failure model used in the postclosure assessment is based are reevaluated. In particular, the conservations involved in assuming that crevice initiation would occur, and that sufficient oxygen would be present to maintain crevice propagation to failure, are discussed. Unless the period required to saturate the environment around the container can be specified with some certainty, it remains necessary to assume corrosion would initiate rapidly on all containers. A modified version of the container lifetime model has been developed which avoids the need to use averaged temperature profiles. In this model, these profiles are converted to propagation rate profiles using an experimental activation energy and then numerically integrated to predict container failure times. A damage function is developed relating the maximum depth of penetration by crevice corrosion to either the time since emplacement in the vault or the total amount of oxygen consumed. This function is used to estimate the maximum penetration depth expected if all the oxygen available in a borehole is consumed in crevice corrosion and to determine the impact on container lifetimes of various repassivation criteria. The factors likely to cause repassivation are summarized, and a number of engineering approaches to extending container lifetimes suggested. (author). 22 refs., 1 tab., 17 figs

  11. Separation of Pu and Nd from Uranium matrix by equilibrated cation exchanger for burnup measurement of irradiated nuclear fuel

    International Nuclear Information System (INIS)

    Joe, Kih Soo; Kim, Jung Suk; Jeon, Young Shin; Han, Sun Ho; Eom, Tae Yoon

    1993-01-01

    Ion chromatographic method has been applied for burnup measurement of irradiated nuclear fuel by dynamic system using 1-octanesulfonate as a cation exchanger and α-hydroxyisobutyric acid as an eluant. A number of elution techniques were evaluated for the optimum separation of plutonium, uranium and neodymium. These elements were individually separated and collected by gradient elution between 0.05 M and 0.40 M of α-hydroxyisobutyric acid in a single column, and finally determined by isotope dilution mass spectrometry. The burnup data from this method were compared with those from conventional anion exchange method. The results showed a good agreement within 3.5 % of difference between two methods. (Author)

  12. Bio-inspired Construction of Advanced Fuel Cell Cathode with Pt Anchored in Ordered Hybrid Polymer Matrix.

    Science.gov (United States)

    Xia, Zhangxun; Wang, Suli; Jiang, Luhua; Sun, Hai; Liu, Shuang; Fu, Xudong; Zhang, Bingsen; Sheng Su, Dang; Wang, Jianqiang; Sun, Gongquan

    2015-11-05

    The significant use of platinum for catalyzing the cathodic oxygen reduction reactions (ORRs) has hampered the widespread use of polymer electrolyte membrane fuel cells (PEMFCs). The construction of well-defined electrode architecture in nanoscale with enhanced utilization and catalytic performance of Pt might be a promising approach to address such barrier. Inspired by the highly efficient catalytic processes in enzymes with active centers embedded in charge transport pathways, here we demonstrate for the first time a design that allocates platinum nanoparticles (Pt NPs) at the boundaries with dual-functions of conducting both electrons by aid of polypyrrole and protons via Nafion(®) ionomer within hierarchical nanoarrays. By mimicking enzymes functionally, an impressive ORR activity and stability is achieved. Using this brand new electrode architecture as the cathode and the anode of a PEMFC, a high mass specific power density of 5.23 W mg(-1)Pt is achieved, with remarkable durability. These improvements are ascribed to not only the electron decoration and the anchoring effects from the Nafion(®) ionomer decorated PPy substrate to the supported Pt NPs, but also the fast charge and mass transport facilitated by the electron and proton pathways within the electrode architecture.

  13. Surface engineering for corrosion and wear resistance

    National Research Council Canada - National Science Library

    Davis, J. R

    2001-01-01

    ... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .16 Pitting . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .19 Crevice Corrosion . . . . . . . . ....

  14. Corrosion Characteristics of the SMART Materials

    Energy Technology Data Exchange (ETDEWEB)

    Baek, Jong Hyuk; Jeong, Y. H.; Choi, B. K.; Soh, J. R.; Lee, D. J.; Choi, B. S

    2000-05-01

    This report summarized the corrosion characteristics of the candidate steam generator tubes (PT-7M, ASTM Gr.2, Inconel-600), which are considering as the core materials in SMART. Also, this evaluated the waterchemstry conditions of commercial power plant including the PWR, BWR, WWER, PHWR, RBMK plants in comparison with that of SMART. And this report described that the microstructures of as-received PT-7M, ASTM Gr.2, and Inconel-600 as the candidate materials of fuel cladding and steam generator tubes and characterized the corrosion properties of the materials, which were tested systematically in the conditions of standard, ammonia solution and ammonia nodular to evaluate the corrosion resistance.

  15. Corrosion Characteristics of the SMART Materials

    International Nuclear Information System (INIS)

    Baek, Jong Hyuk; Jeong, Y. H.; Choi, B. K.; Soh, J. R.; Lee, D. J.; Choi, B. S.

    2000-05-01

    This report summarized the corrosion characteristics of the candidate steam generator tubes (PT-7M, ASTM Gr.2, Inconel-600), which are considering as the core materials in SMART. Also, this evaluated the waterchemstry conditions of commercial power plant including the PWR, BWR, WWER, PHWR, RBMK plants in comparison with that of SMART. And this report described that the microstructures of as-received PT-7M, ASTM Gr.2, and Inconel-600 as the candidate materials of fuel cladding and steam generator tubes and characterized the corrosion properties of the materials, which were tested systematically in the conditions of standard, ammonia solution and ammonia nodular to evaluate the corrosion resistance

  16. Corrosion Testing Facility

    Data.gov (United States)

    Federal Laboratory Consortium — The Corrosion Testing Facility is part of the Army Corrosion Office (ACO). It is a fully functional atmospheric exposure site, called the Corrosion Instrumented Test...

  17. Super ODS steels R and D for fuel cladding of next generation nuclear systems. 7) Corrosion behavior and mechanism in LBE

    International Nuclear Information System (INIS)

    Sano, H.; Fujisawa, T.; Kimura, A.; Inoue, Masaki; Ukai, S.; Ohnuki, S.; Okuda, T.; Abe, F.

    2009-01-01

    Corrosion of structural materials is one of the serious problems when lead-bismuth eutectic alloy (LBE) is used as a coolant material in next generation nuclear systems. In this study, dissolution experiments of synthetic Fe-Cr-Al alloys and developed super ODS steel candidates into LBE under several partial pressures of oxygen were conducted. Dissolution behaviors of major components in such steels into LBE were investigated. Interfacial behavior between LBE and steels was also observed. In addition, partial potential diagrams of the Fe-Cr-Al-O system at several conditions were established as basic data. From the potential diagrams, the partial pressure range of oxygen was estimated for the stable protective oxide layer formation at the interface. At lower oxygen partial pressure than the pressure that is enough for the formation of the stable oxide layer, a rough oxide layer was formed at the interface in all samples, and the alloy elements dissolved into LBE through it. On the other hand, at the oxygen partial pressure to form stable oxide layer, a dense and very thin oxide layer was formed especially on the higher aluminum content steel, preventing the alloy dissolution into LBE. From the results, aluminum and chromium content in steel were very important for preventing the corrosion by LBE. (author)

  18. Copper canisters for nuclear high level waste disposal. Corrosion aspects

    International Nuclear Information System (INIS)

    Werme, L.; Sellin, P.; Kjellbert, N.

    1992-10-01

    A corrosion analysis of a thick-walled copper canister for spent fuel disposal is discussed. The analysis has shown that there are no rapid mechanisms that may lead to canister failure, indicating an anticipated corrosion service life of several millions years. If further analysis of the copper canister is considered, it should be concentrated on identifying and evaluating processes other than corrosion, which may have a potential for leading to canister failure. (au)

  19. Corrosion resistance testing of high-boron-content stainless steels

    International Nuclear Information System (INIS)

    Petrman, I.; Safek, V.

    1994-01-01

    Boron steels, i.e. stainless steels with boron contents of 0.2 to 2.25 wt.%, are employed in nuclear engineering for the manufacture of baskets or wells in which radioactive fissile materials are stored, mostly spent nuclear fuel elements. The resistance of such steels to intergranular corrosion and uniform corrosion was examined in the Strauss solution and in boric acid; the dependence of the corrosion rate of the steels on their chemical composition was investigated, and their resistance was compared with that of AISI 304 type steel. Corrosion resistance tests in actual conditions of ''wet'' compact storage (demineralized water or a weak boric acid solution) gave evidence that boron steels undergo nearly no uniform corrosion and, as electrochemical measurements indicated, match standard corrosion-resistant steels. Corrosion resistance was confirmed to decrease slightly with increasing boron content and to increase somewhat with increasing molybdenum content. (Z.S.). 3 tabs., 4 figs., 7 refs

  20. Aqueous Corrosion Rates for Waste Package Materials

    Energy Technology Data Exchange (ETDEWEB)

    S. Arthur

    2004-10-08

    The purpose of this analysis, as directed by ''Technical Work Plan for: Regulatory Integration Modeling and Analysis of the Waste Form and Waste Package'' (BSC 2004 [DIRS 171583]), is to compile applicable corrosion data from the literature (journal articles, engineering documents, materials handbooks, or standards, and national laboratory reports), evaluate the quality of these data, and use these to perform statistical analyses and distributions for aqueous corrosion rates of waste package materials. The purpose of this report is not to describe the performance of engineered barriers for the TSPA-LA. Instead, the analysis provides simple statistics on aqueous corrosion rates of steels and alloys. These rates are limited by various aqueous parameters such as temperature (up to 100 C), water type (i.e., fresh versus saline), and pH. Corrosion data of materials at pH extremes (below 4 and above 9) are not included in this analysis, as materials commonly display different corrosion behaviors under these conditions. The exception is highly corrosion-resistant materials (Inconel Alloys) for which rate data from corrosion tests at a pH of approximately 3 were included. The waste package materials investigated are those from the long and short 5-DHLW waste packages, 2-MCO/2-DHLW waste package, and the 21-PWR commercial waste package. This analysis also contains rate data for some of the materials present inside the fuel canisters for the following fuel types: U-Mo (Fermi U-10%Mo), MOX (FFTF), Thorium Carbide and Th/U Carbide (Fort Saint Vrain [FSVR]), Th/U Oxide (Shippingport LWBR), U-metal (N Reactor), Intact U-Oxide (Shippingport PWR, Commercial), aluminum-based, and U-Zr-H (TRIGA). Analysis of corrosion rates for Alloy 22, spent nuclear fuel, defense high level waste (DHLW) glass, and Titanium Grade 7 can be found in other analysis or model reports.

  1. Effect of TiN particulate reinforcement on corrosive behaviour of ...

    Indian Academy of Sciences (India)

    Polarization studies indicate an increase in the corrosion resistance in composites compared to the matrix alloy. EIS study reveals that the polarization resistance (p) increases with increase in TiN content in composites, thus confirming improved corrosion resistance in composites. The observed decrease in corrosion rate ...

  2. Effect of TiN particulate reinforcement on corrosive behaviour of ...

    Indian Academy of Sciences (India)

    6061 alloy. It is understood that after the initiation of corrosion, interfacial corrosion products may have decoupled the conducting ceramic TiN from Al 6061 matrix alloy thus eliminating the galvanic effect between them. Keywords. Composites; corrosion; SEM; EDX; EIS. 1. Introduction. Particulate-reinforced aluminium metal ...

  3. Initial results from dissolution rate testing of N-Reactor spent fuel over a range of potential geologic repository aqueous conditions

    International Nuclear Information System (INIS)

    Gray, W.J.; Einziger, R.E.

    1998-04-01

    Hanford N-Reactor spent nuclear fuel (HSNF) may ultimately be placed in a geologic repository for permanent disposal. To determine whether the engineered barrier system that will be designed for emplacement of light-water-reactor (LWR) spent fuel will also suffice for HSNF, aqueous dissolution rate measurements were conducted on the HSNF. The purpose of these tests was to determine whether HSNF dissolves faster or slower than LWR spent fuel under some limited repository-relevant water chemistry conditions. The tests were conducted using a flowthrough method that allows the dissolution rate of the uranium matrix to be measured without interference by secondary precipitation reactions that would confuse interpretation of the results. Similar tests had been conducted earlier with LWR spent fuel, thereby allowing direct comparisons. Two distinct corrosion modes were observed during the course of these 12 tests. The first, Stage 1, involved no visible corrosion of the test specimen and produced no undissolved corrosion products. The second, Stage 2, resulted in both visible corrosion of the test specimen and left behind undissolved corrosion products. During Stage 1, the rate of dissolution could be readily determined because the dissolved uranium and associated fission products remained in solution where they could be quantitatively analyzed. The measured rates were much faster than has been observed for LWR spent fuel under all conditions tested to date when normalized to the exposed test specimen surface areas. Application of these results to repository conditions, however, requires some comparison of the physical conditions of the different fuels. The surface area of LWR fuel that could potentially be exposed to repository groundwater is estimated to be approximately 100 times greater than HSNF. Therefore, when compared on the basis of mass, which is more relevant to repository conditions, the HSNF and LWR spent fuel dissolve at similar rates

  4. Nuclear fuels

    International Nuclear Information System (INIS)

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

    2009-01-01

    , Bubbles and precipitates, Modeling fuel behavior); Modeling defects and fission products in UO 2 ceramic by ab initio computation (Ab initio computation, Point defects in uranium dioxide, Fission products in uranium dioxide, The indispensable coupling of modeling and experiment); Cladding and assembly materials (What is the purpose of cladding?, Zirconium alloys, Claddings: required to exhibit good mechanical strength, Mechanical behavior of irradiated Zr alloys, Claddings: required to prove corrosion resistant); Pellet-cladding interaction (The phenomena involved in pellet-cladding interaction (PCI), Experimental simulation of PCI and the lessons to be drawn from it, The requirement for an experimental basis, Numerical simulation of PCI, Towards a lifting of PCI-related operating constraints); Advanced UO 2 and MOX ceramics (Chromium oxide-doped UO 2 fuel, Novel MOX microstructures); Mechanical behavior of fuel assemblies (Assembly mechanical behavior in normal operating conditions, Assembly mechanical behavior in accident situations, Fuel in a loss of primary coolant accident (LOCA)); Introduction to LOCA-type accident transients (Overview of thermal-hydraulic and fuel-related aspects, Incidence of LOCA transients on the thermal-metallurgical-mechanical behavior of zirconium-base alloy cladding); Fuel in a reactivity insertion accident (RIA) (Safety criteria); Fuel in a severe accident (The VERCORS analytical program, The Phebus-FP global tests, Control of severe accidents in the EPR reactor); In-core fuel management (Relationships between cycle length, maximum burnup, and batch fraction Enrichment and burnable poisons, The impact of the nature of the fuel used, and its evolution, on the major parameters of core physics, and management Prospects for future trends in core management); Fuel cycle material balances (In-core evolution of materials, Decay heat and potential radiotoxicity, Plutonium management); Long-term behavior of spent fuel (The nature of spent nuclear

  5. Corrosion particles in the primary coolant of VVER-440 reactors

    Energy Technology Data Exchange (ETDEWEB)

    Vajda, N.; Molnar, Z., E-mail: vajdanor@gmail.com [RadAnal Ltd., Budapest (Hungary); Macsik, Z.; Szeles, E.; Hargittai, P., E-mail: szeles@iki.kfki.hu [Inst. of Isotopes of the Hungarian Academy of Sciences, Budapest (Hungary); Csordas, A.; Pinter, T., E-mail: csordas@aeki.kfki.hu [Atomic Energy Research Inst. of the Hungarian Academy of Sciences, Budapest (Hungary); Pinter, T., E-mail: pinter@npp.hu [Paks Nuclear Power Plant, Paks (Hungary)

    2010-07-01

    Corrosion and activity build-up processes are of major concern in ageing and life-extension of nuclear power reactors. Researches to study the migration of radioactive corrosion particles have been initiated at Paks Nuclear Power Plant (NPP), Hungary in order to better understand the corrosion of the primary circuit surfaces, the transport and activation of the particles of corrosion origin and their deposition on in-core and out-of-core surfaces. Radioactive corrosion particles were collected from the primary coolant and the steam generator surfaces of the 4 reactor units and subjected to detailed microanalytical and radioanalytical investigations. Scanning electron microscopy and energy dispersive X-ray microanalysis (SEM-EDX) were used to study the morphology and the composition of the matrix elements in the particles and the deposited corrosion layers. Particles identified by SEM-EDX were re-located under optical microscope by means of a coordinate transformation algorithm and were separated with a micromanipulator for further studies. Activities of γ emitting radionuclides were determined by high resolution γ spectrometry, and those of β decaying isotopes were measured by liquid scintillation (LS) spectrometry after radiochemical processing. High sensitivity of the nuclear measuring techniques allowed us to determine the low activity concentrations of the long-lived radionuclides, i.e. {sup 60}Co, {sup 54}Mn, {sup 63}Ni, {sup 55}Fe in the individual particles. Finally, high resolution sector-field inductively coupled plasma mass spectrometry (SF-ICP-MS) was applied to determine the ultralow concentrations of Co, Fe, Ni in the same particles. Specific activities of {sup 60}Co/Co, {sup 54}Mn/Fe, {sup 55}Fe/Fe and {sup 63}Ni/Ni were derived from the measured activity and concentration data. Specific activities of the radioactive corrosion products reveal the history of activity buildup processes in the particle. Typically, Fe-Cr-Ni oxide particles formed as a

  6. Corrosion particles in the primary coolant of VVER-440 reactors

    International Nuclear Information System (INIS)

    Vajda, N.; Molnar, Z.; Macsik, Z.; Szeles, E.; Hargittai, P.; Csordas, A.; Pinter, T.; Pinter, T.

    2010-01-01

    Corrosion and activity build-up processes are of major concern in ageing and life-extension of nuclear power reactors. Researches to study the migration of radioactive corrosion particles have been initiated at Paks Nuclear Power Plant (NPP), Hungary in order to better understand the corrosion of the primary circuit surfaces, the transport and activation of the particles of corrosion origin and their deposition on in-core and out-of-core surfaces. Radioactive corrosion particles were collected from the primary coolant and the steam generator surfaces of the 4 reactor units and subjected to detailed microanalytical and radioanalytical investigations. Scanning electron microscopy and energy dispersive X-ray microanalysis (SEM-EDX) were used to study the morphology and the composition of the matrix elements in the particles and the deposited corrosion layers. Particles identified by SEM-EDX were re-located under optical microscope by means of a coordinate transformation algorithm and were separated with a micromanipulator for further studies. Activities of γ emitting radionuclides were determined by high resolution γ spectrometry, and those of β decaying isotopes were measured by liquid scintillation (LS) spectrometry after radiochemical processing. High sensitivity of the nuclear measuring techniques allowed us to determine the low activity concentrations of the long-lived radionuclides, i.e. 60 Co, 54 Mn, 63 Ni, 55 Fe in the individual particles. Finally, high resolution sector-field inductively coupled plasma mass spectrometry (SF-ICP-MS) was applied to determine the ultralow concentrations of Co, Fe, Ni in the same particles. Specific activities of 60 Co/Co, 54 Mn/Fe, 55 Fe/Fe and 63 Ni/Ni were derived from the measured activity and concentration data. Specific activities of the radioactive corrosion products reveal the history of activity buildup processes in the particle. Typically, Fe-Cr-Ni oxide particles formed as a result of corrosion of the steel

  7. Corrosion of UN in water

    International Nuclear Information System (INIS)

    Sunder, S.; Miller, N.H.

    1996-10-01

    Corrosion of UN in water was investigated as a function of pH and temperature using X-ray photoelectron spectroscopy (XPS), X-ray diffraction (XRD), and by measuring the amount of ammonia formed due to its corrosion. The XPS results indicate that a freshly fractured surface of UN is quickly converted to U0 2 on exposure to liquid water or water vapours at ambient temperatures. These results show that UN is unstable in contact with water. The corrosion rate of UN is estimated to be ≥40 μmol·m -2 ·h -1 in deaerated water at ∼92 o C. There was no significant difference in corrosion rates measured in water at initial pHs of ∼6 and ∼10.3. These results contradict the literature reports stating that UN is stable in contact with boiling water. The implications of these results on the suitability of UN as a nuclear fuel for reactors are discussed. (author)

  8. Steam generator corrosion 2007; Dampferzeugerkorrosion 2007

    Energy Technology Data Exchange (ETDEWEB)

    Born, M. (ed.)

    2007-07-01

    Eiteneuer); (o) Thermodynamical calculations according to the optimization of spray alloys (Manfred Born, Juergen Korb, Ruediger W. Schuelen); (p) Thermal spraying between claim of erosion and claim of corrosion (Reinhard Polak, Franz Kremsner); (q) Flue gas-lateral cleaning of heating surfaces at operation and standstill of incineration boilers for waste, substitute fuels and biomass (Joerg Krueger); (r) Avoidance of superheater corrosion by external overheating (Reinhard Schu).

  9. Device of capturing for radioactive corrosion products

    International Nuclear Information System (INIS)

    Ohara, Atsushi; Fukushima, Kimichika.

    1984-01-01

    Purpose: To increase the area of contact between the capturing materials for the radioactive corrosion products contained in the coolants and the coolants by producing stirred turbulent flows in the coolant flow channel of LMFBR type reactors. Constitution: Constituent materials for the nuclear fuel elements or the reactor core structures are activated under the neutron irradiation, corroded and transferred into the coolants. While capturing devices made of pure metal nickel are used for the elimination of the corrosion products, since the coolants form laminar flows due to the viscosity thereof near the surface of the capturing materials, the probability that the corrosion products in the coolants flowing through the middle portion of the channel contact the capturing materials is reduced. In this invention, rotating rolls and flow channels in which the balls are rotated are disposed at the upstream of the capturing device to forcively disturb the flow of the liquid sodium, whereby the radioactive corrosion products can effectively be captured. (Kamimura, M.)

  10. Quantitative analysis of hydrogen gas formed by aqueous corrosion of metallic uranium

    International Nuclear Information System (INIS)

    Fonnesbeck, J.

    2000-01-01

    Three unirradiated EBR-II blanket fuel samples containing depleted uranium metal were corrosion tested in simulated J-13 well water at 90 C. The corrosion rate of the blanket uranium metal was then determined relative to H 2 formation. Corrosion of one of the samples was interrupted prior to complete oxidation of the uranium metal and the solid corrosion product was analyzed for UO 2 and UH 3

  11. Influence of irradiation and radiolysis on the corrosion rates and mechanisms of zirconium alloys

    International Nuclear Information System (INIS)

    Verlet, Romain

    2015-01-01

    The nuclear fuel of pressurized water reactors (PWR) in the form of uranium oxide UO 2 pellets (or MOX) is confined in a zirconium alloy cladding. This cladding is very important because it represents the first containment barrier against the release of fission products generated by the nuclear reaction to the external environment. Corrosion by the primary medium of zirconium alloys, particularly the Zircaloy-4, is one of the factors limiting the reactor residence time of the fuel rods (UO 2 pellets + cladding). To optimize core management and to extend the lifetime of the fuel rods in reactor, new alloys based on zirconium-niobium (M5) have been developed. However, the corrosion mechanisms of these are not completely understood because of the complexity of these materials, corrosion environment and the presence of radiation from the nuclear fuel. Therefore, this thesis specifically addresses the effects of radiolysis and defects induced by irradiation with ions in the matrix metal and the oxide layer on the corrosion rate of Zircaloy-4 and M5. The goal is to separate the influence of radiation damage to the metal, that relating to defects created in the oxide and that linked to radiolysis of the primary medium on the oxidation rate of zirconium alloys in reactor. 1) Regarding effect of irradiation of the metal on the oxidation rate: type dislocation loops appear and increase the oxidation rate of the two alloys. For M5, in addition to the first effect, a precipitation of fines needles of niobium reduced the solid solution of niobium concentration in the metal and ultimately in the oxide, which strongly reduces the oxidation rate of the alloy. 2) Regarding the effect of irradiation of the oxide layer on the oxidation rate: defects generated by the nuclear cascades in the oxide increase the oxidation rate of the two materials. For M5, germination of niobium enriched zones in irradiated oxide also causes a decrease of the niobium concentration in solid solution

  12. Corrosion performance of inorganic coatings in seawater

    NARCIS (Netherlands)

    Zhang, X.; Buter, S.J.; Ferrari, G.M.; Westing, E. van; Kowalski, L.

    2011-01-01

    Inorganic coatings are widely used to protect carbon steel hydraulic cylinder rods from wear and corrosion in aggressive offshore environment. Different types of lay-ers such as Ni/Cr, Al2O3, Cr2O3, TiO2, and Inconel 625 layers were applied to the carbon steels by plasma, High Velocity Oxygen Fuel

  13. Corrosion behaviors of ceramics against liquid sodium. Sodium corrosion characteristics of sintering additives

    International Nuclear Information System (INIS)

    Tachi, Yoshiaki; Kano, Shigeki; Hirakawa, Yasushi; Yoshida, Eiichi

    1998-01-01

    It has been progressed as the Frontier Materials Research to research and develop ceramics to apply for several components of fast breeder reactor using liquid sodium as coolant instead of metallic materials. Grain boundary of ceramics has peculiar properties compared with matrix because most of ceramics are produced by hardening and firing their raw powders. Some previous researchers indicated that ceramics were mainly corroded at grain boundaries by liquid sodium, and ceramics could not be used under corrosive environment. Thus, it is the most important for the usage of ceramics in liquid sodium to improve corrosion resistance of grain boundaries. In order to develop the advanced ceramics having good sodium corrosion resistance among fine ceramics, which have recently been progressed in quality and characteristics remarkably, sodium corrosion behaviors of typical sintering additives such as MgO, Y 2 O 3 and AlN etc. have been examined and evaluated. As a result, the followings have been clarified and some useful knowledge about developing advanced ceramics having good corrosion resistance against liquid sodium has been obtained. (1) Sodium corrosion behavior of MgO depended on Si content. Samples containing large amount of Si were corroded severely by liquid sodium, whereas others with low Si contents showed good corrosion resistance. (2) Both Y 2 O 3 and AlN, which contained little Si, showed good sodium corrosion resistance. (3) MgO, Y 2 O 3 and AlN are thought to be corroded by liquid sodium, if they contain some SiO 2 . Therefore, in order to improve sodium corrosion resistance, it is very important for these ceramics to prevent the contamination of matrix with SiO 2 through purity control of their raw powders. (author)

  14. Corrosion risk assessment and comprehensive evaluation of ship sea water pipe systems

    Directory of Open Access Journals (Sweden)

    YANG Guangfu

    2017-05-01

    Full Text Available In order to propose a comprehensive corrosion risk evaluation model based on an analysis of the corrosion risk of sea water pipe systems of ships in service,the modes of potential corrosion and their causes were first analyzed by identifying the origins of the inner and outer corrosion of the sea water pipe systems. In accordance with the matrixes of corrosion occurrence possibility and corrosion risk consequence,a corrosion risk matrix was established for a sea water pipe system. The method for the quantitative evaluation value of each mode of corrosion possibly occurring in a sea water pipe system was then presented. The comprehensive evaluation model of the corrosion was first built using the Analytical Hierarchy Process(AHP,which can be used for the comprehensive corrosion evaluation. The results show that the comprehensive evaluation value will be expressed as the corrosion risk level and corrosion-induced consequence of the sea water pipe system. This will be very useful for accurately protecting the sea water pipe systems of ships in service from corrosion,thereby attaining the goals of economy,rationality and timeliness.

  15. Microbiologically influenced corrosion in petroleum product pipelines--a review.

    Science.gov (United States)

    Muthukumar, N; Rajasekar, A; Ponmariappan, S; Mohanan, S; Maruthamuthu, S; Muralidharan, S; Subramanian, P; Palaniswamy, N; Raghavan, M

    2003-09-01

    Microbiologically influenced corrosion is responsible for most of the internal corrosion problems in oil transportation pipelines and storage tanks. One problematic area in treating gas lines is the occurrence of the stratification of water in the line. Under these conditions, corrosion inhibitors do not come into contact properly and oil and inhibitors undergo degradation. The role of bacteria on oil degradation, the consequences of oil degradation in fuel systems and its influence on corrosion have been explained in detail. Besides, factors influencing on degradation of oil and corrosion inhibitors have also been discussed. Mechanism of microbiologically influenced corrosion in oil pipeline has been explained. Many of the misapplication of biocides/inhibitors occur mainly because the characteristics of biocides/inhibitors are not considered before use in pipeline industry. List of biocides and monitoring programme have been collected from literature and presented.

  16. High temperature corrosion investigation in an oxyfuel combustion test rig

    DEFF Research Database (Denmark)

    Montgomery, Melanie; Bjurman, M.; Hjörnhede, A

    2014-01-01

    constructed by Brandenburg Technical University to gain understanding into oxyfuel firing. Two air‐cooled corrosion probes were exposed in this oxyfuel combustion chamber where the fuel was lignite. Gas composition was measured at the location of testing. Various alloys from a 2½ Cr steel, austenitic steels...... (perhaps carburized) zone was used as a measure of corrosion rates. The lowest alloyed steel had the highest corrosion rate, and the other austenitic and nickel alloys had much lower corrosion rates. Precipitates in the alloy adjacent the corrosion front were revealed for both Sanicro 28 and C‐276. However......Oxyfuel firing and subsequent capture of CO2 is a way to reduce CO2 emissions from coal‐fired boilers. Literature is summarized highlighting results which may contribute to understanding of the corrosion processes in an oxyfuel boiler.Tests were conducted in a 500 kWth oxyfuel test facility...

  17. Waterside corrosion of zirconium alloys in nuclear power plants

    International Nuclear Information System (INIS)

    Jeong, Yong Hwan; Baek, B. J.; Park, S. Y. and others

    1999-08-01

    The overview of corrosion and hydriding behaviors of Zr-based alloy under the conditions of the in-reactor service and in the absence of irradiation is introduced in this report. The metallurgical characteristics of Zr-based alloys and the thermo-mechanical treatments on the microstructures and the textures in the manufacturing process for fuel cladding are also introduced. The factors affecting the corrosion of Zr alloy in reactor are summarized. And the corrosion mechanism and hydrogen up-take are discussed based on the laboratory and in-reactor results. The phenomenological observations of zirconium alloy corrosion in reactors are summarized and the models of in-reactor corrosion are exclusively discussed. Finally, the effects of irradiation on the corrosion process in Zr alloy were investigated mainly based on the literature data. (author). 538 refs., 26 tabs., 105 figs

  18. Waterside corrosion of zirconium alloys in nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Yong Hwan; Baek, B.J.; Park, S.Y. [and others

    1999-08-01

    The overview of corrosion and hydriding behaviors of Zr-based alloy under the conditions of the in-reactor service and in the absence of irradiation is introduced in this report. The metallurgical characteristics of Zr-based alloys and the thermo-mechanical treatments on the microstructures and the textures in the manufacturing process for fuel cladding are also introduced. The factors affecting the corrosion of Zr alloy in reactor are summarized. And the corrosion mechanism and hydrogen up-take are discussed based on the laboratory and in-reactor results. The phenomenological observations of zirconium alloy corrosion in reactors are summarized and the models of in-reactor corrosion are exclusively discussed. Finally, the effects of irradiation on the corrosion process in Zr alloy were investigated mainly based on the literature data. (author). 538 refs., 26 tabs., 105 figs.

  19. Electrochemical corrosion of cermet coatings in artificial marine water

    International Nuclear Information System (INIS)

    Cabot, P.L.; Fernandez, J.; Guilemany, J.M.

    1998-01-01

    The electrochemical corrosion of different WC+12Co coatings sprayed on 34CrMo4 (UNS-G41350) steel by the high velocity oxygen fuel technique has been studied by corrosion potential and impedance measurements considering previous SEM observations and EDX microanalysis. The experiments were conducted in artificial marine water at 20 C and the impedance spectra were obtained at the corresponding corrosion potentials for the substrate, coating and substrate-coating systems. The impedance diagrams indicated that the electrochemical corrosion of the steel-coating systems is controlled by oxygen diffusion through a porous film of corrosion products, as in the case of the shot-blasted steel. In contrast, the corrosion of the coating appeared to be controlled by diffusion of oxygen through the electrolyte. The impedance diagrams obtained for the steel-coating systems depended on the porosities of the cermet coatings, thus being an useful procedure to characterize metals coated by cermets. (orig.)

  20. Long Term Corrosion/Degradation Test Six Year Results

    Energy Technology Data Exchange (ETDEWEB)

    M. K. Adler Flitton; C. W. Bishop; M. E. Delwiche; T. S. Yoder

    2004-09-01

    The Subsurface Disposal Area (SDA) of the Radioactive Waste Management Complex (RWMC) located at the Idaho National Engineering and Environmental Laboratory (INEEL) contains neutron-activated metals from non-fuel, nuclear reactor core components. The Long-Term Corrosion/Degradation (LTCD) Test is designed to obtain site-specific corrosion rates to support efforts to more accurately estimate the transfer of activated elements to the environment. The test is using two proven, industry-standard methods—direct corrosion testing using metal coupons, and monitored corrosion testing using electrical/resistance probes—to determine corrosion rates for various metal alloys generally representing the metals of interest buried at the SDA, including Type 304L stainless steel, Type 316L stainless steel, Inconel 718, Beryllium S200F, Aluminum 6061, Zircaloy-4, low-carbon steel, and Ferralium 255. In the direct testing, metal coupons are retrieved for corrosion evaluation after having been buried in SDA backfill soil and exposed to natural SDA environmental conditions for times ranging from one year to as many as 32 years, depending on research needs and funding availability. In the monitored testing, electrical/resistance probes buried in SDA backfill soil will provide corrosion data for the duration of the test or until the probes fail. This report provides an update describing the current status of the test and documents results to date. Data from the one-year and three-year results are also included, for comparison and evaluation of trends. In the six-year results, most metals being tested showed extremely low measurable rates of general corrosion. For Type 304L stainless steel, Type 316L stainless steel, Inconel 718, and Ferralium 255, corrosion rates fell in the range of “no reportable” to 0.0002 mils per year (MPY). Corrosion rates for Zircaloy-4 ranged from no measurable corrosion to 0.0001 MPY. These rates are two orders of magnitude lower than those specified in

  1. The development and testing of reduced enrichment fuels for Canadian research reactors

    International Nuclear Information System (INIS)

    Wood, J.C.; Foo, M.T.; Berthiaume, L.C.

    1983-01-01

    Fuel rods of uranium silicide dispersed in aluminum and clad in aluminum have been developed and tested in the laboratory and in-reactor. The properties of the dispersion fuel materials proved satisfactory with regard to thermal conductivity, aqueous corrosion resistance, strength and ductility, and thermal stability below 473 K. A vacancy condensation model is proposed to account for the thermally-induced swelling that occurs above 473 K by virtue of the chemical reactions that occur between the dispersed silicide fuel particles and the aluminum matrix. The in-reactor fuel core swelling was less than % after irradiation at high powers 76-131 kW/m) to a high terminal burnup (79.2 at% of U-235 atoms). (author)

  2. A multiphase interfacial model for the dissolution of spent nuclear fuel

    Science.gov (United States)

    Jerden, James L.; Frey, Kurt; Ebert, William

    2015-07-01

    The Fuel Matrix Dissolution Model (FMDM) is an electrochemical reaction/diffusion model for the dissolution of spent uranium oxide fuel. The model was developed to provide radionuclide source terms for use in performance assessment calculations for various types of geologic repositories. It is based on mixed potential theory and consists of a two-phase fuel surface made up of UO2 and a noble metal bearing fission product phase in contact with groundwater. The corrosion potential at the surface of the dissolving fuel is calculated by balancing cathodic and anodic reactions occurring at the solution interfaces with UO2 and NMP surfaces. Dissolved oxygen and hydrogen peroxide generated by radiolysis of the groundwater are the major oxidizing agents that promote fuel dissolution. Several reactions occurring on noble metal alloy surfaces are electrically coupled to the UO2 and can catalyze or inhibit oxidative dissolution of the fuel. The most important of these is the oxidation of hydrogen, which counteracts the effects of oxidants (primarily H2O2 and O2). Inclusion of this reaction greatly decreases the oxidation of U(IV) and slows fuel dissolution significantly. In addition to radiolytic hydrogen, large quantities of hydrogen can be produced by the anoxic corrosion of steel structures within and near the fuel waste package. The model accurately predicts key experimental trends seen in literature data, the most important being the dramatic depression of the fuel dissolution rate by the presence of dissolved hydrogen at even relatively low concentrations (e.g., less than 1 mM). This hydrogen effect counteracts oxidation reactions and can limit fuel degradation to chemical dissolution, which results in radionuclide source term values that are four or five orders of magnitude lower than when oxidative dissolution processes are operative. This paper presents the scientific basis of the model, the approach for modeling used fuel in a disposal system, and preliminary

  3. Neutronic fuel element fabrication

    Science.gov (United States)

    Korton, George

    2004-02-24

    This disclosure describes a method for metallurgically bonding a complete leak-tight enclosure to a matrix-type fuel element penetrated longitudinally by a multiplicity of coolant channels. Coolant tubes containing solid filler pins are disposed in the coolant channels. A leak-tight metal enclosure is then formed about the entire assembly of fuel matrix, coolant tubes and pins. The completely enclosed and sealed assembly is exposed to a high temperature and pressure gas environment to effect a metallurgical bond between all contacting surfaces therein. The ends of the assembly are then machined away to expose the pin ends which are chemically leached from the coolant tubes to leave the coolant tubes with internal coolant passageways. The invention described herein was made in the course of, or under, a contract with the U.S. Atomic Energy Commission. It relates generally to fuel elements for neutronic reactors and more particularly to a method for providing a leak-tight metal enclosure for a high-performance matrix-type fuel element penetrated longitudinally by a multiplicity of coolant tubes. The planned utilization of nuclear energy in high-performance, compact-propulsion and mobile power-generation systems has necessitated the development of fuel elements capable of operating at high power densities. High power densities in turn require fuel elements having high thermal conductivities and good fuel retention capabilities at high temperatures. A metal clad fuel element containing a ceramic phase of fuel intimately mixed with and bonded to a continuous refractory metal matrix has been found to satisfy the above requirements. Metal coolant tubes penetrate the matrix to afford internal cooling to the fuel element while providing positive fuel retention and containment of fission products generated within the fuel matrix. Metal header plates are bonded to the coolant tubes at each end of the fuel element and a metal cladding or can completes the fuel-matrix enclosure

  4. Catastrophes caused by corrosion

    OpenAIRE

    PETROVIC ZORAN C.

    2016-01-01

    For many years, huge attention has been paid to the problem of corrosion damage and destruction of metallic materials. Experience shows that failures due to corrosion problems are very important, and statistics at the world level shows that the damage resulting from the effects of various forms of corrosion is substantial and that, for example, in industrialized countries it reaches 4-5% of national incomes. Significant funds are determined annually for the prevention and control of corrosion...

  5. Modelling of Corrosion Cracks

    DEFF Research Database (Denmark)

    Thoft-Christensen, Palle

    Modelling of corrosion cracking of reinforced concrete structures is complicated as a great number of uncertain factors are involved. To get a reliable modelling a physical and mechanical understanding of the process behind corrosion in needed.......Modelling of corrosion cracking of reinforced concrete structures is complicated as a great number of uncertain factors are involved. To get a reliable modelling a physical and mechanical understanding of the process behind corrosion in needed....

  6. Microbiological corrosion of metals

    International Nuclear Information System (INIS)

    Vladislavlev, V.V.

    1992-01-01

    Problems is considered of development of the microbiological corrosion of the NPP equipment. The main attention is paid to the selective character of microbiological corrosion in zones of welded joints of austenitic steels. It is noted that the presence of technological defects promotes growth of corrosional damages. Methods for microbiological corrosion protection are discussed

  7. Management of Reinforcement Corrosion

    DEFF Research Database (Denmark)

    Küter, André; Geiker, Mette Rica; Møller, Per

    Reinforcement corrosion is the most important cause for deterioration of reinforced concrete structures, both with regard to costs and consequences. Thermodynamically consistent descriptions of corrosion mechanisms are expected to allow the development of innovative concepts for the management...... of reinforcement corrosion....

  8. Corrosion of metallic materials. Dry corrosion, aqueous corrosion and corrosion by liquid metal, methods of protection

    International Nuclear Information System (INIS)

    Helie, Max

    2015-01-01

    This book is based on a course on materials given in an engineering school. The author first gives an overview of metallurgy issues: metallic materials (pure metals, metallic alloys), defects of crystal lattices (point defects, linear defects or dislocations), equilibrium diagrams, steels and cast, thermal processing of steels, stainless steels, aluminium and its alloys, copper and its alloys. The second part addresses the properties and characterization of surfaces and interfaces: singularity of a metal surface, surface energy of a metal, energy of grain boundaries, adsorption at a material surface, metal-electrolyte interface, surface oxide-electrolyte interface, techniques of surface analysis. The third chapter addresses the electrochemical aspects of corrosion: description of the corrosion phenomenon, free enthalpy of a compound and free enthalpy of a reaction, case of dry corrosion (thermodynamic aspect, Ellingham diagram, oxidation mechanisms, experimental study, macroscopic modelling), case of aqueous corrosion (electrochemical thermodynamics and kinetics, experimental determination of corrosion rate). The fourth part addresses the different forms of aqueous corrosion: generalized corrosion (atmospheric corrosion, mechanisms and tests), localized corrosion (galvanic, pitting, cracking, intergranular, erosion and cavitation), particular cases of stress cracking (stress corrosion, fatigue-corrosion, embrittlement by hydrogen), and bi-corrosion (of non alloyed steels, of stainless steels, and of aluminium and copper alloys). The sixth chapter addresses the struggle and the protection against aqueous corrosion: methods of prevention, scope of use of main alloys, geometry-based protection of pieces, use of corrosion inhibitors, use of organic or metallic coatings, electrochemical protection. The last chapter proposes an overview of corrosion types in industrial practices: in the automotive industry, in the oil industry, in the aircraft industry, and in the

  9. 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.

  10. Galvanic corrosion of beryllium welds

    International Nuclear Information System (INIS)

    Hill, M.A.; Butt, D.P.; Lillard, R.S.

    1997-01-01

    Beryllium is difficult to weld because it is highly susceptible to cracking. The most commonly used filler metal in beryllium welds is Al-12 wt.% Si. Beryllium has been successfully welded using Al-Si filler metal with more than 30 wt.% Al. This filler creates an aluminum-rich fusion zone with a low melting point that tends to backfill cracks. Drawbacks to adding a filler metal include a reduction in service temperature, a lowering of the tensile strength of the weld, and the possibility for galvanic corrosion to occur at the weld. To evaluate the degree of interaction between Be and Al-Si in an actual weld, sections from a mock beryllium weldment were exposed to 0.1 M Cl - solution. Results indicate that the galvanic couple between Be and the Al-Si weld material results in the cathodic protection of the weld and of the anodic dissolution of the bulk Be material. While the cathodic protection of Al is generally inefficient, the high anodic dissolution rate of the bulk Be during pitting corrosion combined with the insulating properties of the Be oxide afford some protection of the Al-Si weld material. Although dissolution of the Be precipitate in the weld material does occur, no corrosion of the Al-Si matrix was observed

  11. Corrosion in Electronics

    DEFF Research Database (Denmark)

    Ambat, Rajan; Gudla, Helene Virginie Conseil; Verdingovas, Vadimas

    2017-01-01

    and high density packing combined with the use of several materials, which can undergo electrochemical corrosion in the presence of water film formed due to humidity exposure and bias conditions on the PCBA surface. This article provides a short review of the corrosion reliability issues of electronics due...... to the use of electronics under varying humidity conditions. Important PCBA aspects, which are fundamental to the corrosion cell formation under humid conditions, are discussed. Effect of hygroscopic residues from the process and service and their role in assisting water film build up and corrosion...... is presented. Various failure modes resulting from the corrosion and influence factors are discussed including humid and gaseous conditions....

  12. Corrosion of structural materials and electrochemistry in high temperature water of nuclear power systems

    International Nuclear Information System (INIS)

    Uchida, Shunsuke

    2014-01-01

    The latest experiences with corrosion in the cooling systems of nuclear power plants are reviewed. High temperature cooling water causes corrosion of structural materials, which often leads to adverse effects in the plants, e.g., generating defects in materials of major components and fuel claddings, increasing shutdown radiation and increasing the volume of radwaste sources. Corrosion behaviors are much affected by water qualities and differ according to the values of water qualities and the materials themselves. In order to establish reliable operation, each plant requires its own unique optimal water chemistry control based on careful consideration of its system, materials and operational history. Electrochemistry is one of key issues that determine corrosion related problems but it is not the only issue. Most phenomena for corrosion related problems, e.g., flow-accelerated corrosion (FAC), intergranular stress corrosion cracking (IGSCC), primary water stress corrosion cracking (PWSCC) and thinning of fuel cladding materials, can be understood based on an electrochemical index, e.g., electrochemical corrosion potential (ECP), conductivities and pH. The most important electrochemical index, ECP, can be measured at elevated temperature and applied to in situ sensors of corrosion conditions to detect anomalous conditions of structural materials at their very early stages. In the paper, theoretical models based on electrochemistry to estimate wall thinning rate of carbon steel piping due to flow-accelerated corrosion and corrosive conditions determining IGSCC crack initiation and growth rate are introduced. (author)

  13. Waterside corrosion of zirconium alloys in nuclear power plants

    International Nuclear Information System (INIS)

    1998-01-01

    Technically the study of corrosion of zirconium alloys in nuclear power reactors is a very active field and both experimental work and understanding of the mechanisms involved are going through rapid changes. As a result, the lifetime of any publication in this area is short. Because of this it has been decided to revise IAEA-TECDOC-684 - Corrosion of Zirconium Alloys in Nuclear Power Plants - published in 1993. This updated, revised and enlarged version includes major changes to incorporate some of the comments received about the first version. Since this review deals exclusively with the corrosion of zirconium and zirconium based alloys in water, and another separate publication is planned to deal with the fuel-side corrosion of zirconium based fuel cladding alloys, i.e. stress corrosion cracking, it was decided to change the original title to Waterside Corrosion of Zirconium Alloys in Nuclear Power Plants. The rapid changes in the field have again necessitated a cut-off date for incorporating new data. This edition incorporates data up to the end of 1995; including results presented at the 11 International Symposium on Zirconium in the Nuclear Industry held in Garmisch-Partenkirchen, Germany, in September 1995. The revised format of the review now includes: Introductory chapters on basic zirconium metallurgy and oxidation theory; A revised chapter discussing the present extent of our knowledge of the corrosion mechanism based on laboratory experiments; a separate and revised chapter discussing hydrogen uptake; a completely reorganized chapter summarizing the phenomenological observations of zirconium alloy corrosion in reactors; a new chapter on modelling in-reactor corrosion; a revised chapter devoted exclusively to the manner in which irradiation might influence the corrosion process; finally, a summary of our present understanding of the corrosion mechanisms operating in reactor

  14. Effect of ageing time and temperature on corrosion behaviour of aluminum alloy 2014

    Science.gov (United States)

    Gadpale, Vikas; Banjare, Pragya N.; Manoj, Manoranjan Kumar

    2018-03-01

    In this paper, the effect of corrosion behaviour of aluminium alloy 2014 were studied by potentiodynamic polarization in 1 mole of NaCl solution of aged sample. The experimental testing results concluded that, corrosion resistance of Aluminum alloy 2014 degraded with the increasing the temperature (150°C & 200°C) and time of ageing. Corroded surface of the aged specimens was tested under optical microscopes for microstructures for phase analysis. Optical micrographs of corroded surfaces showed general corrosion and pitting corrosion. The corrosion resistance of lower ageing temperature and lower ageing time is higher because of its fine distribution of precipitates in matrix phase.

  15. Corrosion of high purity Fe-Cr-Ni alloys in 13 N boiling nitric acid

    International Nuclear Information System (INIS)

    Ohta, Joji; Mayuzumi, Masami; Kusanagi, Hideo; Takaku, Hiroshi

    1998-01-01

    Corrosion in boiling nitric acid was investigated for high purity Fe-18%Cr-12%Ni alloys and type 304L stainless steels (SS). Owing to very low impurity concentration, the solution treated high purity alloys show almost no intergranular corrosion while the type 304L SS show severe intergranular corrosion. Both in the high purity alloys and type 304L SS, aging treatments ranging from 873 K to 1073 K for 1 h enhance intergranular corrosion. During the aging treatments, impurities should be segregated to the grain boundaries. The corrosion behaviors were discussed from a standpoint of impurity segregation to grain boundaries. This study is of importance for purex reprocessing of spent fuels

  16. Effect of heat treatment on the wear and corrosion behaviors of a gray cast iron coated with a COLMONOY 88 alloy deposited by high velocity oxygen fuel (HVOF thermal spray

    Directory of Open Access Journals (Sweden)

    A. Öz

    2013-07-01

    Full Text Available The present work has been conducted in order to determine the influence of heat treatment on the wear and corrosion behaviours of a gray cast iron substrate coated with a Ni base coating deposited by HVOF thermal spray. The wear resistance of the coatings was obtained using a reciprocating wear tester by rubbing a 10 mm diameter steel ball on the coatings at normal atmospheric conditions. Corrosion tests were performed using potentiodynamic polarization measurements in a 3,5 % NaCl solution. It was observed that the corrosion and wear resistance of the coatings increased along with the reduction of porosity and roughness by the heat treatment.

  17. Double shell slurry low-temperature corrosion tests

    International Nuclear Information System (INIS)

    Divine, J.R.; Bowen, W.M.; McPartland, S.A.; Elmore, R.P.; Engel, D.W.

    1983-09-01

    A series of year-long tests have been completed on potential double shell slurry (DSS) compositions at temperatures up to 100 0 C. These tests have sought data on uniform corrosion, pitting, and stress-corrosion cracking. No indication of the latter two types of corrosion were observed within the test matrix. Corrosion rates after four months were generally below the 1 mpy (25 μm/y) design limit. By the end of twelve months all results were below this limit and, except for very concentrated mixtures, all were below 0.5 mpy. Prediction equations were generated from a model fitted to the data. The equations provide a rapid means of estimating the corrosion rate for proposed DSS compositions

  18. A state-of-the art report on the investigation of the various corrosion models for zirconium-based alloy

    International Nuclear Information System (INIS)

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

    1999-02-01

    The desire to increase uranium utilization and to minimize spent fuel storage requirements provides an incentive to extend the average fuel rod discharge burnup to about 70,000MWd/MTU. For these higher burnups data are needed to determine if waterside corrosion of the cladding may be a life-limiting feature of fuel rod design. It is apparent that many factors can influence waterside corrosion, and these need to be better understood in order to minimize corrosion at these higher target burnups. The objective of this report is to review published data relevant to the corrosion of Zircaloy under PWR operating conditions. (author). 100 refs., 4 tabs., 21 figs

  19. Irradiation experience with FRAMATOME fuel

    International Nuclear Information System (INIS)

    Rim, C.S.; Texier, C.; Traccucci, R.

    1978-01-01

    This program principally consists of: - monitoring of the reactor coolant activity due to fission and corrosion products, - on-site non-destructive examinations (visual, dimensional, gamma spectrometry, etc.) on irradiated fuel assemblies, - on-site and hot cell examinations on removable fuel rods. Additional tests are also in progress in order to improve models used in fuel rod and fuel assembly design and to verify the technical limits of fuel operation with respect to power ramp, daily load follow, load regulation, etc. The objective of this paper is to review the behavior of FRAMATOME fuel in power reactors

  20. Hydrogen in CANDU fuel elements

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  1. Increased corrosion resistance of basalt reinforced cement compositions with nanosilica

    Directory of Open Access Journals (Sweden)

    URKHANOVA Larisa Alekseevna

    2014-08-01

    Full Text Available Disperse fiber reinforcement is used to improve deformation and shrinkage characteristics, flexural strength of concrete. Basalt roving and thin staple fiber are often used as mineral fibers. The paper considers the problems of using thin basalt fiber produced by centrifugal-blow method. Evaluation of the corrosion resistance of basalt fiber as part of the cement matrix was performed. Nanodispersed silica produced by electron beam accelerator was used to increase corrosion resistance of basalt fiber.

  2. Nuclear Repository steel canister: experimental corrosion rates

    Science.gov (United States)

    Caporuscio, F.; Norskog, K.

    2017-12-01

    The U.S. Spent Fuel & Waste Science & Technology campaign evaluates various generic geological repositories for the disposal of spent nuclear fuel. This experimental work analyzed and characterized the canister corrosion and steel interface mineralogy of bentonite-based EBS 304 stainless steel (SS), 316 SS, and low-carbon steel coupons in brine at higher heat loads and pressures. Experiments contrasted EBS with and without an argillite wall rock. Unprocessed bentonite from Colony, Wyoming simulated the clay buffer and Opalinus Clay represented the wall rock. Redox conditions were buffered at the magnetite-iron oxygen fugacity univariant curve. A K-Na-Ca-Cl-based brine was chosen to replicate generic granitic groundwater compositions, while Opalinous Clay groundwater was used in the wall rock series of experiments. Most experiments were run at 150 bar and 300°C for 4 to 6 weeks and one was held at elevated conditions for 6 months. The two major experimental mixtures were 1) brine-bentonite clay- steel, and 2) brine-bentonite clay-Opalinus Clay-steel. Both systems were equilibrated at a high liquid/clay ratio. Mineralogy and aqueous geochemistry of each experiment were evaluated to monitor the reactions that took place. In total 4291 measurements were obtained: 2500 measured steel corrosion depths and 1791 were of phyllosilicate mineral reactions/growths at the interface. The low carbon steel corrosion mechanism was via pit corrosion, while 304 SS and 316 SS were by general corrosion. The low carbon steel corrosion rate (1.95 μm/day) was most rapid. The 304 SS corrosion rate (0.37 μm/day) was slightly accelerated versus the 316 SS corrosion rate (0.26 μm/day). Note that the six month 316 SS experiment shows inhibited corrosion rates (0.07 μm/day). This may be in part due to mantling by the Fe-saponite/chlorite authigenic minerals. All phyllosilicate growth rates at the interface exhibit similar growth rate patterns to the steels (i.e. LCS>304>316> 316 six month).

  3. Monitoring Microbially Influenced Corrosion

    DEFF Research Database (Denmark)

    Hilbert, Lisbeth Rischel

    Abstract Microbially influenced corrosion (MIC) of carbon steel may occur in media with microbiological activity of especially sulphate-reducing bacteria (SRB). The applicability and reliability of a number of corrosion monitoring techniques for monitoring MIC has been evaluated in experiments...... and diffusional effects and unreliable corrosion rates, when biofilm and ferrous sulphide corrosion products cover the steel surface. Corrosion rates can be overestimated by a factor of 10 to 100 by electrochemical techniques. Weight loss coupons and ER are recommended as necessary basic monitoring techniques....... EIS might be used for detection of MIC as the appearance of very large capacitances can be attributed to the combined ferrous sulphide and biofilm formation. Capacitance correlates directly with sulphide concentration in sterile sulphide media. Keywords: Corrosion monitoring, carbon steel, MIC, SRB...

  4. SRB seawater corrosion project

    Science.gov (United States)

    Bozack, M. J.

    1991-01-01

    The corrosion behavior of 2219 aluminum when exposed to seawater was characterized. Controlled corrosion experiments at three different temperatures (30, 60 and 100 C) and two different environments (seawater and 3.5 percent salt solution) were designed to elucidate the initial stages in the corrosion process. It was found that 2219 aluminum is an active catalytic surface for growth of Al2O3, NaCl, and MgO. Formation of Al2O3 is favored at lower temperatures, while MgO is favored at higher temperatures. Visible corrosion products are formed within 30 minutes after seawater exposure. Corrosion characteristics in 3.5 percent salt solution are different than corrosion in seawater. Techniques utilized were: (1) scanning electron microscopy, (2) energy dispersive x-ray spectroscopy, and (3) Auger electron spectroscopy.

  5. New Storage Mode for Spent Fuel at the Budapest Research Reactor (Encapsulation of the fuel for semi-dry storage)

    International Nuclear Information System (INIS)

    Hargitai, T.; Vidovszky, I.

    2002-01-01

    The Budapest Research Reactor is a light water cooled and moderated tank type reactor. The reactor was first commissioned in 1959; its principal functions at that time were to serve as a facility for basic research experiments in the frameworks of research programmes of the Hungarian Academy of Sciences and in industrial development projects. In the first years of operation (until 1966) the power was 2 MW; the type of the fuel used was EK-10. In 1967 the reactor was upgraded, the power was increased to 5 MW, as a new fuel WWR-SM (36%) was introduced, the core grid was replaced and also beryllium reflector was put around the core. In this upgrading project a new wet spent fuel storage facility (AFR pond) was constructed near to the reactor hall. In 1986 the second (full scale) reconstruction of the reactor started, the new reactor with unchanged fuel type was first critical in December 1992. After the reconstruction a new type of fuel was introduced, the so-called VVR-M2, with the same geometry and enrichment as VVR-SM, but with different material composition, i.e. UO 2 granulate dispersed in aluminum matrix. Up to the present the reactor has been in operation for more than 43 years and all the spent fuel produced since the first commissioning of the reactor is stored in wet facilities on site. There is no real experience of wet storage for so long time worldwide, but it is well known that aluminium corrosion can accelerate rapidly once it has started. That is the reason why the operating organization of the BRR facility decided to encapsulate all the stored assemblies and keep them in inert gas atmosphere in order to slow down or to stop corrosion processes. The present paper describes the semi-dry storage transition option, i.e. the design and the operation of the canning device, as well as the canning procedure in detail. (author)

  6. Microbiologically Influenced Corrosion

    Science.gov (United States)

    2015-11-05

    Widdel, F. (2012) Marine sulfate-reducing bacteria cause serious corrosion of iron under electroconductive biogenic mineral crust. Envi- ronmental...inCluding oil and gas, from to 2011 [2]. Over that period of time approximately of all releases were attributed to corrosion . National ;sociati<Jn...c>fComJsicmEngineers (NACE) International [3] the cost of corrosion for onshore gas and liquid nsn1ission pipelines was $7 billion. However, there

  7. Matrix theory

    CERN Document Server

    Franklin, Joel N

    2003-01-01

    Mathematically rigorous introduction covers vector and matrix norms, the condition-number of a matrix, positive and irreducible matrices, much more. Only elementary algebra and calculus required. Includes problem-solving exercises. 1968 edition.

  8. Grain boundary corrosion of copper canister material

    International Nuclear Information System (INIS)

    Fennell, P.A.H.; Graham, A.J.; Smart, N.R.; Sofield, C.J.

    2001-03-01

    The proposed design for a final repository for spent fuel and other long-lived residues in Sweden is based on the multi-barrier principle. The waste will be encapsulated in sealed cylindrical canisters, which will then be placed in granite bedrock and surrounded by compacted bentonite clay. The canister design is based on a thick cast inner container fitted inside a corrosion-resistant copper canister. During fabrication of the outer copper canisters there will be some unavoidable grain growth in the welded areas. As grains grow they will tend to concentrate impurities within the copper at the new grain boundaries. The work described in this report was undertaken to determine whether there is any possibility of enhanced corrosion at grain boundaries within the copper canister. The potential for grain boundary corrosion was investigated by exposing copper specimens, which had undergone different heat treatments and hence had different grain sizes, to aerated artificial bentonite-equilibrated groundwater with two concentrations of chloride, for increasing periods of time. The degree of grain boundary corrosion was determined by atomic force microscopy (AFM) and optical microscopy. AFM showed no increase in grain boundary 'ditching' for low chloride groundwater. In high chloride groundwater the surface was covered uniformly with a fine-grained oxide. No increases in oxide thickness were observed. No significant grain boundary attack was observed using optical microscopy either. The work suggests that in aerated artificial groundwaters containing chloride ions, grain boundary corrosion of copper is unlikely to adversely affect SKB's copper canisters

  9. Oxidation kinetics of reaction products formed in uranium metal corrosion

    International Nuclear Information System (INIS)

    Totemeier, T. C.

    1998-01-01

    The oxidation behavior of uranium metal ZPPR fuel corrosion products in environments of Ar-4%O 2 and Ar-20%O 2 were studied using thermo-gravimetric analysis (TGA). These tests were performed to extend earlier work in this area specifically, to assess plate-to-plate variations in corrosion product properties and the effect of oxygen concentration on oxidation behavior. The corrosion products from two relatively severely corroded plates were similar, while the products from a relatively intact plate were not reactive. Oxygen concentration strongly affected the burning rate of reactive products, but had little effect on low-temperature oxidation rates

  10. Coatings and Corrosion Laboratory

    Data.gov (United States)

    Federal Laboratory Consortium — Purpose: The mission of the Coatings and Corrosion Laboratory is to develop and analyze the effectiveness of innovative coatings test procedures while evaluating the...

  11. Localized Corrosion Behavior of Type 304SS with a Silica Layer Under Atmospheric Corrosion Environments

    International Nuclear Information System (INIS)

    E. Tada; G.S. Frankel

    2006-01-01

    The U.S. Department of Energy (DOE) has proposed a potential repository for spent nuclear fuel and high-level radioactive waste at the Yucca Mountain site in Nevada. [I] The temperature could be high on the waste packages, and it is possible that dripping water or humidity could interact with rock dust particulate to form a thin electrolyte layer with concentrated ionic species. Under these conditions, it is possible that highly corrosion-resistant alloys (CRAs) used as packages to dispose the nuclear waste could suffer localized corrosion. Therefore, to better understand long-term corrosion performance of CRAs in the repository, it is important to investigate localized corrosion under a simulated repository environment. We measured open circuit potential (OCP) and galvanic current (i g ) for silica-coated Type 304SS during drying of salt solutions under controlled RH environments to clarify the effect of silica layer as a dust layer simulant on localized corrosion under atmospheric environments. Type 304SS was used as a relatively susceptible model CRA instead of the much more corrosion resistant alloys, such as Alloy 22, that are being considered as, waste package materials

  12. Radiation resistant ceramic matrix composites

    International Nuclear Information System (INIS)

    Jones, R.H.; Steiner, D.; Heinisch, H.L.; Newsome, G.A.; Kerch, H.M.

    1997-01-01

    Ceramic matrix composites are of interest for nuclear applications because of their high-temperature properties, corrosion resistance, fracture toughness relative to monolithic ceramics, and low neutron activation and after heat. Evaluations of the radiation resistance of commercially available SiC/SiC composites have revealed their promise for this application, but also the need for further development to achieve the desired performance. This paper summarizes the results of a workshop cosponsored by the Offices of Fusion Energy and Basic Energy Sciences of the US Department of Energy and Lockheed-Martin Corporation with forty attendees from national laboratories, universities and industry. A number of promising routes for optimizing the radiation stability of ceramic matrix composites were identified at this workshop. These routes included the newer, more stoichiometric fibers and alternate fiber/matrix interfaces and matrix processing routes. (orig.)

  13. Zirconium alloy barrier having improved corrosion resistance

    International Nuclear Information System (INIS)

    Adamson, R.B.; Rosenbaum, H.S.

    1983-01-01

    A nuclear fuel element for use in the core of a nuclear reactor has a composite cladding container having a substrate and a dilute zirconium alloy liner bonded to the inside surface of the substrate. The dilute zirconium alloy liner forms about 1 to about 20 percent of the thickness of the cladding and is comprised of zirconium and a metal selected from the group consisting of iron, chromium, iron plus chromium, and copper. The dilute zirconium alloy liner shields the substrate from impurities or fission products from the nuclear fuel material and protects the substrate from stress corrosion and stress cracking. The dilute zirconium alloy liner displays greater corrosion resistance, especially to oxidation by hot water or steam than unalloyed zirconium. The substrate material is selected from conventional cladding materials, and preferably is a zirconium alloy. (author)

  14. Corrosion in the oil industry

    Energy Technology Data Exchange (ETDEWEB)

    Brondel, D. (Sedco Forex, Montrouge (France)); Edwards, R. (Schlumberger Well Services, Columbus, OH (United States)); Hayman, A. (Etudes et Productions Schlumberger, Clamart (France)); Hill, D. (Schlumberger Dowell, Tulsa, OK (United States)); Mehta, S. (Schlumberger Dowell, St. Austell (United Kingdom)); Semerad, T. (Mobil Oil Indonesia, Inc., Sumatra (Indonesia))

    1994-04-01

    Corrosion costs the oil industry billions of dollars a year, a fact that makes the role of the corrosion engineer an increasingly important one. Attention is paid to how corrosion affects every aspect of exploration and production, from offshore rigs to casing. Also the role of corrosion agents such as drilling and production fluids is reviewed. Methods of control and techniques to monitor corrosion are discussed, along with an explanation of the chemical causes of corrosion. 21 figs., 32 refs.

  15. Contribution and modelization of basic processes for development of the spent nuclear fuel matrix; Contribucion experimental y modelizacion de procesos basicos para el desarrollo del modelo de alteracion de la matriz delcombustible irradiado

    Energy Technology Data Exchange (ETDEWEB)

    Pablo, J. de; Casa, I.; Clarens, F.; Gimenez, J.; Rovira, M.

    2003-07-01

    The behaviour of the spent nuclear fuel in the conditions expected in a deep geological repository depends on some basic processes such as the formation of oxidizing species due to the radiolysis of water, the effect of these species on the spent nuclear fuel matrix oxidation and dissolution, and the precipitation of secondary solid phases. In this work, some of those processes have been studied in detail. In particular, both uranium dioxide oxidation and dissolution rates have been determined in the presence of oxygen an hydrogen per-oxide (the most important molecular species formed in the radiolysis of water). In addition, the precipitation of secondary phases on the UO{sub 2} surfaces in the presence of hydrogen peroxide has also been fool owed by means of the Atomic Force Microscope allowing the identification of studies (UO{sub 4}.H{sub 2}O). On the other hand, the radiolytical models of the dissolution of spent nuclear fuel need to estimate the surface site densities of the solid. In this sense, we have studied surface site densities of three different uranium oxides: UO{sub 2}, U{sub 3}O{sub 8}, and UO{sub 3}.4H{sub 2}O. (Author) 42 refs.

  16. Microprobe study of fission product behavior in high-burnup HTR fuels

    International Nuclear Information System (INIS)

    Kleykamp, H.

    Electron microprobe analysis of irradiated coated particles with high burnup (greater than 50 percent fima) gives detailed information on the chemical state and the transport behavior of the fission products in UO 2 and UC 2 kernels and in the coatings. In oxide fuel kernels, metallic inclusions and ceramic precipitations are observed. The solubility behavior of the fission products in the fuel matrix has been investigated. Fission product inclusions could not be detected in carbide fuel kernels; post irradiation annealed UC 2 kernels, however, give information on the element combinations of some fission product phases. Corresponding to the chemical state in the kernel, Cs, Sr, Ba, Pd, Te and the rare earths are released easily and diffuse through the entire pyrocarbon coating. These fission products can be retained by a silicon carbide layer. The initial stage of a corrosive attack of the SiC coating by the fission products is evidenced

  17. The anaerobic corrosion of carbon steel in concrete

    International Nuclear Information System (INIS)

    Naish, C.C.; Balkwill, P.H.; O'Brien, T.M.; Taylor, K.J.; Marsh, G.P.

    1990-11-01

    The report describes the work of a two year programme investigating the anaerobic corrosion of carbon steel embedded in a range of candidate repository cements and concretes at laboratory temperatures. The factors investigated in the study were the rate of the anaerobic corrosion reaction, the effect of hydrogen overpressure on the reaction rate and the form of the corrosion product. Both electrochemical and sample weight loss corrosion rate measurements were used. The cements and concretes used were prepared both with and without small additions of chloride (2% by weight of mix water). The results indicate that the corrosion rate is low, < 1 μm/year, the effect of hydrogen overpressure is not significant over the range of pressures investigated, 1-100 atmospheres, and that the corrosion product is dependent on the cement used to cast the samples. Magnetite was identified in the case of blast furnace slag replacement cements but for pulverised fuel ash and ordinary Portland cements no corrosion product was evident either from X-ray diffraction or laser Raman measurements. Further work is presently underway to investigate the effects of elevated temperatures and chloride levels on the anaerobic corrosion reaction and the rate of hydrogen gas production. (author)

  18. Alternating current techniques for corrosion monitoring in water reactor systems

    International Nuclear Information System (INIS)

    Isaacs, H.S.; Weeks, J.R.

    1977-01-01

    Corrosion in both nuclear and fossil fueled steam generators is generally a consequence of the presence of aggressive impurities introduced into the coolant system through condenser leakage. The impurities concentrate in regions of the steam generator protected from coolant flow, in crevices or under deposited corrosion products and adjacent to heat transfer surfaces. These three factors, the aggressive impurity, crevice type areas and heat transfer surfaces appear to be the requirements for the onset of rapid corrosion. Under conditions where coolant impurities do not concentrate the corrosion rates are low, easily measured and can be accounted for by allowances in the design of the steam generator. Rapid corrosion conditions cannot be designed for and must be suppressed. The condition of the surfaces when rapid corrosion develops must be markedly different from those during normal operation and these changes should be observable using electrochemical techniques. This background formed the basis of a design of a corrosion monitoring device, work on which was initiated at BNL. The basic principles of the technique are described. The object of the work is to develop a corrosion monitoring device which can be operated with PWR steam generator secondary coolant feed water

  19. Solution exchange corrosion testing with the glass-zeolite ceramic waste form in demineralized water at 900C

    International Nuclear Information System (INIS)

    Simpson, L. J.

    1998-01-01

    A ceramic waste form of glass-bonded zeolite is being developed for the long-term disposition of fission products and transuranic elements in wastes from the U.S. Department of Energy's spent nuclear fuel conditioning activities. Solution exchange corrosion tests were performed on the ceramic waste form and its potential base constituents of glass, zeolite 5A, and sodalite as part of an effort to qualify the ceramic waste form for acceptance into the Civilian Radioactive Waste Management System. Solution exchange tests were performed at 90 C by replacing 80 to 90% of the leachate with fresh demineralized water after set time intervals. The results from these tests provide information about corrosion mechanisms and the ability of the ceramic waste form and its constituent materials to retain waste components. The results from solution exchange tests indicate that radionuclides will be preferentially retained in the zeolites without the glass matrix and in the ceramic waste form, with respect to cations like Li, K, and Na. Release results have been compared for simulated waste from candidate ceramic waste forms with zeolite 5A and its constituent materials to determine the corrosion behavior of each component

  20. Impact of Aluminum on Anticipated Corrosion in a Flooded SNF Multi Canister Overpack (MCO)

    Energy Technology Data Exchange (ETDEWEB)

    DUNCAN, D.R.

    1999-07-06

    Corrosion reactions in a flooded MCO are examined to determine the impact of aluminum corrosion products (from aluminum basket grids and spacers) on bound water estimates and subsequent fuel/environment reactions during storage. The mass and impact of corrosion products were determined to be insignificant, validating the choice of aluminum as an MCO component and confirming expectations that no changes to the Technical Databook or particulate mass or water content are necessary.

  1. Archaeological analogs and corrosion

    International Nuclear Information System (INIS)

    David, D.

    2008-01-01

    In the framework of the high level and long life radioactive wastes disposal deep underground, the ANDRA built a research program on the material corrosion. In particular they aim to design containers for a very long time storage. Laboratory experiments are in progress and can be completed by the analysis of metallic archaeological objects and their corrosion after hundred years. (A.L.B.)

  2. Corrosion evaluation technology

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Uh Chul; Han, Jeong Ho; Nho, Kye Ho; Lee, Eun Hee; Kim, Hong Pyo; Hwang, Seong Sik; Lee, Deok Hyun; Hur, Do Haeng; Kim, Kyung Mo

    1997-09-01

    A multifrequency ACPD system was assembled which can measure very small crack. Stress corrosion cracking test system with SSRT operating high temperature was installed. Stress corrosion cracking test of newly developed alloy 600 and existing alloy 600 was carried out in steam atmosphere of 400 deg C. No crack was observed in both materials within a test period of 2,000 hrs. Corrosion fatigue test system operating at high temperature was installed in which fatigue crack was measured by CDPD. Lead enhanced the SCC of the Alloy 600 in high temperature water, had a tendency to modify a cracking morphology from intergranular to transgranular. Pit initiation preferentially occurred at Ti-rich carbide. Resistance to pit initiation decreased with increasing temperature up to 300 deg C. Test loop for erosion corrosion was designed and fabricated. Thin layer activation technique was very effective in measuring erosion corrosion. Erosion corrosion of a part of secondary side pipe was evaluated by the Check Family Codes of EPRI. Calculated values of pipe thickness by Check Family Codes coincided with the pipe thickness measured by UT with an error of {+-} 20%. Literature review on turbine failure showed that failure usually occurred in low pressure turbine rotor disc and causes of failure are stress corrosion cracking and corrosion fatigue. (author). 12 refs., 20 tabs., 77 figs.

  3. Demystifying Controlling Copper Corrosion

    Science.gov (United States)

    The LCR systematically misses the highest health and corrosion risk sites for copper. Additionally, there are growing concerns for WWTP copper in sludges and discharge levels. There are many corrosion control differences between copper and lead. This talk explains the sometimes c...

  4. Effect of TiN particulate reinforcement on corrosive behaviour of ...

    Indian Academy of Sciences (India)

    2010), quartz (Nair et al 1985) and glass fibre (Pohlman. 1978) exhibited higher corrosion resistance compared to their matrix alloys. Recent studies on the corrosion resistance of TiB2 particulate-reinforced A356 alloy show a marked. ∗. Author for correspondence (anandkps7@yahoo.com) decrease with increase in TiB2 ...

  5. Molten carbonate fuel cell separator

    Science.gov (United States)

    Nickols, R.C.

    1984-10-17

    In a stacked array of molten carbonate fuel cells, a fuel cell separator is positioned between adjacent fuel cells to provide isolation as well as a conductive path therebetween. The center portion of the fuel cell separator includes a generally rectangular, flat, electrical conductor. Around the periphery of the flat portion of the separator are positioned a plurality of elongated resilient flanges which form a gas-tight seal around the edges of the fuel cell. With one elongated flange resiliently engaging a respective edge of the center portion of the separator, the sealing flanges, which are preferably comprised of a noncorrosive material such as an alloy of yttrium, iron, aluminum or chromium, form a tight-fitting wet seal for confining the corrosive elements of the fuel cell therein. This arrangement permits a good conductive material which may be highly subject to corrosion and dissolution to be used in combination with a corrosion-resistant material in the fuel cell separator of a molten carbonate fuel cell for improved fuel cell conductivity and a gas-tight wet seal.

  6. Corrosive Poisonings in Adults

    Science.gov (United States)

    Chibishev, Andon; Pereska, Zanina; Chibisheva, Vesna; Simonovska, Natasa

    2012-01-01

    Ingestion of corrosive substances may cause severe to serious injuries of the upper gastrointestinal tract and the poisoning can even result in death. Acute corrosive intoxications pose a major problem in clinical toxicology since the most commonly affected population are the young with psychic disorders, suicidal intent and alcohol addiction. The golden standard for determination of the grade and extent of the lesion is esophagogastroduodenoscopy performed in the first 12-24 hours following corrosive ingestion. The most common late complications are esophageal stenosis, gastric stenosis of the antrum and pyloris, and rarely carcinoma of the upper gastrointestinal tract. Treatment of the acute corrosive intoxications include: neutralization of corrosive agents, antibiotics, anti-secretory therapy, nutritional support, collagen synthesis inhibitors, esophageal dilation and stent placement, and surgery. PMID:23678319

  7. Oxidative dissolution of spent fuel and release of nuclides from a copper/iron canister. Model developments and applications

    Energy Technology Data Exchange (ETDEWEB)

    Longcheng Liu

    2001-12-01

    Three models have been developed and applied in the performance assessment of a final repository. They are based on accepted theories and experimental results for known and possible mechanisms that may dominate in the oxidative dissolution of spent fuel and the release of nuclides from a canister. Assuming that the canister is breached at an early stage after disposal, the three models describe three sub-systems in the near field of the repository, in which the governing processes and mechanisms are quite different. In the model for the oxidative dissolution of the fuel matrix, a set of kinetic descriptions is provided that describes the oxidative dissolution of the fuel matrix and the release of the embedded nuclides. In particular, the effect of autocatalytic reduction of hexavalent uranium by dissolved H{sub 2}, using UO{sub 2} (s) on the fuel pellets as a catalyst, is taken into account. The simulation results suggest that most of the radiolytic oxidants will be consumed by the oxidation of the fuel matrix, and that much less will be depleted by dissolved ferrous iron. Most of the radiolytically produced hexavalent uranium will be reduced by the autocatalytic reaction with H{sub 2} on the fuel surface. It will reprecipitate as UO{sub 2} (s) on the fuel surface, and thus very little net oxidation of the fuel will take place. In the reactive transport model, the interactions of multiple processes within a defective canister are described, in which numerous redox reactions take place as multiple species diffuse. The effect of corrosion of the cast iron insert of the canister and the reduction of dissolved hexavalent uranium by ferrous iron sorbed onto iron corrosion products and by dissolved H{sub 2} are particularly included. Scoping calculations suggest that corrosion of the iron insert will occur primarily under anaerobic conditions. The escaping oxidants from the fuel rods will migrate toward the iron insert. Much of these oxidants will, however, be consumed

  8. Material problems in high temperature corrosive environments and their solution by surface technologies

    International Nuclear Information System (INIS)

    Anzai, Toshio; Shibata, Keiichi

    1992-01-01

    The materials for high temperature facilities are usually determined by the heat resistant strength except special corrosive environment. High temperature corrosion has been considered in the case of using low quality fuel which is avoided recently. Even though fuel is clean, high temperature corrosion occurs due to the rise of design temperature and the intake of sea salt particles. There is also one more high temperature corrosion problem on process side that consumes heat energy. The research on high temperature corrosion is important for realizing new energy system such as molten carbonate type fuel cell, pulverized coal-firing supercritical power generation and high temperature gas-cooled nuclear reactor. In this report, the examples experienced actually in a chemical plant are examined, and the possibility of material technology against high temperature corrosion is reviewed. Corrosion phenomena in high temperature environment, the examples of corrosion in the high temperature facilities in a chemical plant and the material technology as the countermeasures for corrosion prevention are reported. (K.I.)

  9. Boiler corrosion (citations from the NTIS data base). Report for 1964-Jul 76

    International Nuclear Information System (INIS)

    Smith, M.F.

    1976-07-01

    Research on design, improved efficiency, materials, cathodic protection, corrosion inhibiting additives and combustion in coal and fuel oil fired boilers are cited. Corrosion from limestone injection for pollution control, magnetohydrodynamics, ship boilers, and nuclear power plant boilers are included. (This updated bibliography contains 86 abstracts, 9 of which are new entries to the previous edition.)

  10. Corrosion of structural materials and electrochemistry in high temperature water of nuclear power systems

    International Nuclear Information System (INIS)

    Uchida, Shunsuke

    2008-01-01

    The latest experiences with corrosion in the cooling systems of nuclear power plants are reviewed. High temperature cooling water causes corrosion of structural materials, which often leads to adverse effects in the plants, e.g., increased shutdown radiation, generation of defects in materials of major components and fuel claddings, and increased volume of radwaste sources. Corrosion behavior is greatly affected by water quality and differs according to the water quality values and the materials themselves. In order to establish reliable operation, each plant requires its own unique optimal water chemistry control based on careful consideration of its system, materials and operational history. Electrochemistry is one of the key issues that determine corrosion-related problems, but it is not the only issue. Most corrosion-related phenomena, e.g., flow accelerated corrosion (FAC), intergranular stress corrosion cracking (IGSCC), primary water stress corrosion cracking (PWSCC) and thinning of fuel cladding materials, can be understood based on an electrochemical index, e.g., the electrochemical corrosion potential (ECP), conductivities and pH. The most important electrochemical index, the ECP, can be measured at elevated temperature and applied to in situ sensors of corrosion conditions to detect anomalous conditions of structural materials at their very early stages. (orig.)

  11. Corrosion Failures in Marine Environment

    OpenAIRE

    R. Krishnan

    1985-01-01

    This paper gives a brief description of typical marine environments and the most common form of corrosion of materials used in this environment. Some typical case histories of failures pertaining to pitting, bimetallic corrosion, dealloying, cavitation and stress corrosion cracking are illustrated as typical examples of corrosion failures.

  12. Automated Methods Of Corrosion Measurements

    DEFF Research Database (Denmark)

    Bech-Nielsen, Gregers; Andersen, Jens Enevold Thaulov; Reeve, John Ch

    1997-01-01

    The chapter describes the following automated measurements: Corrosion Measurements by Titration, Imaging Corrosion by Scanning Probe Microscopy, Critical Pitting Temperature and Application of the Electrochemical Hydrogen Permeation Cell.......The chapter describes the following automated measurements: Corrosion Measurements by Titration, Imaging Corrosion by Scanning Probe Microscopy, Critical Pitting Temperature and Application of the Electrochemical Hydrogen Permeation Cell....

  13. 77 FR 35677 - Regulation of Fuel and Fuel Additives; Modification to Octamix Waiver (TOLAD)

    Science.gov (United States)

    2012-06-14

    ...'s gasoline-alcohol fuel, OCTAMIX. ADDRESSES: EPA has established a docket for this action under... Corporation received a waiver under CAA section 211(f)(4) for a gasoline-alcohol fuel blend, known as OCTAMIX... alternative corrosion inhibitor, TOLAD\\TM\\ MFA-10A, in the OCTAMIX gasoline-alcohol fuel blend which otherwise...

  14. Nuclear fuel rods

    International Nuclear Information System (INIS)

    Wada, Toyoji.

    1979-01-01

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

  15. Drying of mock spent nuclear fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Crepeau, J.C.; Reese, S.; McIlroy, H.M. Jr. [Univ. of Idaho, Idaho Falls, ID (United States). Dept. of Mechanical Engineering; Lords, R.E. [Lockheed Martin Idaho Technologies Co., Idaho Falls, ID (United States)

    1998-03-01

    Spent nuclear fuel elements are stored in underwater cooling pools until the elements can be safely handled and prepared for interim dry storage. The fuel was intended for short-term storage in water before it was to be reprocessed. However, the fuel will no longer be reprocessed, and extended storage in water has caused many of the aluminum-clad elements to degrade, exposing the uranium fuel. In addition, sludge, comprised of corroded aluminum and sediment, has accumulated in and around the fuel plates. The water in the sludge must be removed before the spent fuel elements can be placed in dry storage. Experiments have been performed on mock spent fuel elements with simulated corrosion product applied between the plates. A series of vacuum and heating cycles were used to dry the elements, and a mixture of clay and aluminum oxide was used to simulate corrosion products on the elements. The procedures used in the experiments were determined to be adequate to dry the mock spent fuel elements, and the temperature behavior of the simulated corrosion product within the fuel elements could be used to determine when the element was dry. On plates where areas of wet simulant were found, a sharp frying front was observed that separated the wet and dry parts of the simulated corrosion product. The drying front propagated inward towards the center of the mock fuel elements over time.

  16. Fuel flexible fuel injector

    Science.gov (United States)

    Tuthill, Richard S; Davis, Dustin W; Dai, Zhongtao

    2015-02-03

    A disclosed fuel injector provides mixing of fuel with airflow by surrounding a swirled fuel flow with first and second swirled airflows that ensures mixing prior to or upon entering the combustion chamber. Fuel tubes produce a central fuel flow along with a central airflow through a plurality of openings to generate the high velocity fuel/air mixture along the axis of the fuel injector in addition to the swirled fuel/air mixture.

  17. Fuel cell

    Energy Technology Data Exchange (ETDEWEB)

    Enomoto, Hirofumi.

    1989-05-22

    This invention aims to maintain a long-term operation with stable cell output characteristics by uniformly supplying an electrolyte from the reserver to the matrix layer over the entire matrix layer, and further to prevent the excessive wetting of the catalyst layer by smoothly absorbing the volume change of the electrolyte, caused by the repeated stop/start-up of the fuel cell, within the reserver system. For this purpose, in this invention, an electrolyte transport layer, which connects with an electrolyte reservor formed at the electrode end, is partly formed between the electrode material and the catalyst layer; a catalyst layer, which faces the electrolyte transport layer, has through-holes, which connect to the matrix, dispersely distributed. The electrolyte-transport layer is a thin sheet of a hydrophilic fibers which are non-wovens of such fibers as carbon, silicon carbide, silicon nitride or inorganic oxides. 11 figs.

  18. Surface films and corrosion of copper

    International Nuclear Information System (INIS)

    Hilden, J.; Laitinen, T.; Maekelae, K.; Saario, T.; Bojinov, M.

    1999-03-01

    In Sweden and Finland the spent nuclear fuel is planned to be encapsulated in cast iron canisters that have an outer shield made of copper. The copper shield is responsible for the corrosion protection of the canister construction. General corrosion of the copper is not expected to be the limiting factor in the waste repository environment when estimating the life-time of the canister construction. However, different forms of localised corrosion, i.e. pitting, stress corrosion cracking, or environmentally assisted creep fracture may cause premature failure of the copper shield. Of the probable constituents in the groundwater, nitrites, chlorides, sulphides and carbonates have been suggested to promote localised corrosion of copper. The main assumption made in planning this research program is that the surface films forming on copper in the repository environment largely determine the susceptibility of copper to the different forms of localised corrosion. The availability of reactants, which also may become corrosion rate limiting, is investigated in several other research programs. This research program consists of a set of successive projects targeted at characterising the properties of surface films on copper in repository environment containing different detrimental anions. A further aim was to assess the significance of the anion-induced changes in the stability of the oxide films with regard to localised corrosion of copper. This report summarises the results from a series of investigations on properties of surface films forming on copper in water of pH = 8.9 at temperature of 80 deg C and pressure of 2 MPa. The main results gained so far in this research program are as follows: The surface films forming on copper in the thermodynamic stability region of monovalent copper at 80 deg C consist of a bulk part (about 1 mm thick) which is a good ionic and electronic conductor, and an outer, interfacial layer (0.001 - 0.005 mm thick) which shows p-type semiconductor

  19. 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.

  20. Corrosion of PWR steam generators

    International Nuclear Information System (INIS)

    Garnsey, R.

    1979-01-01

    Some designs of pressurized water reactor (PWR) steam generators have experienced a variety of corrosion problems which include stress corrosion cracking, tube thinning, pitting, fatigue, erosion-corrosion and support plate corrosion resulting in 'denting'. Large international research programmes have been mounted to investigate the phenomena. The operational experience is reviewed and mechanisms which have been proposed to explain the corrosion damage are presented. The implications for design development and for boiler and feedwater control are discussed. (author)

  1. CORROSION OF LEAD SHIELDING IN MODEL 9975 PACKAGE

    Energy Technology Data Exchange (ETDEWEB)

    Subramanian, K

    2006-03-15

    Experiments were performed to determine the corrosion rate of lead when exposed to off-gas or degradation products of organic materials used in the model 9975 package.[1] The experiments were completed within the framework of a parametric test matrix with variables of organic configuration, temperature, humidity and the effect of durations of exposure on the corrosion of lead in the 9975 package. The room temperature vulcanizing (RTV) sealant was the most corrosive organic species in the testing, followed by the polyvinyl acetate (PVAc) glue. The Celotex{copyright} material uniquely induced measurable corrosion only in situations with condensed water, and to a much lesser extent than the PVAc glue and RTV. The coupons exhibited faster corrosion at higher temperatures than at room temperatures. There was a particularly pronounced effect of condensed water as the coupons exposed in the cells with condensed water exhibited much higher corrosion rates. In the 9975 package, the PVAc glue was determined to be the most aggressive due to it's proximity in the design. The condition considered most representative of the package conditions is that of the coupon exposed to the Celotex{copyright}/glue organic exposed in the ambient humidity conditions. The corrosion rate of 2 mpy measured in the laboratory experiments for this condition is considered to be a bounding condition to the 9975 package conditions when the laboratory results are extrapolated to actual package conditions, and is recommended as a conservative estimate for package performance calculations.

  2. The corrosion of copper in compacted clay

    Energy Technology Data Exchange (ETDEWEB)

    King, F.; Ryan, S.R.; Litke, C.D

    1997-12-01

    The uniform corrosion behaviour of copper has been investigated in the presence of compacted clay under simulated disposal vault conditions. The compacted clay is used to simulate the buffer material that would surround copper nuclear fuel waste containers in a Canadian disposal vault. The effect of the speciation of dissolved Cu has been investigated using three synthetic groundwaters of different salinity and various dissolved [O{sub 2}]. The formation of cuprous species is favoured by low [O{sub 2}] and high [C1{sup -}], with Cu(II) species formed at high [O{sub 2}] and low [C1{sup -}]. Because the Na-bentonite clay is a cation-exchange material, positively charged Cu(II) species are found to adsorb more strongly than negatively charged CuC1{sup -} complexes. The impact of the Cu speciation on four experimental parameters is reported: the corrosion rate, the interfacial [Cu] in the clay, the [Cu] profile through the clay layer, and the Cu(l):Cu(ll) ratio in the precipitated corrosion products. In agreement with previous studies, the overall rate-controlling process is believed to be the diffusion of dissolved Cu away from the corroding surface. Adsorption acts as a driving force for corrosion by immobilizing dissolved Cu. Under the conditions used in these experiments, the diffusion of dissolved O{sub 2} to the Cu surface was not rate controlling. (author)

  3. Replacing fuel alignment in Germany

    International Nuclear Information System (INIS)

    Poetz, F.; Kalthoff, W.

    1991-01-01

    Up to the end of 1989 varying numbers of broken fuel alignment pins were detected in several German PWRs (80 broken pins in all). The distribution of these broken pins over the core cross-section was more or less random. The problem was due to the stress corrosion cracking of the pin material and was restricted to individual pins. It was concluded that all fuel alignment pins made of Inconel X-750 should be replaced. The development of a new pin, more resistant to intergranular stress corrosion, and the replacement technique are outlined. (author)

  4. Corrosion behaviors of SMART materials in the ammonia atmosphere

    International Nuclear Information System (INIS)

    Baek, J. H.; Lee, M. H.; Choi, B. S.; Kim, J. P.; Jung, Y. H.; Lee, D. J.

    1999-01-01

    The corrosion characteristics of the zirconium-based alloy(Low-Sn Zircaloy-4) and titanium-based alloys(PT-7M and ASTM Gr.2), which would be used for fuel cladding tube and steam generator tube in the SMART, were investigated at 360 deg C, 400 deg C, 500 deg C, and 520 deg C in the ammonia atmosphere. In all test conditions, the resistance to uniform and nodular corrosion of zirconium-based alloy was inferior to that of titanium-based alloys. In the case of 360 deg C test, the corrosion rate of zirconium-based alloy decreased slightly with increasing the ammonia concentration, while that of titanium-based alloys increased. The test results above 400 deg C showed that the corrosion resistance of PT-7M alloys was superior to that of ASTM Gr.2 alloy and was not influenced from the variation of ammonia concentration

  5. Wear and impact resistance of HVOF sprayedceramic matrix composites coating

    Science.gov (United States)

    Prawara, B.; Martides, E.; Priyono, B.; Ardy, H.; Rikardo, N.

    2016-02-01

    Ceramic coating has the mechanical properties of high hardness and it is well known for application on wear resistance, but on the other hand the resistance to impact load is low. Therefore its use is limited to applications that have no impact loading. The aim of this research was to obtain ceramic-metallic composite coating which has improved impact resistance compared to conventional ceramic coating. The high impact resistance of ceramic-metallic composite coating is obtained from dispersed metallic alloy phase in ceramic matrix. Ceramic Matrix Composites (CMC) powder with chrome carbide (Cr3C2) base and ceramic-metal NiAl-Al2O3 with various particle sizes as reinforced particle was deposited on mild steel substrate with High Velocity Oxygen Fuel (HVOF) thermal spray coating. Repeated impact test showed that reinforced metallic phase size influenced impact resistance of CMC coating. The ability of CMC coating to absorb impact energy has improved eight times and ten times compared with original Cr3C2 and hard chrome plating respectively. On the other hand the high temperature corrosion resistance of CMC coating showed up to 31 cycles of heating at 800°C and water quenching cooling.

  6. The anaerobic corrosion of carbon steel in concrete

    International Nuclear Information System (INIS)

    Naish, C.C.; Balkwill, P.H.; O'Brien, T.M.; Taylor, K.J.; Marsh, G.P.

    1990-11-01

    The report describes the work of a two year programme investigating the anaerobic corrosion of carbon steel embedded in a range of candidate repository cements and concretes at laboratory ambient temperatures. The factors investigated in the study were the rate of the anaerobic corrosion reaction, the effect of hydrogen overpressure on the reaction rate and the form of the corrosion product. Both electrochemical and sample weight loss corrosion rate measurements were used. The cements and concretes used were prepared both with and without small additions of chloride (2% by weight of mix water). The results indicate that the corrosion rate is low, <1 μm/year, the effect of hydrogen overpressure is not significant over the range of pressures investigated, 1-100 atmospheres, and that the corrosion product is dependent on the cement used to cast the samples. Magnetite was identified in the case of blast furnace slag replacement cements but for pulverised fuel ash and ordinary Portland cements no corrosion product was evident either from X-ray diffraction or laser Raman measurements. (Author)

  7. Furniture Rack Corrosion Coupon Surveillance - 2012 Update

    International Nuclear Information System (INIS)

    Mickalonis, J. I.; Murphy, T. R.; Berry, C. J.

    2012-01-01

    Under the L Basin corrosion surveillance program furniture rack coupons immersed for 14 years (FY2009 coupons) and 16 years (FY2011 coupons) were analyzed and the results trended with coupons exposed for shorter times. In addition, a section harvested from an actual furniture rack that was immersed for 14 years was analyzed for pitting in the weld and heat-affected-zone (HAZ) regions. The L Basin operations maintained very good water quality over the entire immersion period for these samples. These results for FY2009 and FY2011 coupons showed that the average pit depths for the 6061 and 6063 base metal are 1 and 2 mils, respectively, while those for the weld and HAZ are 3 and 4 mils, respectively. The results for the weld and HAZ regions are similar to coupons removed during the period of FY2003 to FY2007. These similarities indicate that the pit development occurred quickly followed by slow kinetics of increase in pit depth. For the actual furniture rack sample average pits of 5 and 2 mils were measured for the HAZ and weld, respectively. These results demonstrate that pitting corrosion of the aluminum furniture racks used to support the spent fuel occurs in waters of good quality. The corrosion kinetics or pit depth growth rate is much less that 1 mil/year, and would not impact long-term use of this material system for fuel storage racks in L Basin if good water quality is maintained

  8. Furniture Rack Corrosion Coupon Surveillance - 2012 Update

    Energy Technology Data Exchange (ETDEWEB)

    Mickalonis, J. I.; Murphy, T. R.; Berry, C. J.

    2012-10-01

    Under the L Basin corrosion surveillance program furniture rack coupons immersed for 14 years (FY2009 coupons) and 16 years (FY2011 coupons) were analyzed and the results trended with coupons exposed for shorter times. In addition, a section harvested from an actual furniture rack that was immersed for 14 years was analyzed for pitting in the weld and heat-affected-zone (HAZ) regions. The L Basin operations maintained very good water quality over the entire immersion period for these samples. These results for FY2009 and FY2011 coupons showed that the average pit depths for the 6061 and 6063 base metal are 1 and 2 mils, respectively, while those for the weld and HAZ are 3 and 4 mils, respectively. The results for the weld and HAZ regions are similar to coupons removed during the period of FY2003 to FY2007. These similarities indicate that the pit development occurred quickly followed by slow kinetics of increase in pit depth. For the actual furniture rack sample average pits of 5 and 2 mils were measured for the HAZ and weld, respectively. These results demonstrate that pitting corrosion of the aluminum furniture racks used to support the spent fuel occurs in waters of good quality. The corrosion kinetics or pit depth growth rate is much less that 1 mil/year, and would not impact long-term use of this material system for fuel storage racks in L Basin if good water quality is maintained.

  9. BWR steel containment corrosion

    Energy Technology Data Exchange (ETDEWEB)

    Tan, C.P.; Bagchi, G.

    1996-04-01

    The report describes regulatory actions taken after corrosion was discovered in the drywell at the Oyster Creek Plant and in the torus at the Nine Mile Point 1 Plant. The report describes the causes of corrosion, requirements for monitoring corrosion, and measures to mitigate the corrosive environment for the two plants. The report describes the issuances of generic letters and information notices either to collect information to determine whether the problem is generic or to alert the licensees of similar plants about the existence of such a problem. Implementation of measures to enhance the containment performance under severe accident conditions is discussed. A study by Brookhaven National Laboratory (BNL) of the performance of a degraded containment under severe accident conditions is summarized. The details of the BNL study are in the appendix to the report.

  10. Corrosion Inhibitors for Aluminum.

    Science.gov (United States)

    Muller, Bodo

    1995-01-01

    Describes a simple and reliable test method used to investigate the corrosion-inhibiting effects of various chelating agents on aluminum pigments in aqueous alkaline media. The experiments that are presented require no complicated or expensive electronic equipment. (DDR)

  11. Corrosion of valve metals

    International Nuclear Information System (INIS)

    Draley, J.E.

    1976-01-01

    A general survey related to the corrosion of valve metals or film-forming metals. The way these metals corrode with some general examples is described. Valve metals form relatively perfect oxide films with little breakdown or leakage when anodized

  12. Tunneling corrosion mechanism of the hot forged austenitic stainless steel in highly oxidizing nitric acid

    International Nuclear Information System (INIS)

    Nagano, Hiroo; Kajimura, Haruhiko

    1993-01-01

    Austenitic Stainless Steels have been used for reprocessing plants where spent nuclear fuels are dealt with in hot nitric acid. Conventional stainless steels are resistant enough to nitric acid. However, they are prone to localized corrosion when nitric acid becomes highly oxidizing with birth of oxidants such as Ce 4+ or Cr 6+ ion during the reprocessing. Pitting type corrosion, so-called tunneling or end-grain corrosion occurred on the forgings of 25%-20%-Nb stainless steel (310Nb stainless steel) in such nitric acid solutions because of transpassive corrosion. It has been well known that metal surfaces of steel products casted, forged or rolled are susceptible to the tunneling corrosion in aggressive corrosion media. Nevertheless, neither clear explanations of the mechanism nor definite countermeasures have been proposed yet. This paper describes the mechanism and countermeasures on the tunneling corrosion of stainless steels in nitric acid relevant to spent nuclear fuel reprocessing. The results obtained are as follows: both general and intergranular corrosion occur on austenitic stainless steels in boiling 8N HNO 3 with Cr 6+ ions. Tunneling corrosion is initiated and propagates at the metal surfaces of 310Nb stainless steel forgings along chromium depleted areas vertical to metal flows. The grooves due to the tunneling corrosion are of diameters of 0.5 to 2 mm with a maximum depth of 6mm depending on exposure time and Cr 6+ concentration in nitric acid. Tunneling corrosion proceeds by build up of galvanic corrosion cells with Cr depleted parts as anodes and their neighborhoods as cathodes. The Cr depleted parts are formed during solidification of ingots and still retained parallel to the metal flow even after forging. The ESR (Electro Slag Remelting) is one of the useful preventive methods to tunneling corrosion from the view point of steel homogenization

  13. Study of radiation effects on zircaloy 4 microstructure (Impact on susceptibility to fuel pellet-cladding interaction in PWR)

    International Nuclear Information System (INIS)

    Lefebvre, F.

    1989-01-01

    In PWR the fast neutron flux is an important parameter for fuel can aging by modification of zircaloy-4 microstructure: amorphisation and dissolution of intermetallic precipitates. These phenomena are both analysed and their influence on fuel-cladding interaction is discussed. Irradiations by 1 MeV electrons, Ar ions, Kr ions and fast neutrons are realized for comparison of damages with different defect creation kinetics. Amorphisation is explained as the crystal amorphous state transformation allowing precipitate dissolution by creation of a chemical potential gradient between matrix and amorphous phase. Progressive dissolution of precipitates produced by irradiation decrease the number of potential sites for stress corrosion cracking, improving rupture resistance of the alloy by fuel-cladding interaction [fr

  14. 3D study of intermetallics and their effect on the corrosion morphology of rheocast aluminium alloy

    International Nuclear Information System (INIS)

    Mingo, B.; Arrabal, R.; Pardo, A.; Matykina, E.; Skeldon, P.

    2016-01-01

    In the present study, the effect of heat treatment T6.1 on the microstructure and corrosion behaviour of rheocast aluminium alloy A356 is investigated on the basis of 2D/3D characterization techniques and electrochemical and SKPFM measurements. Heat treatment strengthens the α-Al matrix, modifies the intermetallic particles and spheroidizes eutectic Si. These changes do not modify significantly the corrosion behaviour of the alloy. 3D SEM-Tomography clearly shows that the corrosion advances in the shape of narrow paths between closely spaced intermetallics without a major influence of eutectic Si. - Highlights: • T6.1 spheroidizes Si, strengthens the matrix and modifies the intermetallics. • Electrochemical behaviour of untreated and heat-treated alloys is similar. • 3D SEM-Tomography provides additional information on the corrosion morphology. • Corrosion advances as paths between intermetallics with little influence of Si.

  15. Matrix calculus

    CERN Document Server

    Bodewig, E

    1959-01-01

    Matrix Calculus, Second Revised and Enlarged Edition focuses on systematic calculation with the building blocks of a matrix and rows and columns, shunning the use of individual elements. The publication first offers information on vectors, matrices, further applications, measures of the magnitude of a matrix, and forms. The text then examines eigenvalues and exact solutions, including the characteristic equation, eigenrows, extremum properties of the eigenvalues, bounds for the eigenvalues, elementary divisors, and bounds for the determinant. The text ponders on approximate solutions, as well

  16. Corrosion of beryllium

    International Nuclear Information System (INIS)

    Mueller, J.J.; Adolphson, D.R.

    1987-01-01

    The corrosion behavior of beryllium in aqueous and elevated-temperature oxidizing environments has been extensively studied for early-intended use of beryllium in nuclear reactors and in jet and rocket propulsion systems. Since that time, beryllium has been used as a structural material in les corrosive environments. Its primary applications include gyro systems, mirror and reentry vehicle structures, and aircraft brakes. Only a small amount of information has been published that is directly related to the evaluation of beryllium for service in the less severe or normal atmospheric environments associated with these applications. Despite the lack of published data on the corrosion of beryllium in atmospheric environments, much can be deduced about its corrosion behavior from studies of aqueous corrosion and the experiences of fabricators and users in applying, handling, processing, storing, and shipping beryllium components. The methods of corrosion protection implemented to resist water and high-temperature gaseous environments provide useful information on methods that can be applied to protect beryllium for service in future long-term structural applications

  17. Monitoring corrosion rates and localised corrosion in low conductivity water

    DEFF Research Database (Denmark)

    Hilbert, Lisbeth Rischel

    2006-01-01

    Monitoring of low corrosion rates and localised corrosion in a media with low conductivity is a challenge. In municipal district heating, quality control may be improved by implementing on-line corrosion monitoring if a suitable technique can be identified to measure both uniform and localised...... corrosion. Electrochemical techniques (LPR, EIS, crevice corrosion current) as well as direct measurement techniques (high-sensitive electrical resistance, weight loss) have been applied in operating plants. Changes in the corrosion processes are best monitored in non-aggressive, low conductivity media...

  18. Corrosion behaviour of the AlSi6Cu4 alloy and cast AlSi6Cu4-graphite particles composite

    Directory of Open Access Journals (Sweden)

    S. Holecek

    2009-04-01

    Full Text Available The corrosion behaviour of the AlSi6Cu4 alloy as a composite matrix and of composites with 8% vol. of graphite particles was investigated. The corrosion experiments were performed over a range of elevated temperatures and were carried out in sea water (3.5%NaCl solution. We have focused our attention to the determination of the mode of corrosion attack and to the determination of the rate ofcorrosion and other corrosion characteristics. Both as-cast and annealed matrix and composite specimens were tested, as well as the99.9% as-cast aluminium for comparison. Corrosion behaviour of the materials was assessed by the corrosion potential (Ec and bypotentiodynamic (polarization curves. As expected, composite is less corrosion resistant than the matrix alloy. In addition to pitting,a severe galvanic corrosion occurs as a result of galvanic couple aluminium/graphite formation. Corrosion potentials imply that examinedmaterials would be sufficiently resistant in non or slightly oxidizing solutions without dissolved oxygen. All studied materials corrode very slowly at potentials negative to corrosion potential, while at potentials positive to corrosion potential the corrosion rate goes up by 1 or 2 orders.

  19. ELECTROCHEMICAL STUDIES OF URANIUM METAL CORROSION MECHANISM AND KINETICS IN WATER

    International Nuclear Information System (INIS)

    Boudanova, Natalya; Maslennikov, Alexander; Peretroukhine, Vladimir F.; Delegard, Calvin H.

    2006-01-01

    During long-term underwater storage of low burn-up uranium metal fuel, a corrosion product sludge forms containing uranium metal grains, uranium dioxide, uranates and, in some cases, uranium peroxide. Literature data on the corrosion of non-irradiated uranium metal and its alloys do not allow unequivocal prediction of the paragenesis of irradiated uranium in water. The goal of the present work conducted under the program 'CORROSION OF IRRADIATED URANIUM ALLOYS FUEL IN WATER' is to study the corrosion of uranium and uranium alloys and the paragenesis of the corrosion products during long-term underwater storage of uranium alloy fuel irradiated at the Hanford Site. The elucidation of the physico-chemical nature of the corrosion of irradiated uranium alloys in comparison with non-irradiated uranium metal and its alloys is one of the most important aspects of this work. Electrochemical methods are being used to study uranium metal corrosion mechanism and kinetics. The present part of work aims to examine and revise, where appropriate, the understanding of uranium metal corrosion mechanism and kinetics in water

  20. A multiphase interfacial model for the dissolution of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Jerden, James L., E-mail: jerden@anl.gov [Argonne National Laboratory, 9700 South Cass Ave., Argonne, IL 60439 (United States); Frey, Kurt [University of Notre Dame, Notre Dame, IN 46556 (United States); Ebert, William [Argonne National Laboratory, 9700 South Cass Ave., Argonne, IL 60439 (United States)

    2015-07-15

    Highlights: • This model accounts for chemistry, temperature, radiolysis, U(VI) minerals, and hydrogen effect. • The hydrogen effect dominates processes determining spent fuel dissolution rate. • The hydrogen effect protects uranium oxide spent fuel from oxidative dissolution. - Abstract: The Fuel Matrix Dissolution Model (FMDM) is an electrochemical reaction/diffusion model for the dissolution of spent uranium oxide fuel. The model was developed to provide radionuclide source terms for use in performance assessment calculations for various types of geologic repositories. It is based on mixed potential theory and consists of a two-phase fuel surface made up of UO{sub 2} and a noble metal bearing fission product phase in contact with groundwater. The corrosion potential at the surface of the dissolving fuel is calculated by balancing cathodic and anodic reactions occurring at the solution interfaces with UO{sub 2} and NMP surfaces. Dissolved oxygen and hydrogen peroxide generated by radiolysis of the groundwater are the major oxidizing agents that promote fuel dissolution. Several reactions occurring on noble metal alloy surfaces are electrically coupled to the UO{sub 2} and can catalyze or inhibit oxidative dissolution of the fuel. The most important of these is the oxidation of hydrogen, which counteracts the effects of oxidants (primarily H{sub 2}O{sub 2} and O{sub 2}). Inclusion of this reaction greatly decreases the oxidation of U(IV) and slows fuel dissolution significantly. In addition to radiolytic hydrogen, large quantities of hydrogen can be produced by the anoxic corrosion of steel structures within and near the fuel waste package. The model accurately predicts key experimental trends seen in literature data, the most important being the dramatic depression of the fuel dissolution rate by the presence of dissolved hydrogen at even relatively low concentrations (e.g., less than 1 mM). This hydrogen effect counteracts oxidation reactions and can limit

  1. Corrosion and anticorrosion. Industrial practice

    International Nuclear Information System (INIS)

    Beranger, G.; Mazille, H.

    2002-01-01

    This book comprises 14 chapters written with the collaboration of about 50 French experts of corrosion. It is complementary to another volume entitled 'corrosion of metals and alloys' and published by the same editor. This volume comprises two parts: part 1 presents the basic notions of corrosion phenomena, the properties of surfaces, the electrochemical properties of corrosion etc.. Part 2 describes the most frequent forms of corrosion encountered in industrial environments and corresponding to specific problems of protection: marine environment, atmospheric corrosion, galvanic corrosion, tribo-corrosion, stress corrosion etc.. The first 8 chapters (part 1) treat of the corrosion problems encountered in different industries and processes: oil and gas production, chemical industry, phosphoric acid industry, PWR-type power plants, corrosion of automobile vehicles, civil engineering and buildings, corrosion of biomaterials, non-destructive testing for the monitoring of corrosion. The other chapters (part 2) deal with anticorrosion and protective coatings and means: choice of materials, coatings and surface treatments, thick organic coatings and enamels, paints, corrosion inhibitors and cathodic protection. (J.S.)

  2. Pitting corrosion of 5052 aluminum alloy

    Science.gov (United States)

    Lockwood, F.; Lee, S.; Faunce, J.; Green, J. A. S.; Ptashnick, W. J.

    1985-01-01

    The relative degree of pitting of 5052 aluminum alloy in a prepaint cleaning process is correlated with the proportion of MgO in the surface oxide of the alloy. Magnesium oxide, formed on the surface during process heat treatments of the alloy, is soluble in the acidic environment of the particular cleaning solutions, thus allowing easier access of corrosive ion to the aluminum metal. The pitting mechanism involves the: (1) formation of microscopic, localized galvanic cells between aluminum (anode) and iron-containing constituents (cathodic sites) normally found in the matrix of 5052 (and other 5xxx aluminum alloys), (2) aggravation of pit formation by chloride, and (3) enhancement of cathodic reactions and hence the overall corrosion process due to penetration of the oxide layer by cupric and ferric ions, and dissolved O 2 present in the cleaning solutions. Pitting was eliminated by adding 0.1% NaNO 3 to the cleaning bath.

  3. Corrosion properties of K-claddings in PWR-simulating loop

    International Nuclear Information System (INIS)

    Park, Jeong Yong; Choi, Byung Kwon; Jeong, Yong Hwan; Jung, Youn Ho

    2004-01-01

    A development of more corrosion resistant Zr alloys than Zircaloy-4 has been required because the corrosion resistance of fuel claddings is one of the main limitations to the extension of fuel burn-up in PWR. A number of recent works concentrated on the development of the advanced Zr alloys such as ZIRLO and M5 have shown that Nb would be the most beneficial alloying element to improve the corrosion resistance of Zr alloys. The optimum Nb content for the superior corrosion performance was changed depending on the other alloy compositions and manufacturing processes. K-cladding, which was manufactured of the advanced Zr alloy with a high Nb content, showed an excellent corrosion resistance in autoclave corrosion test when alloying addition and heat treatment were properly controlled. However, the mechanistic understanding is not enough to predict the in-pile corrosion performance of K-cladding. The objective of this study is to investigate the corrosion properties of K-claddings in PWR-simulating loop. Microstructure and oxide characteristics of K-cladding were also investigated to develop a further mechanistic understanding on the corrosion behavior

  4. Matrix superpotentials

    Science.gov (United States)

    Nikitin, Anatoly G.; Karadzhov, Yuri

    2011-07-01

    We present a collection of matrix-valued shape invariant potentials which give rise to new exactly solvable problems of SUSY quantum mechanics. It includes all irreducible matrix superpotentials of the generic form W=kQ+\\frac{1}{k} R+P, where k is a variable parameter, Q is the unit matrix multiplied by a real-valued function of independent variable x, and P and R are the Hermitian matrices depending on x. In particular, we recover the Pron'ko-Stroganov 'matrix Coulomb potential' and all known scalar shape invariant potentials of SUSY quantum mechanics. In addition, five new shape invariant potentials are presented. Three of them admit a dual shape invariance, i.e. the related Hamiltonians can be factorized using two non-equivalent superpotentials. We find discrete spectrum and eigenvectors for the corresponding Schrödinger equations and prove that these eigenvectors are normalizable.

  5. Spent fuel. Dissolution and oxidation

    International Nuclear Information System (INIS)

    Grambow, B.

    1989-03-01

    Data from studies of the low temperature air oxidation of spent fuel were retrieved in order to provide a basis for comparison between the mechanism of oxidation in air and corrosion in water. U 3 O 7 is formed by diffusion of oxygen into the UO 2 lattice. A diffusion coefficient of oxygen in the fuel matric was calculated for 25 degree C to be in the range of 10 -23 to 10 -25 m 2 /s. The initial rates of U release from spent fuel and from UO 2 appear to be similar. The lowest rates (at 25 degree c >10 -4 g/(m 2 d)) were observed under reducing conditions. Under oxidizing conditions the rates depend mainly of the nature and concentraion of the oxidant and/or on corbonate. In contact with air, typical initial rates at room temperature were in the range between 0.001 and 0.1 g/(m 2 d). A study of apparent U solubility under oxidizing conditions was performed and it was suggested that the controlling factor is the redox potential at the UO 2 surface rather than the E h of the bulk solution. Electrochemical arguments were used to predict that at saturation, the surface potential will eventually reach a value given by the boundaries at either the U 3 O 7 /U 3 O 8 or the U 3 O 7 /schoepite stability field, and a comparison with spent fuel leach data showed that the solution concentration of uranium is close to the calculated U solubility at the U 3 O 7 /U 3 O 8 boundary. The difference in the cumulative Sr and U release was calculated from data from Studsvik laboratory. The results reveal that the rate of Sr release decreases with the square root of time under U-saturated conditions. This time dependence may be rationalized either by grain boundary diffusion or by diffusion into the fuel matrix. Hence, there seems to be a possibility of an agreement between the Sr release data, structural information and data for oxygen diffusion in UO 2 . (G.B.)

  6. Fuel performance in water storage

    International Nuclear Information System (INIS)

    Hoskins, A.P.; Scott, J.G.; Shelton-Davis, C.V.; McDannel, G.E.

    1993-11-01

    Westinghouse Idaho Nuclear Company operates the Idaho Chemical Processing Plant (ICPP) at the Idaho National Engineering Laboratory (INEL) for the Department of Energy (DOE). A variety of different types of fuels have been stored there since the 1950's prior to reprocessing for uranium recovery. In April of 1992, the DOE decided to end fuel reprocessing, changing the mission at ICPP. Fuel integrity in storage is now viewed as long term until final disposition is defined and implemented. Thus, the condition of fuel and storage equipment is being closely monitored and evaluated to ensure continued safe storage. There are four main areas of fuel storage at ICPP: an original underwater storage facility (CPP-603), a modern underwater storage facility (CPP-666), and two dry fuel storage facilities. The fuels in storage are from the US Navy, DOE (and its predecessors the Energy Research and Development Administration and the Atomic Energy Commission), and other research programs. Fuel matrices include uranium oxide, hydride, carbide, metal, and alloy fuels. In the underwater storage basins, fuels are clad with stainless steel, zirconium, and aluminum. Also included in the basin inventory is canned scrap material. The dry fuel storage contains primarily graphite and aluminum type fuels. A total of 55 different fuel types are currently stored at the Idaho Chemical Processing Plant. The corrosion resistance of the barrier material is of primary concern in evaluating the integrity of the fuel in long term water storage. The barrier material is either the fuel cladding (if not canned) or the can material

  7. Development of alternative materials for BWR fuel springs

    International Nuclear Information System (INIS)

    Uruma, Y.; Osato, T.; Yamazaki, K.

    2002-01-01

    Major sources of radioactivity introduced into reactor water of BWR were estimated fuel crud and in-core materials (especially, fuel springs). Fuel springs are used for fixation of fuel cladding tubes with spacer grid. Those are small parts (total length is only within 25 mm) and so many numbers are loaded simultaneously and then total surfaces area are calculated up to about 200 m 2 . Fuel springs are located under high radiation field and high oxidative environment. Conventional fuel spring is made of alloy-X750 which is one of nickel-based alloy and is reported to show relatively higher corrosion release rate. 58 Co and 60 Co will be released directly into reactor water from intensely radio-activated fuel springs surface and increase radioactivity concentrations in primary coolant. Corrosion release control from fuel springs is an important technical item and a development of alternative material instead of alloy-X750 for fuel spring is a key subject to achieve ultra low man-rem exposure BWR plant. In present work, alloy-X718 which started usage for PWR fuel springs and stainless steel type 316L which has many mechanical property data are picked up for alternative materials and compared their corrosion behaviors with conventional material. Corrosion experiment was conducted under vapor-water two phases flow which is simulated fuel cladding surface boiling condition. After exposure, corrosion film formed under corrosion test was analyzed in detail and corrosion film amount and corrosion release amount are estimated among three materials. (authors)

  8. Evaluation of corrosive behavior of SAE 5155 by corrosion environment

    International Nuclear Information System (INIS)

    An, Jae Pil; Park, Keyung Dong

    2005-01-01

    In this study, the influence of shot peening and corrosive condition for corrosion property was investigated on immersed in 3.5% NaCl, 10% HNO 3 + 3% HF, 6% FeCl 3 . The immersion test was performed on two kinds of specimen. The immersion periods was performed 30days. Corrosion potential, weight loss were investigated from experimental results. From test results, the effect of shot peening on the corrosion was evaluated

  9. CFD investigating the air ingress accident for a HTGR simulation of graphite corrosion oxidation

    International Nuclear Information System (INIS)

    Ferng, Y.M.; Chi, C.W.

    2012-01-01

    Highlights: ► A CFD model is proposed to investigate graphite oxidation corrosion in the HTR-10. ► A postulated air ingress accident is assumed in this paper. ► Air ingress flowrate is the predicted result, instead of the preset one. ► O 2 would react with graphite on pebble surface, causing the graphite corrosion. ► No fuel exposure is predicted to be occurred under the air ingress accident. - Abstract: Through a compressible multi-component CFD model, this paper investigates the characteristics of graphite oxidation corrosion in the HTR-10 core under the postulated accident of gas duct rupture. In this accident, air in the steam generator cavity would enter into the core after pressure equilibrium is achieved between the core and the cavity, which is also called as the air ingress accident. Oxygen in the air would react with graphite on pebble surface, subsequently resulting in oxidation corrosion and challenging fuel integrity. In this paper, characteristics of graphite oxidation corrosion during the air ingress accident can be reasonably captured, including distributions of graphite corrosion amount on the different cross-sections, time histories of local corrosion amount at the monitoring points and overall corrosion amount in the core, respectively. Based on the transient simulation results, the corrosion pattern and its corrosion rate would approach to the steady-state conditions as the accident continuously progresses. The total amount of graphite corrosion during a 3-day accident time is predicted to be about 31 kg with the predicted asymptotic corrosion rate. This predicted value is less than that from the previous work of Gao and Shi.

  10. Fuel assembly reconstitution

    International Nuclear Information System (INIS)

    Morgado, Mario M.; Oliveira, Monica G.N.; Ferreira Junior, Decio B.M.; Santos, Barbara O. dos; Santos, Jorge E. dos

    2009-01-01

    Fuel failures have been happened in Nuclear Power Plants worldwide, without lost of integrity and safety, mainly for the public, environment and power plants workers. The most common causes of these events are corrosion (CRUD), fretting and pellet cladding interaction. These failures are identified by increasing the activity of fission products, verified by chemical analyses of reactor coolant. Through these analyses, during the fourth operation cycle of Angra 2 Nuclear Power Plant, was possible to observe fuel failure indication. This indication was confirmed in the end of the cycle during the unloading of reactor core through leakage tests of fuel assembly, using the equipment called 'In Mast Sipping' and 'Box Sipping'. After confirmed, the fuel assembly reconstitution was scheduled, and happened in April, 2007, where was identified the cause and the fuel rod failure, which was substitute by dummy rods (zircaloy). The cause was fretting by 'debris'. The actions to avoid and prevent fuel assemblies failures are important. The goals of this work are to describe the methodology of fuel assembly reconstitution using the FARE (Fuel Assembly Reconstitution Equipment) system, to describe the results of this task in economic and security factors of the company and show how the fuel assembly failures are identified during operation and during the outage. (author)

  11. Corrosive components of nutshells and their chars

    Directory of Open Access Journals (Sweden)

    Karczewski Mateusz

    2016-01-01

    Full Text Available Biomass combustion stands among various technologies pointed at fossil fuels consumption decrease. Biomass can be found in very diversified sources spread more evenly across the globe, can be burned with use of traditional combustion solutions and is more CO2 neutral in combustion than their fossil fuel counterparts. On the other hand biomass has several problems with composition that despite its potential diversity. Problem of excess moisture can be already solved by material selection or by preliminary pyrolysis. The main problem concerns however biomass ash composition. Biomass ashes are more prone to have higher quantities of potentially corrosive components than their coal counterparts. The example of such constituents are alkali metals, sulphur and chlorine. Ash basic composition is also important due to various ash properties like its melting temperature and slagging or fouling tendencies. To address the problem, several indices for fast properties prediction and earlier problem identification can be appointed. This work concentrates on ash quality evaluation for potentially attractive biomass fuel from nutshell materials and their corresponding char obtained by pyrolysis in 300, 450 and 550 °C. Pistachio and hazelnut shells with their chars will be analysed for corrosive compounds and their potential influence on combustion process.

  12. Hot corrosion of low cobalt alloys

    Science.gov (United States)

    Stearns, C. A.

    1982-01-01

    The hot corrosion attack susceptibility of various alloys as a function of strategic materials content are investigated. Preliminary results were obtained for two commercial alloys, UDIMET 700 and Mar-M 247, that were modified by varying the cobalt content. For both alloys the cobalt content was reduced in steps to zero. Nickel content was increased accordingly to make up for the reduced cobalt but all other constituents were held constant. Wedge bar test samples were produced by casting. The hot corrosion test consisted of cyclically exposing samples to the high velocity flow of combustion products from an air-fuel burner fueled with jet A-1 and seeded with a sodium chloride aqueous solution. The flow velocity was Mach 0.5 and the sodium level was maintained at 0.5 ppm in terms of fuel plus air. The test cycle consisted of holding the test samples at 900 C for 1 hour followed by 3 minutes in which the sample could cool to room temperature in an ambient temperature air stream.

  13. 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)

  14. A Comparison of Corrosion Behavior of Copper and Its Alloy in Pongamia pinnata Oil at Different Conditions

    Directory of Open Access Journals (Sweden)

    Meenakshi H. N. Parameswaran

    2013-01-01

    Full Text Available Vegetable oils are promising substitutes for petrodiesel as they can be produced from numerous oil seed crops that can be cultivated anywhere and have high energy contents, exhibiting clean combustion behavior with zero CO2 emission and negligible SO2 generation. The impact of biofuel on the corrosion of various industrial metals is a challenge for using biofuel as automotive fuel. Fuel comes in contact with a wide variety of metallic materials under different temperatures, velocities, and loads thereby causing corrosion during storage and flow of fuel. Hence, the present investigation compares the corrosion rates of copper and brass in Pongamia pinnata oil (O100, 3% NaCl, and oil blend with NaCl (O99 obtained by static immersion test and using rotating cage. The corrosivity and conductivity of the test media are positively correlated. This study suggested that the corrosivity of copper is higher than brass in Pongamia pinnata oil (PO.

  15. Corrosion testing facilities in India

    International Nuclear Information System (INIS)

    Viswanathan, R.; Subramanian, Venu

    1981-01-01

    Major types of corrosion tests, establishment of specifications on corrosion testing and scope of their application in practice are briefly described. Important organizations in the world which publish specifications/standards are listed. Indian organizations which undertake corrosion testing and test facilities available at them are also listed. Finally in an appendix, a comprehensive list of specifications relevant to corrosion testing is given. It is arranged under the headings: environmental testing, humidity tests, salt spray/fog tests, immersion tests, specification corrosion phenomena, (tests) with respect to special corrosion media, (tests) with respect to specific corrosion prevention methods, and specific corrosion tests using electrical and electrochemical methods (principles). Each entry in the list furnishes information about: nature of the test, standard number, and its specific application. (M.G.B.)

  16. Corrosion inhibitors for concrete bridges.

    Science.gov (United States)

    2004-12-01

    Deicing salts and salt-water spray can cause serious corrosion problems for reinforced concrete bridge structures. : These problems can lead to costly and labor-intensive repair and even replacement of the structure. Surface applied : corrosion inhib...

  17. Modelling reinforcement corrosion in concrete

    DEFF Research Database (Denmark)

    Michel, Alexander; Geiker, Mette Rica; Stang, Henrik

    2012-01-01

    A physio-chemical model for the simulation of reinforcement corrosion in concrete struc-tures was developed. The model allows for simulation of initiation and subsequent propaga-tion of reinforcement corrosion. Corrosion is assumed to be initiated once a defined critical chloride threshold is rea......, a numerical example is pre-sented, that illustrates the formation of corrosion cells as well as propagation of corrosion in a reinforced concrete structure.......A physio-chemical model for the simulation of reinforcement corrosion in concrete struc-tures was developed. The model allows for simulation of initiation and subsequent propaga-tion of reinforcement corrosion. Corrosion is assumed to be initiated once a defined critical chloride threshold...

  18. Proceedings of the Water Reactor Fuel Performance Meeting - WRFPM / Top Fuel 2009

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2009-06-15

    SFEN, ENS, SNR, ANS, AESJ, CNS KNS, IAEA and NEA are jointly organizing the 2009 International Water Reactor Fuel Performance / TopFuel 2009 Meeting following the 2008 KNS Water Reactor Performance Meeting held during October 19-23, 2008 in Seoul, Korea. This meeting is held annually on a tri-annual rotational basis in Europe, USA and Asia. In 2009, this meeting will be held in Paris, September 6-10, 2009 in coordination with the Global 2009 Conference at the same date and place. That would lead to a common opening session, some common technical presentations, a common exhibition and common social events. The technical scope of the meeting includes all aspects of nuclear fuel from fuel rod to core design as well as manufacturing, performance in commercial and test reactors or on-going and future developments and trends. Emphasis will be placed on fuel reliability in the general context of nuclear 'Renaissance' and recycling perspective. The meeting includes selectively front and/or back end issues that impact fuel designs and performance. In this frame, the conference track devoted to 'Concepts for transportation and interim storage of spent fuels and conditioned waste' will be shared with 'GLOBAL' conference. Technical Tracks: - 1. Fuel Performance, Reliability and Operational Experience: Fuel operating experience and performance; experience with high burn-up fuels; water side corrosion; stress corrosion cracking; MOX fuel performance; post irradiation data on lead fuel assemblies; radiation effects; water chemistry and corrosion counter-measures. - 2. Transient Fuel Behaviour and Safety Related Issues: Transient fuel behavior and criteria (RIA, LOCA, ATWS, Ramp tests..). Fuel safety-related issues such as PCI (pellet cladding interaction), transient fission gas releases and cladding bursting/ballooning during transient events - Advances in fuel performance modeling and core reload methodology, small and large-scale fuel testing

  19. Corrosion-resistant metal surfaces

    Science.gov (United States)

    Sugama, Toshifumi [Wading River, NY

    2009-03-24

    The present invention relates to metal surfaces having thereon an ultrathin (e.g., less than ten nanometer thickness) corrosion-resistant film, thereby rendering the metal surfaces corrosion-resistant. The corrosion-resistant film includes an at least partially crosslinked amido-functionalized silanol component in combination with rare-earth metal oxide nanoparticles. The invention also relates to methods for producing such corrosion-resistant films.

  20. Atmospheric corrosion of mild steel

    OpenAIRE

    Morcillo, M.; de la Fuente, D.; Díaz, I.; Cano, H.

    2011-01-01

    The atmospheric corrosion of mild steel is an extensive topic that has been studied by many authors in different regions throughout the world. This compilation paper incorporates relevant publications on the subject, in particular about the nature of atmospheric corrosion products, mechanisms of atmospheric corrosion and kinetics of the atmospheric corrosion process, paying special attention to two matters upon which relatively less information has been published: a) the morphology of steel c...

  1. Atmospheric corrosion of mild steel

    OpenAIRE

    Morcillo, Manuel; Fuente, Daniel de la; Díaz, Iván; Cano, H.

    2011-01-01

    The atmospheric corrosion of mild steel is an extensive topic that has been studied by many authors in different regions throughout the world. This compilation paper incorporates relevant publications on the subject, in particular about the nature of atmospheric corrosion products, mechanisms of atmospheric corrosion and kinetics of the atmospheric corrosion process, paying special attention to two matters upon which relatively less information has been published: a) the morpholog...

  2. Matrix thermalization

    Energy Technology Data Exchange (ETDEWEB)

    Craps, Ben [Theoretische Natuurkunde, Vrije Universiteit Brussel (VUB), and International Solvay Institutes, Pleinlaan 2, B-1050 Brussels (Belgium); Evnin, Oleg [Department of Physics, Faculty of Science, Chulalongkorn University, Thanon Phayathai, Pathumwan, Bangkok 10330 (Thailand); Theoretische Natuurkunde, Vrije Universiteit Brussel (VUB), and International Solvay Institutes, Pleinlaan 2, B-1050 Brussels (Belgium); Nguyen, Kévin [Theoretische Natuurkunde, Vrije Universiteit Brussel (VUB), and International Solvay Institutes, Pleinlaan 2, B-1050 Brussels (Belgium)

    2017-02-08

    Matrix quantum mechanics offers an attractive environment for discussing gravitational holography, in which both sides of the holographic duality are well-defined. Similarly to higher-dimensional implementations of holography, collapsing shell solutions in the gravitational bulk correspond in this setting to thermalization processes in the dual quantum mechanical theory. We construct an explicit, fully nonlinear supergravity solution describing a generic collapsing dilaton shell, specify the holographic renormalization prescriptions necessary for computing the relevant boundary observables, and apply them to evaluating thermalizing two-point correlation functions in the dual matrix theory.

  3. Structurally Integrated Coatings for Wear and Corrosion

    Energy Technology Data Exchange (ETDEWEB)

    Beardsley, M. Brad; Sebright, Jason L.

    2008-11-18

    Wear and corrosion of structures cuts across industries and continues to challenge materials scientists and engineers to develop cost effective solutions. Industries typically seek mature technologies that can be implemented for production with rapid or minimal development and have little appetite for the longer-term materials research and development required to solve complex problems. The collaborative work performed in this project addressed the complexity of this problem in a multi-year program that industries would be reluctant to undertake without government partnership. This effort built upon the prior development of Advanced Abrasion Resistant Materials conduct by Caterpillar Inc. under DOE Cooperative Agreement No. DE-FC26-01NT41054. In this referenced work, coatings were developed that exhibited significant wear life improvements over standard carburized heat treated steel in abrasive wear applications. The technology used in this referenced work, arc lamp fusing of thermal spray coatings, was one of the primary technical paths in this work effort. In addition to extending the capability of the coating technology to address corrosion issues, additional competitive coating technologies were evaluated to insure that the best technology was developed to meet the goals of the program. From this, plasma transferred arc (PTA) welding was selected as the second primary technology that was investigated. Specifically, this project developed improved, cost effective surfacing materials and processes for wear and corrosion resistance in both sliding and abrasive wear applications. Materials with wear and corrosion performance improvements that are 4 to 5 times greater than heat treated steels were developed. The materials developed were based on low cost material systems utilizing ferrous substrates and stainless steel type matrix with hard particulates formed from borides and carbides. Affordability was assessed against other competing hard surfacing or coating

  4. Effect of Bi on the corrosion resistance of zirconium alloys

    International Nuclear Information System (INIS)

    Yao Meiyi; Zhou Bangxin; Li Qiang; Zhang Weipeng; Zhu Li; Zou Linghong; Zhang Jinlong; Peng Jianchao

    2014-01-01

    In order to investigate systematically the effect of Bi addition on the corrosion resistance of zirconium alloys, different zirconium-based alloys, including Zr-4 (Zr-l.5Sn-0.2Fe-0.1Cr), S5 (Zr-0.8Sn-0.35Nb-0.4Fe-0.1Cr), T5 (Zr-0.7Sn-l.0Nb-0.3Fe-0.1Cr) and Zr-1Nb, were adopted to prepare the zirconium alloys containing Bi of 0∼0.5% in mass fraction. These alloys were denoted as Zr-4 + xBi, S5 + xBi, T5 + xBi and Zr-1Nb + xBi, respectively. The corrosion behavior of these specimens was investigated by autoclave testing in lithiated water with 0.01 M LiOH or deionized water at 360 ℃/18.6 MPa and in superheated steam at 400 ℃/10.3 MPa. The microstructure of the alloys was examined by TEM and the second phase particles (SPPs) were analyzed by EDS. Microstructure observation shows that the addition of Bi promotes the precipitation of Sn as second phase particles (SPPs) because Sn is in solid solution in α-Zr matrix in Zr-4, S5 and T5 alloys. The concentration of Bi dissolved in α-Zr matrix increase with the increase of Nb in the alloys, and the excess Bi precipitates as Bi-containing SPPs. The corrosion results show that the effect of Bi addition on the corrosion behavior of different zirconium-based alloys is very complicated, depending on their compositions and corrosion conditions. In the case of higher Bi concentration in α-Zr, the zirconium alloys exhibit better corrosion resistance. However, in the case of precipitation of Bi-containing SPPs, the corrosion resistance gets worse. This indicates that the solid solution of Bi in α-Zr matrix can improve the corrosion resistance, while the precipitation of the Bi-containing SPPs is harmful to the corrosion resistance. (authors)

  5. Accelerated cyclic corrosion tests

    Directory of Open Access Journals (Sweden)

    Prošek T.

    2016-06-01

    Full Text Available Accelerated corrosion testing is indispensable for material selection, quality control and both initial and residual life time prediction for bare and painted metallic, polymeric, adhesive and other materials in atmospheric exposure conditions. The best known Neutral Salt Spray (NSS test provides unrealistic conditions and poor correlation to exposures in atmosphere. Modern cyclic accelerated corrosion tests include intermittent salt spray, wet and dry phases and eventually other technical phases. They are able to predict the material performance in service more correctly as documented on several examples. The use of NSS should thus be restricted for quality control.

  6. On the role of H2 as an inhibitor of UO2 matrix dissolution

    International Nuclear Information System (INIS)

    Merino, Juan; Gaona, Xavier; Duro, Lara; Bruno, Jordi; Martinez-Esparza, Aurora

    2007-01-01

    The study of spent fuel behaviour under disposal conditions is usually based on conservative approaches assuming oxidising conditions produced by water radiolysis at the fuel/water interface. However, the presence of H 2 from container corrosion can inhibit the dissolution of the UO 2 matrix and enhance its long-term stability. Several studies have confirmed the decrease in dissolution rates when H 2 is present in the system, although the exact mechanisms of interaction have not been fully established. This paper deals with a radiolytic modelling exercise to explore the consequences of the interaction of H 2 with radicals generated by radiolysis in the homogeneous phase. The main conclusion is that in all the modelled cases the presence of H 2 in the system leads to a decrease in matrix dissolution. The extent of the inhibition, and the threshold partial pressure for the inhibition to take place, both depend in a complex way on the chemical composition of the water and the type of radiation present in the system. (authors)

  7. Automatic identification of corrosion damage using image processing techniques

    Energy Technology Data Exchange (ETDEWEB)

    Bento, Mariana P.; Ramalho, Geraldo L.B.; Medeiros, Fatima N.S. de; Ribeiro, Elvis S. [Universidade Federal do Ceara (UFC), Fortaleza, CE (Brazil); Medeiros, Luiz C.L. [Petroleo Brasileiro S.A. (PETROBRAS), Rio de Janeiro, RJ (Brazil)

    2009-07-01

    This paper proposes a Nondestructive Evaluation (NDE) method for atmospheric corrosion detection on metallic surfaces using digital images. In this study, the uniform corrosion is characterized by texture attributes extracted from co-occurrence matrix and the Self Organizing Mapping (SOM) clustering algorithm. We present a technique for automatic inspection of oil and gas storage tanks and pipelines of petrochemical industries without disturbing their properties and performance. Experimental results are promising and encourage the possibility of using this methodology in designing trustful and robust early failure detection systems. (author)

  8. Combating corrosion in biomass and waste fired plant

    Energy Technology Data Exchange (ETDEWEB)

    Henderson, Pamela [Vattenfall AB, Stockholm (Sweden). Research and Development; Hjoernhede, Anders [Vattenfall AB, Gothenburg (Sweden). Power Consultant

    2010-07-01

    Many biomass- or waste-fired plants have problems with high temperature corrosion especially if the steam temperature is greater than 500 C. An increase in the combustion of waste fuels means that an increasing number of boilers have had problems. Therefore, there is great interest in reducing the costs associated with high temperature corrosion and at the same time there exists a desire to improve the electrical efficiency of a plant by the use of higher steam temperatures. Assuming that the fuel is well-mixed and that there is good combustion control, there are in addition a number of other measures which can be used to reduce superheater corrosion in biomass and waste fired plants, and these are described in this paper. These include the use of fuel additives, specifically sulphur-containing ones; design aspects like placing superheaters in less corrosive positions in a boiler, using tube shielding, a wider pitch between the tubes; operational considerations such as more controlled soot-blowing and the use of better materials. (orig.)

  9. The existence state of uranium(VI) in portland cement matrix material immobilization body

    International Nuclear Information System (INIS)

    Tan Hongbin; Li Yuxiang

    2005-01-01

    The basis of Portland cement material reaction with uranium, the corrosion of uranium minerals in nature and the state of study on immobilization of uranium by Portland cement matrix material are introduced, and some considerations are presented. (authors)

  10. Characterization and prediction of carbon steel corrosion in diluted seawater containing pentaborate

    Science.gov (United States)

    Fukaya, Yuichi; Watanabe, Yutaka

    2018-01-01

    This study addresses the influence of Na2B10O16, which may be used for criticality control of fuel debris in the Fukushima Daiichi Nuclear Power Station, on the corrosion behavior of carbon steel in diluted artificial seawater. The corrosion forms of carbon steel were categorized as uniform corrosion, localized corrosion, and passivity based on the balance between the dilution ratio of artificial seawater and the concentration of Na2B10O16. The changes in corrosion forms were arranged on a water quality region map. Passivity was maintained by adding 3.7 × 10-2 M or more of Na2B10O16 to artificial seawater with a dilution ratio of 100-fold or more. The criticality control of the fuel debris and corrosion mitigation of the carbon steel components may be achieved simultaneously in the water quality. The prediction of the corrosion form of carbon steel was attempted by the extended Larson-Skold Index (LSI) = ([Cl-] + 2[SO42-])/([HCO3-] + 2[B10O162-]). However, because the passivating action of B10O162- was remarkably stronger than that of HCO3-, the prediction was difficult under the simple addition of equivalent concentrations. The localized corrosion of carbon steel under the addition of Na2B10O16 preferentially occurred from the crevices of the test specimens, as was the case in stainless steel.

  11. Microstructure Analysis of Neutron Absorber Al/B4C Metal Matrix Composites

    Directory of Open Access Journals (Sweden)

    Jianxin Lin

    2017-12-01

    Full Text Available The microstructure of Al/B4C metal matrix composites (MMCs used as neutron absorbers in both dry storage casks and wet storage pools of spent nuclear fuel was analyzed by SEM and TEM. A polishing method of a focused Ga+ ion beam was used to obtain an ideal sample surface with very low roughness, which was used to statistically analyze the distribution characteristics and size factor of B4C particles in the aluminum matrix. The area of B4C particles mainly ranged from 0 to 0.5 μm2, which was the proportion of 64.29%, 86.99% and 76.86% of total statistical results for the Al-15%B4C, Al-25%B4C and Al-30%B4C MMCs, respectively. The average area of B4C particles in the Al-15%B4C, Al-25%B4C and Al-30%B4C MMCs were about 1.396, 0.528 and 1.183 μm2, respectively. The nanoscale precipitates were analyzed by the element mappings in scanning transmission electron microscopy (STEM mode and electron energy loss spectroscopy (EELS mode, which included elliptic alloy precipitates with elemental Cu, Cr, Fe and Si, except for Al, and B4C nanoparticles with polygonal shape. The interface characteristics showed that the (021 crystal plane of B4C particle and (111 crystal plane of aluminum matrix grew together. The lattice misfit was about 1.68% for (111Al//(021B4C. The corrosion properties and corresponding mechanism of Al/B4C MMCs were investigated in an aqueous solution with 5000 ppm boric acid at 100 °C and atmospheric pressure, which showed that the mass increment rate was first decreased with increasing corrosion time and then increased.

  12. 77 FR 66074 - Regulation of Fuel and Fuel Additives: Modification to Octamix Waiver (TOLAD MFA-10A)

    Science.gov (United States)

    2012-11-01

    ... Texas Methanol's gasoline-alcohol fuel, OCTAMIX. This correction notice explains that TOLAD MFA-10A is... inhibitor, TOLAD TM MFA-10A, at a concentration of 42 mg/l, in the OCTAMIX gasoline-alcohol fuel blend which... corrosion inhibitor in Texas Methanol's gasoline- alcohol fuel, OCTAMIX.\\2\\ \\1\\ EPA-HQ-OAR-2011-0894-0001...

  13. Effect of heat treatment on corrosion behavior of duplex coatings

    Directory of Open Access Journals (Sweden)

    K. Raghu Ram Mohan Reddy

    2017-01-01

    Full Text Available In the present paper, duplex WC-Co/NiCrAlY coating is coated onto Ti6Al4V substrate and vacuum heat treatment is employed to investigate the corrosion behavior of heat treated samples as well as Ti6Al4V substrate for comparison. In this duplex coating system, High Velocity Oxy Fuel (HVOF process is used to deposit NiCrAlY interlayer with a constant thickness of 200 μm and WC-Co ceramic top layer with varying thickness of 250 μm, 350 μm and 450 μm deposited by Detonation Spray (DS process. Different heat treatment temperatures (600–1150 °C were employed for the coated samples to study the microstructure and the effect on corrosion resistance of the duplex coatings. Potentiodynamic polarization tests were carried to investigate the corrosion performance of duplex coated heat treated samples and the substrate in Ringer’s solution at 37 °C and prepared the pH to 5.7. The microstructure upon corrosion after heat treatment was characterized by SEM analysis to understand the corrosion behavior. The results disclosed that at all heat treatment temperatures, all the coated samples exhibited better corrosion resistance than the base substrate. However, during 950 °C and 1150 °C heat treatment temperatures, it was observed highest corrosion potential than 600 °C and 800 °C. The 350 μm thickness, coated sample exhibited highest corrosion resistance compared to other two coated samples and the substrate at all heat treatment temperatures.

  14. Study of biofuels corrosiveness; Estudo da corrosividade de biocombustiveis

    Energy Technology Data Exchange (ETDEWEB)

    Nascimento, Rafaela da Conceicao; Branco, Lucas da Paz Nogueira; Guimaraes, Maria Jose de Oliveria Cavalcanti; Seidl, Peter Rudolf [Universidade Federal do Rio de Janeiro (UFRJ), RJ (Brazil). Escola de Quimica

    2012-07-01

    The replacement of crude oil derivatives with biofuels is part of an energy strategy, in which efficiency and conservation should play a leading role. The renewable nature of biodiesel is based in the fact that it is derived from raw materials from agricultural activities, instead of petroleum. Thinking about the increasingly production of biodiesel and its storage to make the proper mixtures, the corrosion of biodiesel tanks can occur over time. The copper corrosion test is used as the standard method for biodiesel corrosivity, however it is designed for diesel fuel from petroleum. Petrodiesel contains sulfur, which attacks the copper alloys. Since biodiesel is free of this element, it should be characterized in a different way. This work intends to analyze existing methodologies in order to evaluate the corrosivity of biodiesel. A literature search was performed in Web of Science database, selecting keywords for a comprehensive search. For the final analysis, a selection of papers classified according to publication year, country and main subject covered was used. Research on biofuels corrosivity is recent and major countries where published (USA, Malaysia and Brazil) which have advantages in terms of size and production of oil seeds. Biodiesel is more corrosive than petroleum diesel, but there is no sufficient evidences as to verify if the level of corrosion found in biodiesel is within acceptable limits for automotive components. Ferrous alloys are more resistant to biodiesel attack than non-ferrous alloys. Among non-ferrous alloys, copper and lead alloys are the most vulnerable, followed by aluminum. It is necessary to develop appropriate methodologies to determine and characterize the corrosivity of biodiesel on engine parts and alloys used in its manufacture and storage. (author)

  15. Water chemistry and corrosion control of cladding and primary circuit components. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    1999-12-01

    Corrosion is the principal life limiting degradation mechanism in nuclear steam supply systems, especially taking into account the trends to increase fuel burnup, thermal rate and cycle length. Primary circuit components of water cooled power reactors have an impact on Zr-based alloys behaviour due to crud (primary circuit corrosion products) formation, transport and deposition on heat transfer surfaces. Crud deposits influence water chemistry, radiation and thermal hydraulic conditions near cladding surface, and by this way-Zr-based alloy corrosion. During the last decade, significant improvements were achieved in the reduction of the corrosion and dose rates by changing the cladding material for one more resistant to corrosion or by the improvement of water chemistry conditions. However, taking into account the above mentioned tendency for heavier fuel duties, corrosion and water chemistry, control will remain a serious task to work with for nuclear power plant operators and scientists, as well as development of generally accepted corrosion model of Zr-based alloys in a water environment in a new millennium. Upon the recommendation of the International Working Group on Water Reactor Fuel Performance and Technology, water chemistry and corrosion of cladding and primary circuit components are in the focus of the IAEA activities in the area of fuel technology and performance. At present the IAEA performs two co-ordinated research projects (CRPs): on On-line High Temperature Monitoring of Water Chemistry and Corrosion (WACOL) and on Activity Transport in Primary Circuits. Two CRPs deal with hydrogen and hydride degradation of the Zr-based alloys. A state-of-the-art review entitled: 'Waterside Corrosion of Zirconium Alloys in Nuclear Power Plants' was published in 1998. Technical Committee meetings on the subject were held in 1985 (Cadarache, France), 1989 (Portland, USA), 1993 (Rez, Czech Republic). During the last few years extensive exchange of experience in

  16. Corrosion of bio implants

    Indian Academy of Sciences (India)

    Chemical stability, mechanical behaviour and biocompatibility in body fluids and tissues are the basic requirements for successful application of implant materials in bone fractures and replacements. Corrosion is one of the major processes affecting the life and service of orthopaedic devices made of metals and alloys used ...

  17. Corrosion resistant steel

    International Nuclear Information System (INIS)

    Zubchenko, A.S.; Borisov, V.P.; Latyshev, V.B.

    1980-01-01

    Corrosion resistant steel for production of sheets and tubes containing C, Mn, Cr, Si, Fe is suggested. It is alloyed with vanadium and cerium for improving tensile properties and ductility. The steel can be melted by a conventional method in electric-arc or induction furnaces. The mentioned steel is intended to be used as a substitute for nickel-bearing austenitic steels

  18. Corrosion in seawater systems

    International Nuclear Information System (INIS)

    Henrikson, S.

    1988-01-01

    Highly alloyed stainless steels have been exposed to natural chlorinated and chlorine-free seawater at 35 deg. C. Simulated tube-tubesheet joints, weld joints and galvanic couples with titanium, 90/10 CuNi and NiAl bronze were tested and evaluated for corrosion. The corrosion rates of various anode materials - zinc, aluminium and soft iron - were also determined. Finally the risk of hydrogen embrittlement of tubes of ferritic stainless steels and titanium as a consequence of cathodic protection was studied. An attempt was also made to explain the cracking mechanism of the ferritic steels by means of transmission electron microscopy. One important conclusion of the project is that chlorinated seawater is considerably more corrosive to stainless steels than chlorine-free water, whereas chlorination reduces the rate of galvanic corrosion of copper materials coupled to stainless steels. Hydrogen embrittlement of ferritic stainless steels and titanium as a consequence of cathodic protection of carbon steel or cast iron in the same structure can be avoided by strict potentiostatic control of the applied potential. (author)

  19. The Reciprocal Pascal Matrix

    OpenAIRE

    Richardson, Thomas M.

    2014-01-01

    The reciprocal Pascal matrix is the Hadamard inverse of the symmetric Pascal matrix. We show that the ordinary matrix inverse of the reciprocal Pascal matrix has integer elements. The proof uses two factorizations of the matrix of super Catalan numbers.

  20. Smart Coatings for Corrosion Protection

    Science.gov (United States)

    Calle, Luz Marina; Li, Wendy; Buhrow, Jerry W.; Johnsey, Marissa N.

    2016-01-01

    Nearly all metals and their alloys are subject to corrosion that causes them to lose their structural integrity or other critical functionality. It is essential to detect corrosion when it occurs, and preferably at its early stage, so that action can be taken to avoid structural damage or loss of function. Protective coatings are the most commonly used method of corrosion control. However, progressively stricter environmental regulations have resulted in the ban of many commercially available corrosion protective coatings due to the harmful effects of their solvents or corrosion inhibitors. This work concerns the development of a multifunctional, smart coating for the autonomous control of corrosion. This coating is being developed to have the inherent ability to detect the chemical changes associated with the onset of corrosion and respond autonomously to indicate it and control it.

  1. Water chemistry and fuel performance

    International Nuclear Information System (INIS)

    Loner, H.; Blaser, W.; Zwicky, H.-U.

    1998-01-01

    A data base in water chemistry and fuel performance has been built up over 13 years of operation in the Leibstadt Nuclear Power Plant. Changes in the water chemistry are reflected in the data base. Measured crud build-up was compared with estimated values from mass calculations. Fuel cladding problems such as enhanced spacer shadow corrosion (ESSC) occurred after the last major change in the water chemistry. The affected fuel is SVEA-96 with Atoms LK2 type Zircaloy-2 as cladding material and Inconel 750 as spacer material. Changes in the fuel crud composition and different modifications at Zircaloy-2 on ESSC is shown. The water chemistry has now coordinated iron to zinc in the feed water which are thought to be the key parameters in the ESSC. However, cladding modifications (LK2 + ) may also contribute to an improvement. Recently the corrosion behaviour of fuel from different fuel suppliers is observed carefully. NPP Leibstadt will run fuel elements from three different suppliers in the following cycle. After a detailed surveillance program the most suitable fuel will be chosen

  2. 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)

  3. Mobile evaporator corrosion test results

    International Nuclear Information System (INIS)

    Rozeveld, A.; Chamberlain, D.B.

    1997-05-01

    Laboratory corrosion tests were conducted on eight candidates to select a durable and cost-effective alloy for use in mobile evaporators to process radioactive waste solutions. Based on an extensive literature survey of corrosion data, three stainless steel alloys (304L, 316L, AL-6XN), four nickel-based alloys (825, 625, 690, G-30), and titanium were selected for testing. The corrosion tests included vapor phase, liquid junction (interface), liquid immersion, and crevice corrosion tests on plain and welded samples of candidate materials. Tests were conducted at 80 degrees C for 45 days in two different test solutions: a nitric acid solution. to simulate evaporator conditions during the processing of the cesium ion-exchange eluant and a highly alkaline sodium hydroxide solution to simulate the composition of Tank 241-AW-101 during evaporation. All of the alloys exhibited excellent corrosion resistance in the alkaline test solution. Corrosion rates were very low and localized corrosion was not observed. Results from the nitric acid tests showed that only 316L stainless steel did not meet our performance criteria. The 316L welded interface and crevice specimens had rates of 22.2 mpy and 21.8 mpy, respectively, which exceeds the maximum corrosion rate of 20 mpy. The other welded samples had about the same corrosion resistance as the plain samples. None of the welded samples showed preferential weld or heat-affected zone (HAZ) attack. Vapor corrosion was negligible for all alloys. All of the alloys except 316L exhibited either open-quotes satisfactoryclose quotes (2-20 mpy) or open-quotes excellentclose quotes (<2 mpy) corrosion resistance as defined by National Association of Corrosion Engineers. However, many of the alloys experienced intergranular corrosion in the nitric acid test solution, which could indicate a susceptibility to stress corrosion cracking (SCC) in this environment

  4. Corrosion behavior of Al6061 alloy weldment produced by friction stir welding process

    Directory of Open Access Journals (Sweden)

    Farhad Gharavi

    2015-07-01

    Full Text Available In this work, the corrosion behavior of welded lap joints of AA6061-T6 aluminum alloy produced by friction stir welding process has been investigated. Corrosion properties of welded lap joints were studied by cyclic polarization and electrochemical impedance spectroscopy tests. All tests were performed in an aerated 0.6 mol L−1 NaCl aqueous solution with pH = 6.5 at a temperature of 30 °C to characterize corrosion morphology and realize corrosion features of weld regions as opposed to the parent alloy. The microstructure of weld nugget (WN, heated affected zone (HAZ, and parent alloy were analyzed using scanning electron microscopy and energy dispersive spectroscopy. The experimental results indicated that the welding process has a major effect on the corrosion resistance, which possibly associated to the break-down and dissolution of intermetallic particles. It is supposed that an increasing in intermetallic distributed throughout the matrix of weld regions increases the galvanic corrosion couples. Furthermore, by decreasing the grain size in the weld regions, the susceptibility to corrosion is enhanced. The pitting corrosion and intergranular attack are the dominant corrosion types in the weld regions and the parent alloy.

  5. Fundamental aspects of corrosion on zirconium base alloys in water reactor environments

    International Nuclear Information System (INIS)

    1990-09-01

    The purpose of this meeting was to discuss the state of knowledge of zirconium alloy corrosion mechanisms. Forty-five participants from 16 countries attended the meeting, and 25 papers were presented and discussed. One additional paper was provided only in written form. The papers were presented in seven sub-sessions under the following headings: Electrochemistry, Coolant Chemistry Effects, Irradiation Effects, Characteristics of Zirconium Oxide, Effects of Alloying on Corrosion, Corrosion Modeling and Effect of Zirconium Base Metal Properties on Corrosion. There is still a need for a laboratory corrosion test that reliably predicts in-pile corrosion in BWR's and PWR's. This holds particularly if out-of-pile tests are used for developing new Zr base alloy compositions. The role of the precipitates and of the solute elements in the matrix has still to be clarified. As it appears, a combination of both influences is necessary to explain the mechanistic aspect of the corrosion of Zircaloy. It is clear that mechanistic understanding of zirconium alloy corrosion is still some way off, although a significant amount of progress has been made toward experimental determination of the micro-scale phenomena. The papers presented a status report of our knowledge of these corrosion mechanisms, but they also served to illustrate the fact that much of the work done to date has been phenomenological rather than mechanistic. The summaries of individual sessions detail the specific conclusions and recommendations made at the meeting. Refs, figs and tabs

  6. Matrix inequalities

    CERN Document Server

    Zhan, Xingzhi

    2002-01-01

    The main purpose of this monograph is to report on recent developments in the field of matrix inequalities, with emphasis on useful techniques and ingenious ideas. Among other results this book contains the affirmative solutions of eight conjectures. Many theorems unify or sharpen previous inequalities. The author's aim is to streamline the ideas in the literature. The book can be read by research workers, graduate students and advanced undergraduates.

  7. Method for storing spent nuclear fuel in repositories

    Science.gov (United States)

    Schweitzer, D.G.; Sastre, C.; Winsche, W.

    A method for storing radioactive spent fuel in repositories containing sulfur as the storage medium is disclosed. Sulfur is non-corrosive and not subject to radiation damage. Thus, storage periods of up to 100 years are possible.

  8. Method for storing spent nuclear fuel in repositories

    Science.gov (United States)

    Schweitzer, Donald G.; Sastre, Cesar; Winsche, Warren

    1981-01-01

    A method for storing radioactive spent fuel in repositories containing sulfur as the storage medium is disclosed. Sulfur is non-corrosive and not subject to radiation damage. Thus, storage periods of up to 100 years are possible.

  9. Reduction of uranium in disposal conditions of spent nuclear fuel

    International Nuclear Information System (INIS)

    Myllykylae, E.

    2008-02-01

    This literature study is a summary of publications, in which the reduction of uranium by iron has been investigated in anaerobic groundwater conditions or in aqueous solution in general. The basics of the reduction phenomena and the oxidation states, complexes and solubilities of uranium and iron in groundwaters are discussed as an introduction to the subject, as well as, the Finnish disposal concept of spent nuclear fuel. The spent fuel itself mainly (∼96 %) consists of a sparingly soluble uranium(IV) dioxide, UO 2 (s), which is stable phase in the anticipated reducing disposal conditions. If spent fuel gets in contact with groundwater, oxidizing conditions might be induced by the radiolysis of water, or by the intrusion of oxidizing glacial melting water. Under these conditions, the oxidation and dissolution of uranium dioxide to more soluble U(VI) species could occur. This could lead to the mobilization of uranium and other components of spent fuel matrix including fission products and transuranium elements. The reduction of uranium back to oxidation state U(IV) can be considered as a favourable immobilization mechanism in a long-term, leading to precipitation due to the low solubility of U(IV) species. The cast iron insert of the disposal canister and its anaerobic corrosion products are the most important reductants under disposal conditions, but dissolved ferrous iron may also function as reductant. Other iron sources in the buffer or near-field rock, are also considered as possible reductants. The reduction of uranium is a very challenging phenomenon to investigate. The experimental studies need e.g. well-controlled anoxic conditions and measurements of oxidation states. Reduction and other simultaneous phenomena are difficult to distinghuish. The groundwater conditions (pH, Eh and ions) influence on the prevailing complexes of U and Fe and on forming corrosion products of iron and, thus they determine also the redox chemistry. The partial reduction of

  10. Investigation on a "tentacle-like" corrosion feature on Bronze Age tin-bronze objects

    Science.gov (United States)

    Piccardo, Paolo; Mödlinger, Marianne; Ghiara, Giorgia; Campodonico, Serena; Bongiorno, Valeria

    2013-12-01

    Studying the micro-structure of Austrian, Bosnian and Croatian Bronze Age objects made of tin bronze, a rare kind of corrosion feature, called in the following "tentacle-like" according to its specific way of penetrating the metallic matrix, was noted and investigated. Differing from the more classical intergranular, pitting, or crevice corrosion features, the "tentacle-like" corrosion is not following the grain boundaries, nor precisely positioned under the etching area, but penetrates mainly the crystal matrix without any apparent order. This paper discusses the first results achieved and the following hypotheses formulated in respect of the typology of this corrosion. The analyses were carried out by optical microscopy, Raman microspectroscopy, and scanning electron microscopy equipped with EDX spectroscopy for quantitative analyses.

  11. Chemical Engineering Division Fuel Cycle Programs. Quarterly progress report, October--December 1977

    Energy Technology Data Exchange (ETDEWEB)

    Steindler, M. J.; Ader, M.; Barletta, R. E.

    1978-01-01

    Fuel cycle studies reported for this period include studies of advanced solvent extraction techniques focussed on the development of centrifugal contactors for use in Purex processes. Miniature single-stage and eight-stage centrifugal contactors are being employed in performance studies applicable to larger units. In other work, literature on the dispersion of reagents as a result of explosions is being reviewed to develop systematic data applicable to fuel reprocessing and useful in identifying source terms. In yet other work, scouting studies were performed to obtain criteria for identifying organic solutions suitable for the separation of actinides from fission products. A program has been initiated on pyrochemical and dry processing of nuclear fuel. Literature reviews have been initiated on material development, carbide fuel reprocessing, and thorium-uranium reprocessing in fused salts. A review and evaluation of the encapsulation of high-level waste in a metal matrix is under way. Corrosion and leach rates of simulated waste forms are being measured and a model has been proposed to describe the reaction between solidified high-level waste and metals. In other work, criteria for the handling of fuel assembly hulls are being developed on the basis of past work on the pyrophoricity of zirconium alloys and related criteria from several sources. Experimental work is underway to determine whether nuclear wastes can be safely confined in geologic formations. Information is being obtained on the migration of radionuclides in aqueous solution-rock systems. 17 figures, 27 tables.

  12. Fuel Fabrication and Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-02-02

    The uranium from the enrichment plant is still in the form of UF6. UF6 is not suitable for use in a reactor due to its highly corrosive chemistry as well as its phase diagram. UF6 is converted into UO2 fuel pellets, which are in turn placed in fuel rods and assemblies. Reactor designs are variable in moderators, coolants, fuel, performance etc.The dream of energy ‘too-cheap to meter’ is no more, and now the nuclear power industry is pushing ahead with advanced reactor designs.

  13. Fuel Fabrication and Nuclear Reactors

    International Nuclear Information System (INIS)

    Karpius, Peter Joseph

    2017-01-01

    The uranium from the enrichment plant is still in the form of UF 6 . UF 6 is not suitable for use in a reactor due to its highly corrosive chemistry as well as its phase diagram. UF 6 is converted into UO 2 fuel pellets, which are in turn placed in fuel rods and assemblies. Reactor designs are variable in moderators, coolants, fuel, performance etc.The dream of energy ''too-cheap to meter'' is no more, and now the nuclear power industry is pushing ahead with advanced reactor designs.

  14. Corrosion research from the practical view - 1987

    International Nuclear Information System (INIS)

    Behrens, D.; Rahmel, A.; Baselt, J.P.

    1987-09-01

    This volume contains brief descriptions of R+D projects in corrosion research, including those just terminated, those still going on, and those due to begin soon. On the whole, the brief accounts cover 133 individual projects. The topics of the nine project groups are: Hydrogen-induced material damage; stress-cracking corrosion; fatigue cracking corrosion and local corrosion; fluid-flow-induced corrosion; high-temperature corrosion; material behaviour in waters and soils; corrosion in special media; material behaviour in seawater; corrosion protection by means of coatings and coverings; testing procedures for detecting material damage due to corrosion. (orig./MM) [de

  15. SNF Interim Storage Canister Corrosion and Surface Environment Investigations

    Energy Technology Data Exchange (ETDEWEB)

    Bryan, Charles R. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Enos, David G. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States)

    2015-09-01

    This progress report describes work being done at Sandia National Laboratories (SNL) to assess the localized corrosion performance of container/cask materials used in the interim storage of spent nuclear fuel (SNF). Of particular concern is stress corrosion cracking (SCC), by which a through-wall crack could potentially form in a canister outer wall over time intervals that are shorter than possible dry storage times. In order for SCC to occur, three criteria must be met. A corrosive environment must be present on the canister surface, the metal must susceptible to SCC, and sufficient tensile stress to support SCC must be present through the entire thickness of the canister wall. SNL is currently evaluating the potential for each of these criteria to be met.

  16. Influence of seawater-component on corrosion of steel in nitric acid solution containing vanadium and ruthenium

    International Nuclear Information System (INIS)

    Irisawa, Eriko; Ueno, Fumiyoshi; Uchida, Naoki; Taguchi, Katsuya

    2015-01-01

    The corrosion rate of stainless steel in boiling-nitiric-acid solutions containing oxidizing metallic ions (vanadium and ruthenium) and seawater components were investigated by the immersion tests in order to evaluate the influence of the seawater components on the corrosion of steel in the nitric acid solution containing oxidizing metallic ions. The corrosion rates of 310Nb in the solution containing seawater components were lower than that in the solution without seawater components. These results described that contamination of seawater components in the spent fuel reprocessing fluid seems not to accelerate the corrosion of steels, which are material for the devices treating the nitric-acid reprocessing solutions. (author)

  17. Development of corrosion resistant surfaces via friction stir processing for bio implant applications

    Science.gov (United States)

    Pal Singh Sodhi, Gurvinder; Singh, Harpreet

    2018-01-01

    Current investigation presents the use of Friction Stir Processing (FSP) to improve the corrosion resistance of pure magnesium for biomedical applications. FSP has been used to incorporate hydroxyapetite (HAP) into Mg-surface so as to modify the chemical composition. FSP was done within a matrix of different parameters and conditions. Influence of various parameters on microstructure was also clearly observed. XRD analysis confirmed the presence of HAP, whereas SEM images revealed a uniform distribution of the imbedded phase. Microhardness and in-vitro corrosion studies were also performed. Influence of grain size on hardness was validated by Hall-Petch relationship. Corrosion behavior was explained on the basis of texture, which indicated better corrosion resistance in comparison to the pure Mg. Therefore, the study reveals that the proposed FSP methodology can be useful tool to improve mechanical and corrosion properties of pure Mg for biomedical applications.

  18. Interfacial valence electron localization and the corrosion resistance of Al-SiC nanocomposite

    Science.gov (United States)

    Mosleh-Shirazi, Sareh; Hua, Guomin; Akhlaghi, Farshad; Yan, Xianguo; Li, Dongyang

    2015-12-01

    Microstructural inhomogeneity generally deteriorates the corrosion resistance of materials due to the galvanic effect and interfacial issues. However, the situation may change for nanostructured materials. This article reports our studies on the corrosion behavior of SiC nanoparticle-reinforced Al6061 matrix composite. It was observed that the corrosion resistance of Al6061 increased when SiC nanoparticles were added. Overall electron work function (EWF) of the Al-SiC nanocomposite increased, along with an increase in the corrosion potential. The electron localization function of the Al-SiC nanocomposite was calculated and the results revealed that valence electrons were localized in the region of SiC-Al interface, resulting in an increase in the overall work function and thus building a higher barrier to hinder electrons in the nano-composite to participate in corrosion reactions.

  19. Corrosion Evaluation and Corrosion Control of Steam Generators

    Energy Technology Data Exchange (ETDEWEB)

    Maeng, W. Y.; Kim, U. C.; Sung, K. W.; Na, J. W.; Lee, Y. H.; Lee, D. H.; Kim, K. M

    2008-06-15

    Corrosion damage significantly influences the integrity and efficiency of steam generator. Corrosion problems of steam generator are unsolved issues until now even though much effort is made around world. Especially the stress corrosion cracking of heat exchange materials is the first issue to be solved. The corrosion protection method of steam generator is important and urgent for the guarantee of nuclear plant's integrity. The objectives of this study are 1) to evaluate the corrosion properties of steam generator materials, 2) to optimize the water chemistry of steam generator and 3) to develop the corrosion protection method of primary and secondary sides of steam generator. The results will be reflected to the water chemistry guideline for improving the integrity and efficiency of steam generator in domestic power plants.

  20. Corrosion Evaluation and Corrosion Control of Steam Generators

    International Nuclear Information System (INIS)

    Maeng, W. Y.; Kim, U. C.; Sung, K. W.; Na, J. W.; Lee, Y. H.; Lee, D. H.; Kim, K. M.

    2008-06-01

    Corrosion damage significantly influences the integrity and efficiency of steam generator. Corrosion problems of steam generator are unsolved issues until now even though much effort is made around world. Especially the stress corrosion cracking of heat exchange materials is the first issue to be solved. The corrosion protection method of steam generator is important and urgent for the guarantee of nuclear plant's integrity. The objectives of this study are 1) to evaluate the corrosion properties of steam generator materials, 2) to optimize the water chemistry of steam generator and 3) to develop the corrosion protection method of primary and secondary sides of steam generator. The results will be reflected to the water chemistry guideline for improving the integrity and efficiency of steam generator in domestic power plants

  1. Scanning reference electrode techniques in localized corrosion

    International Nuclear Information System (INIS)

    Isaacs, H.S.; Vyas, B.

    1979-04-01

    The principles, advantages, and implementations of scanning reference electrode techniques are reviewed. Data related to pitting, intergranular corrosion, welds and stress corrosion cracking are presented. The technique locates the position of localized corrosion and can be used to monitor the development of corrosion and changes in the corrosion rate under a wide range of conditions

  2. Corrosion control for low-cost reliability

    International Nuclear Information System (INIS)

    Anon.

    1993-01-01

    This conference was held September 19-24, 1993 in Houston, Texas to provide a forum for exchange of state-of-the-art information on corrosion. Topics of interest focus on the following: atmospheric corrosion; chemical process industry corrosion; high temperature corrosion; and corrosion of plant materials. Individual papers have been processed separately for inclusion in the appropriate data bases

  3. Corrosion of dissimilar metal crevices in simulated concentrated ground water solutions at elevated temperature

    Energy Technology Data Exchange (ETDEWEB)

    Ikeda, B.M.; Quinn, M.J

    2003-01-01

    The disposal of high-level nuclear waste in the Yucca Mountain, Nevada is under consideration by the US Department of Energy. The proposed facility will be located in the unsaturated zone approximately 300 m below the surface and 300 m above the water table. The proposed waste container consists of an outer corrosion-resistant Alloy 22 shell surrounding a 316 NG stainless steel structural inner container that encapsulates the used nuclear fuel waste. A titanium drip shield is proposed to protect the waste container from ground water seepage arid rock-fail. A cycle of dripping/evaporation could result in the generation of concentrated aggressive solutions, which could contact the waste container. The waste container material could be susceptible to crevice corrosion from such solutions. The experiments described in this report support the modeling of waste package degradation processes. The intent was to provide parameter values that are required to model crevice corrosion chemistry, as it relates to hydrogen pick-up, and stress corrosion cracking for selected candidate waste package materials. The purpose of the experiments was to study the crevice corrosion behavior of various candidate materials under near freely corroding conditions and to determine the pH developed in crevice solutions. Experimental results of crevice corrosion of dissimilar metal pairs (Alloy 22, Grade-7 and -16 titanium and 316 stainless steel) immersed in a simulated concentrated ground water at {approx}90{sup o}C are reported. The corrosion potential was measured during exposure periods of between 330 and 630 h. Following the experiments, the pH of the crevice solution was measured. The results indicate that a limited degree of crevice acidification occurred during the experiment. The values for corrosion potential suggest that crevice corrosion may have initiated. The total corrosion was limited, with little visible evidence for crevice corrosion being observed on the sample coupon faces

  4. Coal-fueled diesel technology development -- Fuel injection equipment for coal-fueled diesel engines

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, R.N.; Hayden, H.L.

    1994-01-01

    Because of the abrasive and corrosive nature of coal water slurries, the development of coal-fueled diesel engine technology by GE-Transportation Systems (GE-TS) required special fuel injection equipment. GE-Corporate Research and Development (GE-CRD) undertook the design and development of fuel injectors, piston pumps, and check valves for this project. Components were tested at GE-CRD on a simulated engine cylinder, which included a cam-actuated jerk pump, prior to delivery to GE-TS for engine testing.

  5. pH-Sensitive Microparticles with Matrix-Dispersed Active Agent

    Science.gov (United States)

    Li, Wenyan (Inventor); Buhrow, Jerry W. (Inventor); Jolley, Scott T. (Inventor); Calle, Luz M. (Inventor)

    2014-01-01

    Methods to produce pH-sensitive microparticles that have an active agent dispersed in a polymer matrix have certain advantages over microcapsules with an active agent encapsulated in an interior compartment/core inside of a polymer wall. The current invention relates to pH-sensitive microparticles that have a corrosion-detecting or corrosion-inhibiting active agent or active agents dispersed within a polymer matrix of the microparticles. The pH-sensitive microparticles can be used in various coating compositions on metal objects for corrosion detecting and/or inhibiting.

  6. Review of oxidation rates of DOE spent nuclear fuel : Part 1 : nuclear fuel

    International Nuclear Information System (INIS)

    Hilton, B.A.

    2000-01-01

    The long-term performance of Department of Energy (DOE) spent nuclear fuel (SNF) in a mined geologic disposal system depends highly on fuel oxidation and subsequent radionuclide release. The oxidation rates of nuclear fuels are reviewed in this two-volume report to provide a baseline for comparison with release rate data and technical rationale for predicting general corrosion behavior of DOE SNF. The oxidation rates of nuclear fuels in the DOE SNF inventory were organized according to metallic, Part 1, and non-metallic, Part 2, spent nuclear fuels. This Part 1 of the report reviews the oxidation behavior of three fuel types prototypic of metallic fuel in the DOE SNF inventory: uranium metal, uranium alloys and aluminum-based dispersion fuels. The oxidation rates of these fuels were evaluated in oxygen, water vapor, and water. The water data were limited to pure water corrosion as this represents baseline corrosion kinetics. Since the oxidation processes and kinetics discussed in this report are limited to pure water, they are not directly applicable to corrosion rates of SNF in water chemistry that is significantly different (such as may occur in the repository). Linear kinetics adequately described the oxidation rates of metallic fuels in long-term corrosion. Temperature dependent oxidation rates were determined by linear regression analysis of the literature data. As expected the reaction rates of metallic fuels dramatically increase with temperature. The uranium metal and metal alloys have stronger temperature dependence than the aluminum dispersion fuels. The uranium metal/water reaction exhibited the highest oxidation rate of the metallic fuel types and environments that were reviewed. Consequently, the corrosion properties of all DOE SNF may be conservatively modeled as uranium metal, which is representative of spent N-Reactor fuel. The reaction rate in anoxic, saturated water vapor was essentially the same as the water reaction rate. The long-term intrinsic

  7. Literature Survey of Copper Corrosion Modelling under Deep Geological Disposal Environment

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Heui-Joo; Kook, Dong-Hak; Lee, Min-Soo; Choi, Heui-Joo

    2008-03-15

    As a one of the solution for the spent nuclear fuel problem, deep repository research have been under way in KAERI. To raise the resistance against repository cask corrosion, copper canister was adopted for the outer material of the cask. Duration of the repository with million year order makes the copper corrosion research under the repository environment very important and the corrosion modelling which could trace the real world precisely and predict the corrosion behavior very well is subsequently necessary. With in progress of plan to manufacture unique copper cask corresponding to our country repository system, survey for the preceding research papers in recent research direction and kernel points is expected to significant. This paper arranged the representative literatures for the important corrosion mechanism which are recently published.

  8. Corrosion of nickel in molten LiCl-Li 2O at 750 °C

    Science.gov (United States)

    Liu, R. Y.; Wang, X.; Zhang, J. S.; Wang, X. M.

    2004-05-01

    Corrosion of the reduction vessel induced by molten LiCl-Li 2O is an important problem in the lithium reduction technique for the spent nuclear fuel management. This study investigates the corrosion of nickel in molten LiCl-10 wt% Li 2O at 750 °C by immersion experiments, with the aim of unraveling the corrosion behavior. Nickel corrodes very fast in the melt, forming a layer of NiO, which is converted to Li 2Ni 8O 10 and then to LiNiO 2. The weight loss curve shows a linear rate law, which is owing to the harmful reaction between oxides and melt at the melt/scale boundary with the formation of the porous corrosion scale. The corrosion mechanism of nickel is also discussed.

  9. FP corrosion dependence on carbon and chromium content in Fe-Cr steel

    International Nuclear Information System (INIS)

    Sasaki, Koei; Tanigaki, Takanori; Fukumoto, Ken-ichi; Uno, Masayoshi

    2015-01-01

    In an attempt to investigate Cs or Cs-Te corrosion dependence on chromium or carbon content in Fe-Cr steel, cesium and Cs-Te corrosion test were performed to three specimens, Fe-9Cr-0C, Fe-9Cr-0.14C and Fe-13Cr-0.14C, for 100 hours at 973K in simulated high burn-up fuel pin environment. Cesium corrosion depth has no dependence on chromium or carbon content in Fe-Cr steel. Cs-Te corrosion was appeared in only Fe-13Cr-0.14C which has chromium carbides ranged along grain boundary. Appearance of the Cs-Te corrosion was determined by distribution or arrangement of chromium carbides which depends on chromium and carbon content. (author)

  10. FY17 Status Report: Research on Stress Corrosion Cracking of SNF Interim Storage Canisters.

    Energy Technology Data Exchange (ETDEWEB)

    Schindelholz, Eric John [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Bryan, Charles R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Alexander, Christopher L. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2017-09-01

    This progress report describes work done in FY17 at Sandia National Laboratories (SNL) to assess the localized corrosion performance of container/cask materials used in the interim storage of spent nuclear fuel (SNF). Of particular concern is stress corrosion cracking (SCC), by which a through-wall crack could potentially form in a canister outer wall over time intervals that are shorter than possible dry storage times. Work in FY17 refined our understanding of the chemical and physical environment on canister surfaces, and evaluated the relationship between chemical and physical environment and the form and extent of corrosion that occurs. The SNL corrosion work focused predominantly on pitting corrosion, a necessary precursor for SCC, and process of pit-to-crack transition; it has been carried out in collaboration with university partners. SNL is collaborating with several university partners to investigate SCC crack growth experimentally, providing guidance for design and interpretation of experiments.

  11. States and transport of hydrogen in the corrosion process of an AZ91 magnesium alloy in aqueous solution

    International Nuclear Information System (INIS)

    Chen Jian; Wang Jianqiu; Han Enhou; Dong Junhua; Ke Wei

    2008-01-01

    Mott-Schottky measurement and secondary ion mass spectroscopy (SIMS) were used to investigate the states and transport of hydrogen during the corrosion behavior of an AZ91 magnesium alloy in 0.1 M sodium sulfate solution. The results showed that when samples were immersed or charged in solution, hydrogen atoms diffused into the film and reacted with vacancy to cause the increases of the carrier concentration (excess electron or hole carrier) and diffusion rate of hydrogen. Some hydrogen atoms diffused to interior of matrix and enriched in β phase while others resorted in the corrosive film. With the increase of immersion or charging time, magnesium hydride would be brittle fractured when the inner stress caused by hydrogen pressure and expansion stress of formation of magnesium hydride was above the fracture strength, which provided the direct experimental evidence of the hydrogen embrittlement (HE) mechanism of magnesium and its alloys. After immersion in solution, the transfer of excess electrons to the interfaces of corrosion film and solution would destroy the charge equilibrium in the film and stimulate the adsorption of SO 4 2- , which resulted in the initiation of localized corrosion; after cathodic charging and then immersion, the enrichment of hydrogen atoms at interior of corrosion film would combine into hydrogen gas to form high pressure and result in the rupture of corrosion film, and localized corrosion initiated and developed at surface. Therefore, localized corrosion nucleated earlier on the charged samples than on the uncharged samples. Hydrogen invasion accelerated the corrosion of matrix

  12. Fuel Exhaling Fuel Cell.

    Science.gov (United States)

    Manzoor Bhat, Zahid; Thimmappa, Ravikumar; Devendrachari, Mruthyunjayachari Chattanahalli; Kottaichamy, Alagar Raja; Shafi, Shahid Pottachola; Varhade, Swapnil; Gautam, Manu; Thotiyl, Musthafa Ottakam

    2018-01-18

    State-of-the-art proton exchange membrane fuel cells (PEMFCs) anodically inhale H 2 fuel and cathodically expel water molecules. We show an unprecedented fuel cell concept exhibiting cathodic fuel exhalation capability of anodically inhaled fuel, driven by the neutralization energy on decoupling the direct acid-base chemistry. The fuel exhaling fuel cell delivered a peak power density of 70 mW/cm 2 at a peak current density of 160 mA/cm 2 with a cathodic H 2 output of ∼80 mL in 1 h. We illustrate that the energy benefits from the same fuel stream can at least be doubled by directing it through proposed neutralization electrochemical cell prior to PEMFC in a tandem configuration.

  13. Spent fuel and fuel pool component integrity. Annual report, FY 1979

    International Nuclear Information System (INIS)

    Johnson, A.B. Jr.; Bailey, W.J.; Schreiber, R.E.; Kustas, F.M.

    1980-05-01

    International meetings under the BEFAST program and under INFCE Working Group No. 6 during 1978 and 1979 continue to indicate that no cases of fuel cladding degradation have developed on pool-stored fuel from water reactors. A section from a spent fuel rack stand, exposed for 1.5 y in the Yankee Rowe (PWR) pool had 0.001- to 0.003-in.-deep (25- to 75-μm) intergranular corrosion in weld heat-affected zones but no evidence of stress corrosion cracking. A section of a 304 stainless steel spent fuel storage rack exposed 6.67 y in the Point Beach reactor (PWR) spent fuel pool showed no significant corrosion. A section of 304 stainless steel 8-in.-dia pipe from the Three Mile Island No. 1 (PWR) spent fuel pool heat exchanger plumbing developed a through-wall crack. The crack was intergranular, initiating from the inside surface in a weld heat-affected zone. The zone where the crack occurred was severely sensitized during field welding. The Kraftwerk Union (Erlangen, GFR) disassembled a stainless-steel fuel-handling machine that operated for 12 y in a PWR (boric acid) spent fuel pool. There was no evidence of deterioration, and the fuel-handling machine was reassembled for further use. A spent fuel pool at a Swedish PWR was decontaminated. The procedure is outlined in this report

  14. PEM fuel cell bipolar plate material requirements for transportation applications

    Energy Technology Data Exchange (ETDEWEB)

    Borup, R.L.; Stroh, K.R.; Vanderborgh, N.E. [Los Alamos National Lab., NM (United States)] [and others

    1996-04-01

    Cost effective bipolar plates are currently under development to help make proton exchange membrane (PEM) fuel cells commercially viable. Bipolar plates separate individual cells of the fuel cell stack, and thus must supply strength, be electrically conductive, provide for thermal control of the fuel stack, be a non-porous materials separating hydrogen and oxygen feed streams, be corrosion resistant, provide gas distribution for the feed streams and meet fuel stack cost targets. Candidate materials include conductive polymers and metal plates with corrosion resistant coatings. Possible metals include aluminium, titanium, iron/stainless steel and nickel.

  15. Development of a Matrix Alteration Model (MAM)

    International Nuclear Information System (INIS)

    Martinez Esparza, A.; Cunado, M. A.; Gago, J. A.; Quinones, J.; Iglesias, E.; Cobos, J.; Gonzalez de la Huebra, A.; Cera, E.; Merino, J.; Bruno, J.; Pablos, J. de; Casas, I.; Clarens, F.; Gimenez, J.

    2005-01-01

    The present report is a summary of the main tasks carried out within the WP4 of the SFS project (5th Framework Programme of the European Commission) by ENRESA and collaborators, mainly focused on the development of the so-called Matrix Alteration Model (MAM), a model to study the long-term oxidant dissolution of the spent fuel matrix under repository conditions. A variety of issues have been addressed: development of the MAM conceptual model, integration of a new matrix alteration mechanism in the radiolytic model, calibration and testing of the model, calculations for base case in Performance Assessment exercises, sensitivity analysis and an assessment of applicability of the MAM. The conceptual model for the UO2 oxidant dissolution is based on the processes expected to occur in the long term under repository conditions. Briefly, when water will enter in contact with the fuel surface, the first process we may expect is the radiolysis of water. Water radiolysis will generate reductant and oxidants and we may expect local oxidising conditions. Because of these local conditions, the surface of the fuel will be oxidized. The oxidation of the matrix and the attachment of aqueous ligands able to form strong complexes with its major component will favour the dissolution of the matrix. The integration of the matrix alteration (oxidation and dissolution) mechanism in the radiolytic model by means of elemental reactions has been mainly elucidated from the mechanistic models developed for non-irradiated UO2 dissolution experiments. Moreover, flow-through dissolution experiments with unirradiated UO2 have been used to calibrate the oxidative dissolution mechanism of UO2. The model developed has been able to reproduce experimental dissolution rates for pH > 5 and [HCO3 -] < 10-2 M when the oxidant is O2 at partial pressures lower than 21%, and 3 < pH < 9 and [HCO3 -] = 210-3 M and when the oxidant is H2O2 at concentrations below 10-4 M. These ranges cover the geochemical

  16. Anti corrosion layer for stainless steel in molten carbonate fuel cell - comprises phase vapour deposition of titanium nitride, aluminium nitride or chromium nitride layer then oxidising layer in molten carbonate electrolyte

    DEFF Research Database (Denmark)

    2000-01-01

    Forming an anticorrosion protective layer on a stainless steel surface used in a molten carbonate fuel cell (MCFC) - comprises the phase vapour deposition (PVD) of a layer comprising at least one of titanium nitride, aluminium nitride or chromium nitride and then forming a protective layer in situ...

  17. Microbiologically Influenced Corrosion

    Science.gov (United States)

    2009-01-01

    Aliphatic amines and nitrites used as corrosion inhibitors can be degraded by microorganisms, decreasing the effectiveness of the compounds and...workers (123) reported that silicone rubber dental liners could be degraded by yeasts. Moisture and chemical resistant polymeric coatings and...to be susceptible. 7.5. Concrete. Concrete is an inert aggregate, such as rock and gravel, surrounded by a cement binder. Concrete is a moderately

  18. Corrosion Protection of Aluminum

    Science.gov (United States)

    Dalrymple, R. S.; Nelson, W. B.

    1963-07-01

    Treatment of aluminum-base metal surfaces in an autoclave with an aqueous chromic acid solution of 0.5 to 3% by weight and of pH below 2 for 20 to 50 hrs at 160 to 180 deg C produces an extremely corrosion-resistant aluminum oxidechromium film on the surface. A chromic acid concentration of 1 to 2% and a pH of about 1 are preferred.

  19. Influence of mictrostructure features on the corrosion behaviourof AZ91 alloy in chloride media

    Directory of Open Access Journals (Sweden)

    Lenka Bukovinová

    2014-11-01

    Full Text Available The influence of the microstructure of as-cast AZ91 magnesium alloy, which applied to solution annealing treatment and ageing treatment respectively, was evaluated in terms of its corrosion behaviour in 0.1 M NaCl solution at room temperature. The corrosion process was monitored by electrochemical impedance spectroscopy (EIS and the surface was characterized by scanning Kelvin probe force microscopy (SKPFM. The extent of corrosion damage was dependent on the microstructure. Surface potential maps indicated that, the surface potential of α-matrix is more positive than surface potential of β phase.

  20. Advanced Corrosion-Resistant Zr Alloys for High Burnup and Generation IV Applications

    International Nuclear Information System (INIS)

    Arthur Motta; Yong Hwan Jeong; R.J. Comstock; G.S. Was; Y.S. Kim

    2006-01-01

    The objective of this collaboration between four institutions in the US and Korea is to demonstrate a technical basis for the improvement of the corrosion resistance of zirconium-based alloys in more extreme operating environments (such as those present in severe fuel duty, cycles high burnup, boiling, aggressive chemistry) and to investigate the feasibility (from the point of view of corrosion rate) of using advanced zirconium-based alloys in a supercritical water environment