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

International Nuclear Information System (INIS)

Wang, Lumin; Wierschke, Jonathan Brett

2015-01-01

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

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

International Nuclear Information System (INIS)

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

1988-11-01

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

Estimation of sesqui-carbide fraction for MARK-I fuel

International Nuclear Information System (INIS)

Vana Varamban, S.; Ananthasivan, K.

2016-01-01

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

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

International Nuclear Information System (INIS)

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

2017-01-01

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

High temperature corrosion of silicon carbide and silicon nitride in the presence of chloride compound

International Nuclear Information System (INIS)

McNallan, M.

1993-01-01

Silicon carbide and silicon nitride are resistant to oxidation because a protective silicon dioxide films on their surfaces in most oxidizing environments. Chloride compounds can attack the surface in two ways: 1) chlorine can attack the silicon directly to form a volatile silicon chloride compound or 2) alkali compounds combined with the chlorine can be transported to the surface where they flux the silica layer by forming stable alkali silicates. Alkali halides have enough vapor pressure that a sufficient quantity of alkali species to cause accelerated corrosion can be transported to the ceramic surface without the formation of a chloride deposit. When silicon carbide is attacked simultaneously by chlorine and oxygen, the corrosion products include both volatile and condensed spices. Silicon nitride is much more resistance to this type of attack than silicon carbide. Silicon based ceramics are exposed to oxidizing gases in the presence of alkali chloride vapors, the rate of corrosion is controlled primarily by the driving force for the formation of alkali silicate, which can be quantified as the activity of the alkali oxide in equilibrium with the corrosive gas mixture. In a gas mixture containing a fixed partial pressure of KCl, the rate of corrosion is accelerated by increasing the concentration of water vapor and inhibited by increasing the concentration of HCl. Similar results have been obtained for mixtures containing other alkalis and halogens. (Orig./A.B.)

Gas cooled fast breeder reactors using mixed carbide fuel

International Nuclear Information System (INIS)

Kypreos, S.

1976-09-01

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

Advances in carbide fuel element development for fast reactor application

International Nuclear Information System (INIS)

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

1977-01-01

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

Corrosion control in nuclear fuel reprocessing

International Nuclear Information System (INIS)

Steele, D.F.

1986-01-01

This article looks in detail at tribology-related hazards of corrosion in irradiated fuel reprocessing plants and tries to identify and minimize problems which could contribute to disaster. First, the corrosion process is explained. Then the corrosion aspects at each of four stages in reprocessing are examined, with particular reference to oxide fuel reprocessing. The four stages are fuel receipt and storage, fuel breakdown and dissolution, solvent extraction and product concentration and waste management. Results from laboratory and plant corrosion trails are used at the plant design stage to prevent corrosion problems arising. Operational procedures which minimize corrosion if it cannot be prevented at the design stage, are used. (UK)

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-04-08

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

Survey of post-irradiation examinations made of mixed carbide fuels

International Nuclear Information System (INIS)

Coquerelle, M.

1997-01-01

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

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

International Nuclear Information System (INIS)

Knight, Travis; Anghaie, Samim

2002-01-01

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

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

International Nuclear Information System (INIS)

Stauff, N.

2011-01-01

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

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

Directory of Open Access Journals (Sweden)

Hans Jürgen Grabke

2004-03-01

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

Silver diffusion through silicon carbide in microencapsulated nuclear fuels TRISO

International Nuclear Information System (INIS)

Cancino T, F.; Lopez H, E.

2013-10-01

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

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

International Nuclear Information System (INIS)

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

2001-01-01

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

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

International Nuclear Information System (INIS)

Radulescu, H.

2001-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

H. radulescu

2001-09-28

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

Fuel corrosion processes under waste disposal conditions

International Nuclear Information System (INIS)

Shoesmith, D.W.

2000-01-01

The release of the majority of radionuclides from spent nuclear fuel under permanent disposal conditions will be controlled by the rate of dissolution of the UO 2 fuel matrix. In this manuscript the mechanism of the coupled anodic (fuel dissolution) and cathodic (oxidant reduction) reactions which constitute the overall fuel corrosion process is reviewed, and the many published observations on fuel corrosion under disposal conditions discussed. The primary emphasis is on summarizing the overall mechanistic behaviour and establishing the primary factors likely to control fuel corrosion. Included are discussions on the influence of various oxidants including radiolytic ones, pH, temperature, groundwater composition, and the formation of corrosion product deposits. The relevance of the data recorded on unirradiated UO 2 to the interpretation of spent fuel behaviour is included. Based on the review, the data used to develop fuel corrosion models under the conditions anticipated in Yucca Mountain (NV, USA) are evaluated

Corrosion in ICPP fuel storage basins

International Nuclear Information System (INIS)

Dirk, W.J.

1993-09-01

The Idaho Chemical Processing Plant currently stores irradiated nuclear fuel in fuel storage basins. Historically, fuel has been stored for over 30 years. During the 1970's, an algae problem occurred which required higher levels of chemical treatment of the basin water to maintain visibility for fuel storage operations. This treatment led to higher levels of chlorides than seen previously which cause increased corrosion of aluminum and carbon steel, but has had little effect on the stainless steel in the basin. Corrosion measurements of select aluminum fuel storage cans, aluminum fuel storage buckets, and operational support equipment have been completed. Aluminum has exhibited good general corrosion rates, but has shown accelerated preferential attack in the form of pitting. Hot dipped zinc coated carbon steel, which has been in the basin for approximately 40 years, has shown a general corrosion rate of 4 mpy, and there is evidence of large shallow pits on the surface. A welded Type 304 stainless steel corrosion coupon has shown no attack after 13 years exposure. Galvanic couples between carbon steel welded to Type 304 stainless steel occur in fuel storage yokes exposed to the basin water. These welded couples have shown galvanic attack as well as hot weld cracking and intergranular cracking. The intergranular stress corrosion cracking is attributed to crevices formed during fabrication which allowed chlorides to concentrate

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

International Nuclear Information System (INIS)

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

1977-01-01

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

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

International Nuclear Information System (INIS)

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

1979-01-01

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

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

Fuel corrosion processes under waste disposal conditions

Energy Technology Data Exchange (ETDEWEB)

Shoesmith, D.W. [Univ. of Western Ontario, Dept. of Chemistry, London, Ontario (Canada)

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.

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

International Nuclear Information System (INIS)

Kastenberg, W.E.

1975-01-01

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

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

Science.gov (United States)

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

1992-06-01

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

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

International Nuclear Information System (INIS)

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

2015-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-07-01

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

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)

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

Ternary carbide uranium fuels for advanced reactor design applications

International Nuclear Information System (INIS)

Knight, Travis; Anghaie, Samim

1999-01-01

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

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

International Nuclear Information System (INIS)

Harry, G.R.

1983-01-01

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

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.

Aqueous Corrosion Rates for Waste Package Materials

International Nuclear Information System (INIS)

Arthur, S.

2004-01-01

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

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

International Nuclear Information System (INIS)

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

2008-01-01

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

Corrosion kinetics of 316L stainless steel bipolar plate with chromiumcarbide coating in simulated PEMFC cathodic environment

Directory of Open Access Journals (Sweden)

N.B. Huang

Full Text Available Stainless steel with chromium carbide coating is an ideal candidate for bipolar plates. However, the coating still cannot resist the corrosion of a proton exchange membrane fuel cell (PEMFC environment. In this work, the corrosion kinetics of 316L stainless steel with chromium carbide is investigated in simulated PEMFC cathodic environment by combining electrochemical tests with morphology and microstructure analysis. SEM results reveal that the steel’s surface is completely coated by Cr and chromium carbide but there are pinholes in the coating. After the coated 316L stainless steel is polarized, the diffraction peak of Fe oxide is found. EIS results indicate that the capacitive resistance and the reaction resistance first slowly decrease (2–32 h and then increase. The potentiostatic transient curve declines sharply within 2000 s and then decreases slightly. The pinholes, which exist in the coating, result in pitting corrosion. The corrosion kinetics of the coated 316L stainless steel are modeled and accords the following equation: i0 = 7.6341t−0.5, with the corrosion rate controlled by ion migration in the pinholes. Keywords: PEMFC, Metal bipolar plate, Chromium carbide coating, Corrosion kinetics, Pitting corrosion

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

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

International Nuclear Information System (INIS)

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

2004-01-01

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

Characterization of fuel swelling in helium-bonded carbide fuel pins

International Nuclear Information System (INIS)

Louie, D.L.Y.

1987-08-01

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

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

International Nuclear Information System (INIS)

Ombrellaro, P.A.

1977-06-01

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

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

International Nuclear Information System (INIS)

Venter, S.

1978-05-01

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

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1977-10-13

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

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

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

International Nuclear Information System (INIS)

Nguyen, D.H.

1980-06-01

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

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

International Nuclear Information System (INIS)

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

1992-01-01

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

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

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)

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

Energy Technology Data Exchange (ETDEWEB)

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

2008-07-01

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

Corrosion of graphite composites in phosphoric acid fuel cells

Science.gov (United States)

Christner, L. G.; Dhar, H. P.; Farooque, M.; Kush, A. K.

1986-01-01

Polymers, polymer-graphite composites and different carbon materials are being considered for many of the fuel cell stack components. Exposure to concentrated phosphoric acid in the fuel cell environment and to high anodic potential results in corrosion. Relative corrosion rates of these materials, failure modes, plausible mechanisms of corrosion and methods for improvement of these materials are investigated.

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

Energy Technology Data Exchange (ETDEWEB)

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

1979-09-30

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

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.)

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

International Nuclear Information System (INIS)

Huwyler, S.

1977-09-01

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

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

Contribution to the study of U-Ti and U-Pu-Ti carbides

International Nuclear Information System (INIS)

Milet, C.A.

1968-01-01

After having discussed the reasons to use (U,Pu) carbides as fast reactor fuel, we examine the influence of the addition of titanium to these carbides. A preliminary study has been done on the system of U-C-Ti and some properties have been measured such as: density, thermal expansion, electrical resistivity, atmospheric corrosion and compatibility with stainless steel. The systems U-Pu-C-Ti (Pu/U + Pu equal to 15 per cent) and U-C-Ti have been found to be very similar. There exists a two phases region (U,Pu)C + TiC, an eutectic between (U,Pu)C and TiC for approximately 15 at %. The solubilities of U + Pu in TiC and of Ti in (U,Pu)C is less than 1 % at. The addition of titanium does not markedly change thermal expansion coefficients of (U,Pu)C. However the resistance to atmospheric corrosion and compatibility with stainless steel is improved. Thermal conductivity, calculated from electrical resistivity, has increased. On the other side, the density of fissile material is lowered. The combination of (U,Pu)C + TiC seems to be the most promising alloy for application as nuclear fuel. (author) [fr

On the improvement of HTGR fuel elements corrosion resistance

International Nuclear Information System (INIS)

Chernikov, A.S.; Kurbakov, S.D.

1996-01-01

The results of corrosion tests of matrix graphite based on calcinated (30PG graphite) and non-calcinated (MPG graphite) petroleum cokes in helium containing 0.01-1 vol.% water vapour in the temperature range 600-1200degC are presented. The results of investigation of matrix graphite components reactivity are considered. It is shown that the filler graphite 30PG has the minimum activity towards the water vapour. The influence of impurities content on the oxidation rate are considered. The results of corrosion tests of matrix graphite coated with protective layers (silicon carbide and aluminium phosphates) in the air environment at 1600degC, 1 h, are given. (author)

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

Corrosion of research reactor Al-clad spent fuel in water

International Nuclear Information System (INIS)

Bendereskaya, O.S.; De, P.K.; Haddad, R.; Howell, J.P.; Johnson, A.B. Jr.; Laoharojanaphand, S.; Luo, S.; Ramanathan, L.V.; Ritchie, I.; Hussain, N.; Vidowsky, I.; Yakovlev, V.

2002-01-01

A significant amount of aluminium-clad spent nuclear fuel from research and test reactors worldwide is currently being stored in water-filled basins while awaiting final disposition. As a result of corrosion issues, which developed from the long-term wet storage of aluminium-clad fuel, the International Atomic Energy Agency (IAEA) implemented a Co-ordinated Research Project (CRP) in 1996 on the 'Corrosion of Research Reactor Aluminium-Clad Spent Fuel in Water'. The investigations undertaken during the CRP involved ten institutes in nine different countries. The IAEA furnished corrosion surveillance racks with aluminium alloys generally used in the manufacture of the nuclear fuel cladding. The individual countries supplemented these racks with additional racks and coupons specific to materials in their storage basins. The racks were immersed in late 1996 in the storage basins with a wide range of water parameters, and the corrosion was monitored at periodic intervals. Results of these early observations were reported after 18 months at the second research co-ordination meeting (RCM) in Sao Paulo, Brazil. Pitting and crevice corrosion were the main forms of corrosion observed. Corrosion caused by deposition of iron and other particles on the coupon surfaces was also observed. Galvanic corrosion of stainless steel/aluminium coupled coupons and pitting corrosion caused by particle deposition was observed. Additional corrosion racks were provided to the CRP participants at the second RCM and were immersed in the individual basins by mid-1998. As in the first set of tests, water quality proved to be the key factor in controlling corrosion. The results from the second set of tests were presented at the third and final RCM held in Bangkok, Thailand in October 2000. An IAEA document giving details about this CRP and other guidelines for spent fuel storage is in pres. This paper presents some details about the CRP and the basis for its extension. (author)

Radiolysis effects on fuel corrosion within a failed nuclear waste container

International Nuclear Information System (INIS)

Sunder, S.; Shoeshmith, D.W.; Christensen, H.C.

2003-01-01

The concept of geological disposal of used nuclear fuel in corrosion resistant containers is being investigated in several countries. In the Canadian Nuclear Fuel Waste Management Program (CNFWMP), it is assumed that the used fuel will be disposed of in copper containers. Since the predicted lifetimes of these containers are very long (>106 years), only those containers emplaced with an undetected defect will fail within the period for which radionuclide release from the fuel must be considered. Early failure could lead to the entry of water into the container and subsequent release of radionuclides. The release rate of radionuclides from the used fuel will depend upon its dissolution rate. The primary mechanism for release will be the corrosion of the fuel driven by radiolytically-produced oxidants. The studies carried out to determine the effects of water radiolysis on fuel corrosion are reviewed, and some of the procedures used to predict corrosion rates of used fuel in failed nuclear waste containers described. (author)

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

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.

Corrosion of spent Advanced Test Reactor fuel

International Nuclear Information System (INIS)

Lundberg, L.B.; Croson, M.L.

1994-01-01

The results of a study of the condition of spent nuclear fuel elements from the Advanced Test Reactor (ATR) currently being stored underwater at the Idaho National Engineering Laboratory (INEL) are presented. This study was motivated by a need to estimate the corrosion behavior of dried, spent ATR fuel elements during dry storage for periods up to 50 years. The study indicated that the condition of spent ATR fuel elements currently stored underwater at the INEL is not very well known. Based on the limited data and observed corrosion behavior in the reactor and in underwater storage, it was concluded that many of the fuel elements currently stored under water in the facility called ICPP-603 FSF are in a degraded condition, and it is probable that many have breached cladding. The anticipated dehydration behavior of corroded spent ATR fuel elements was also studied, and a list of issues to be addressed by fuel element characterization before and after forced drying of the fuel elements and during dry storage is presented

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

Nuclear fuel concept for the 21st century

International Nuclear Information System (INIS)

Tulenko, J.S.; Schoessow, G.

1996-01-01

In a previous paper, the author presented his rationale for the fuel cycle for the 21st century. This cycle, driven by both environmental and economic factors, required that the fuel should be able to operate in a range from 90 000 MWd/tonne of heavy metal and above. Such an operation would require the development of a cladding material that would not undergo waterside corrosion at these ultrahigh burnups. The University of Florida is proposing a new fuel arrangement that the authors feel meets the demands of high burnup and provides a safer fuel assembly. It is believed that the liquid-metal bond concept combined with a silicon carbide composite cladding and the collapsible fission gas plenum offers outstanding potential for ultrahigh burnup fuels while providing a potentially ultrasafe reactor operation. Efforts at various facilities are under way to determine the radiation stability of silicon carbide fuel and to fabricate SiC materials that will provide the radiation stability needed. Other parameters offer strong incentives to successfully develop silicon carbide as a cladding material

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

International Nuclear Information System (INIS)

Minato, Kazuo; Ogawa, Toru; Fukuda, Kousaku

1995-03-01

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

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

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.

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

Energy Technology Data Exchange (ETDEWEB)

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

2007-07-31

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

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)

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

Corrosion surveillance for research reactor spent nuclear fuel in wet basin storage

International Nuclear Information System (INIS)

Howell, J.P.

1999-01-01

Foreign and domestic test and research reactor fuel is currently being shipped from locations over the world for storage in water filled basins at the Savannah River Site (SRS). The fuel was provided to many of the foreign countries as a part of the ''Atoms for Peace'' program in the early 1950's. In support of the wet storage of this fuel at the research reactor sites and at SRS, corrosion surveillance programs have been initiated. The International Atomic Energy Agency (IAEA) established a Coordinated Research Program (CRP) in 1996 on ''Corrosion of Research Reactor Aluminum-Clad Spent Fuel in Water'' and scientists from ten countries worldwide were invited to participate. This paper presents a detailed discussion of the IAEA sponsored CRP and provides the updated results from corrosion surveillance activities at SRS. In May 1998, a number of news articles around the world reported stories that microbiologically influenced corrosion (MIC) was active on the aluminum-clad spent fuel stored in the Receiving Basin for Offsite Fuels (RBOF) at SRS. This assessment was found to be in error with details presented in this paper. A biofilm was found on aluminum coupons, but resulted in no corrosion. Cracks seen on the surface were not caused by corrosion, but by stresses from the volume expansion of the oxide formed during pre-conditioning autoclaving. There has been no pitting caused by MIC or any other corrosion mechanism seen in the RBOF basin since initiation of the SRS Corrosion Surveillance Program in 1993

PWR fuel rod corrosion in Japan

International Nuclear Information System (INIS)

Inoue, S.; Mori, K.; Murata, K.; Kobasyashi, S.

1997-01-01

Many particular appearance were observed on the fuel rod surfaces during fuel inspection at reactor outage in 1991. The appearances looked like small black circular nodules. The size was approximately 1 mm. This kind of appearances were found on fuel rods of which burnup exceeded approximately 30 GWd/t and at the second or third spans of the fuel assembly from the top. In order to clarify the cause, PIE was performed. The black nodules were confirmed to be oxide film spalling by visual inspection. Maximum oxide film thickness was 70 μm and spalling was observed where oxide thickness exceeded 40 t0 50 μm. Oxide film thickness was greater than expected. Many small pores were found in the oxide film when the oxide film had become thicker. Many circumferential cracks were also found in the film. It was speculated that these cracks caused the spalling of the oxide film. Hydride precipitates were mainly oriented circumferentially. Dense hydrides were observed near the outer rim of the cladding. No concentrated hydrides were observed near the spalling area. Maximum hydrogen content was 315 ppm. It was confirmed that the results of tensile test showed no significant effects by corrosion. The mechanism of accelerated corrosion was studied in detail. Water chemistry during irradiation was examined. Lithium content was maintained below 2.2 ppm. pH value was kept between 6.9 and 7.2. There was no anomalies in water chemistry during reactor operation. Cladding fabrication record clarified that heat treatment parameter was smaller than the optimum value. In Japan, heat treatment of the cladding was already optimized by improved fabrication process. Also chemical composition optimization of the cladding, such as low Tin and high Silicon content, was adopted for high burnup fuel. These remedies has already reduced fuel cladding corrosion and we believe we have solved this problem. (author). 6 figs, 1 tab

PWR fuel rod corrosion in Japan

Energy Technology Data Exchange (ETDEWEB)

Inoue, S [Kansai Electric Power Co., Inc., Osaka (Japan); Mori, K; Murata, K; Kobasyashi, S [Nuclear Fuel Industries, Ltd, Osaka (Japan)

1997-02-01

Many particular appearance were observed on the fuel rod surfaces during fuel inspection at reactor outage in 1991. The appearances looked like small black circular nodules. The size was approximately 1 mm. This kind of appearances were found on fuel rods of which burnup exceeded approximately 30 GWd/t and at the second or third spans of the fuel assembly from the top. In order to clarify the cause, PIE was performed. The black nodules were confirmed to be oxide film spalling by visual inspection. Maximum oxide film thickness was 70 {mu}m and spalling was observed where oxide thickness exceeded 40 t0 50 {mu}m. Oxide film thickness was greater than expected. Many small pores were found in the oxide film when the oxide film had become thicker. Many circumferential cracks were also found in the film. It was speculated that these cracks caused the spalling of the oxide film. Hydride precipitates were mainly oriented circumferentially. Dense hydrides were observed near the outer rim of the cladding. No concentrated hydrides were observed near the spalling area. Maximum hydrogen content was 315 ppm. It was confirmed that the results of tensile test showed no significant effects by corrosion. The mechanism of accelerated corrosion was studied in detail. Water chemistry during irradiation was examined. Lithium content was maintained below 2.2 ppm. pH value was kept between 6.9 and 7.2. There was no anomalies in water chemistry during reactor operation. Cladding fabrication record clarified that heat treatment parameter was smaller than the optimum value. In Japan, heat treatment of the cladding was already optimized by improved fabrication process. Also chemical composition optimization of the cladding, such as low Tin and high Silicon content, was adopted for high burnup fuel. These remedies has already reduced fuel cladding corrosion and we believe we have solved this problem. (author). 6 figs, 1 tab.

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.

Corrosion behaviour of zircaloy 4 fuel rod cladding in EDF power plants

Energy Technology Data Exchange (ETDEWEB)

Romary, H; Deydier, D [EDF, Direction de l` Equipment SEPTEN, Villeurbanne (France)

1997-02-01

Since the beginning of the French nuclear program, a surveillance of fuel has been carried out in order to evaluate the fuel behaviour under irradiation. Until now, nuclear fuels provided by suppliers have met EDF requirements concerning fuel behaviour and reliability. But, the need to minimize the costs and to increase the flexibility of the power plants led EDF to the definition of new targets: optimization of the core management and fuel cycle economy. The fuel behaviour experience shows that some of these new requirements cannot be fully fulfilled by the present standard fuel due to some technological limits. Particularly, burnup enhancement is limited by the oxidation and the hydriding of the Zircaloy 4 fuel rod cladding. Also, fuel suppliers and EDF need to have a better knowledge of the Zy-4 cladding behaviour in order to define the existing margins and the limiting factors. For this reason, in-reactor fuel characterization programs have been set up by fuel suppliers and EDF for a few years. This paper presents the main results and conclusions of EDF experience on Zy-4 in-reactor corrosion behaviour. Data obtained from oxide layer or zirconia thickness measurements show that corrosion performance of Zy-4 fuel rod cladding, as irradiated until now in EDF reactors, is satisfactory but not sufficient to meet the future needs. The fuel suppliers propose in order to improve the corrosion resistance of fuel rod cladding, low tin Zy-4 cladding and then optimized Zy-4 cladding. Irradiation of these claddings are ongoing. The available corrosion data show the better in-reactor corrosion resistance of optimized Zy-4 fuel rod cladding compared to the standard Zy-4 cladding. The scheduled fuel surveillance program will confirm if the optimized Zy-4 fuel rod cladding will meet the requirements for the future high burnup and high flexibility fuel. (author). 10 refs, 19 figs, 4 tabs.

Effect of water chemistry and fuel operation parameters on Zr + 1% Nb cladding corrosion

Energy Technology Data Exchange (ETDEWEB)

Kritsky, V G; Petrik, N G; Berezina, I G; Doilnitsina, V V [VNIPIET, St. Petersburg (Russian Federation)

1997-02-01

In-pile corrosion of Zr + 1%Nb fuel cladding has been studied. Zr-oxide and hydroxide solubilities at various temperatures and pH values have been calculated and correlations obtained between post-transition corrosion and the solubilities nodular corrosion and fuel operation parameters, as well as between the rate of fuel cladding degradation and water chemistry. Extrapolations of fuel assemblies behaviour to higher burnups have also performed. (author). 12 refs, 11 figs.

Problems raised by corrosion in the nuclear fuel cycle

International Nuclear Information System (INIS)

Tricot, R.; Boutonnet, G.; Perrot, M.; Blum, J.-M.

1977-01-01

In the uranium ore processing industry, materials which resist both mechanical abrasion and corrosion in an acid medium are required. Different typical cases are examined. For the reprocessing of irradiated fuels, two processes are possible: the conventional wet process, of the Purex type, and the fluoride volatilization process. In the latter case, the problems raised by fluoride corrosion in the presence of fission products is examined. The other parts of the fuel cycle are examined in the same manner [fr

Criteria for Corrosion Protection of Aluminum-Clad Spent Nuclear Fuel in Interim Wet Storage

International Nuclear Information System (INIS)

Howell, J.P.

1999-01-01

Storage of aluminum-clad spent nuclear fuel at the Savannah River Site (SRS) and other locations in the U. S. and around the world has been a concern over the past decade because of the long time interim storage requirements in water. Pitting corrosion of production aluminum-clad fuel in the early 1990''s at SRS was attributed to less than optimum quality water and corrective action taken has resulted in no new pitting since 1994. The knowledge gained from the corrosion surveillance testing and other investigations at SRS over the past 8 years has provided an insight into factors affecting the corrosion of aluminum in relatively high purity water. This paper reviews some of the early corrosion issues related to aluminum-clad spent fuel at SRS, including fundamentals for corrosion of aluminum alloys. It updates and summarizes the corrosion surveillance activities supporting the future storage of over 15,000 research reactor fuel assemblies from countries over the world during the next 15-20 years. Criteria are presented for providing corrosion protection for aluminum-clad spent fuel in interim storage during the next few decades while plans are developed for a more permanent disposition

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

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

International Nuclear Information System (INIS)

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

1983-01-01

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

Microstructural, phase evolution and corrosion properties of silicon carbide reinforced pulse electrodeposited nickel-tungsten composite coatings

Science.gov (United States)

Singh, Swarnima; Sribalaji, M.; Wasekar, Nitin P.; Joshi, Srikant; Sundararajan, G.; Singh, Raghuvir; Keshri, Anup Kumar

2016-02-01

Silicon carbide (SiC) reinforced nickel-tungsten (Ni-W) coatings were successfully fabricated on steel substrate by pulse electrodeposition method (PED) and the amount of SiC was varied as 0 g/l, 2 g/l, and 5 g/l in Ni-W coating. Effect of subsequent addition of SiC on microstructures, phases and on corrosion property of the coating was investigated. Field emission scanning electron microscopy (FE-SEM) image of the surface morphology of the coating showed the transformation from the dome like structure to turtle shell like structure. X-ray diffraction (XRD) of Ni-W-5 g/l SiC showed the disappearance of (220) plane of Ni(W), peak splitting in major peak of Ni(W) and formation of distinct peak of W(Ni) solid solution. Absence of (220) plane, peak splitting and presence of W(Ni) solid solution was explained by the high resolution transmission electron microscopy (HR-TEM) images. Tafel polarization plot was used to study the corrosion property of the coatings in 0.5 M NaCl solution. Ni-W-5 g/l SiC coating was showed higher corrosion resistance (i.e. ∼21% increase in corrosion potential, Ecorr) compared to Ni-W coating. Two simultaneous phenomena have been identified for the enhanced corrosion resistance of Ni-W-5 g/l SiC coating. (a) Presence of crystallographic texture (b) formation of continuous double barrier layer of NiWO4 and SiO2.

Impact of Zr + 2.5% Nb alloy corrosion upon operability of RBMK-1000 fuel channels

International Nuclear Information System (INIS)

Kovyrshin, V.; Zaritsky, N.

1999-01-01

The basic components of RBMK-1000 core (fuel channels, bimetal adapters, claddings of fuel elements, etc.) are of zirconium alloys. Their corrosion is one of factors influencing upon fuel channels operability. Dynamics of channel tubes nodular corrosion development is presented by the results of in-reactor investigation at ChNPP. Radiation-induced mechanism of corrosion damage of tubes surface in contact with coolant was formulated and substantiated by data of post-reactor studies. Within the certain time period of operation corrosion of zirconium alloy of lower bimetal adapter along with removal from there of corrosion products are predominant within the whole process of reactor elements corrosion. The experimental and calculating method was proposed and substantiated to predict time duration up to loss of fuel channels leak tightness. The approaches were generalized to control state of fuel channels material to assess their operability under operation of RBMK-1000 reactors. (author)

Corrosion of research reactor aluminium clad spent fuel in water

International Nuclear Information System (INIS)

2003-01-01

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

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

International Nuclear Information System (INIS)

K.L. Goluoglu

2000-01-01

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

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.

Oxidative corrosion of spent UO2 fuel in vapor and dripping groundwater at 900C

International Nuclear Information System (INIS)

Finch, R. J.

1999-01-01

Corrosion of spent UO 2 fuel has been studied in experiments conducted for nearly six years. Oxidative dissolution in vapor and dripping groundwater at 90 C occurs via general corrosion at fuel-fragment surfaces. Dissolution along fuel-grain boundaries is also evident in samples contacted by the largest volumes of groundwater, and corroded grain boundaries extend at least 20 or 30 grains deep (> 200 microm), possibly throughout millimeter-sized fragments. Apparent dissolution of fuel along defects that intersect grain boundaries has created dissolution pits that are 50 to 200 nm in diameter. Dissolution pits penetrate 1-2 microm into each grain, producing a ''worm-like'' texture along fuel-grain-boundaries. Sub-micrometer-sized fuel shards are common between fuel grains and may contribute to the reactive surface area of fuel exposed to groundwater. Outer surfaces of reacted fuel fragments develop a fine-grained layer of corrosion products adjacent to the fuel (5-15 microm thick). A more coarsely crystalline layer of corrosion products commonly covers the fine-grained layer, the thickness of which varies considerably among samples (from less than 5 microm to greater than 40 microm). The thickest and most porous corrosion layers develop on fuel fragments exposed to the largest volumes of groundwater. Corrosion-layer compositions depend strongly on water flux, with uranyl oxy-hydroxides predominating in vapor experiments, and alkali and alkaline earth uranyl silicates predominating in high drip-rate experiments. Low drip-rate experiments exhibit a complex assemblage of corrosion products, including phases identified in vapor and high drip-rate experiments

The corrosion of Zircaloy-4 fuel cladding in pressurized water reactors

International Nuclear Information System (INIS)

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

1991-01-01

This paper reports on the effects of thermo-mechanical processing of cladding on the corrosion of Zircaloy-4 in commercial PWRs that have been investigated. Visual observations and nondestructive measurements at poolside, augmented by observations in the hot cell, indicate that the initial black oxide transforms into a grey or tan later white oxide layer at a thickness of 10 to 15 μm independent of the thermal processing history of the tubing. At an oxide layer thickness of 60 to 80 μm, the oxide may spall depending somewhat on the particular oxide morphology formed and possibly on the frequency of power and temperature changes of the fuel rods. Because spalling of oxide lowers the metal-to-oxide interface temperature of fuel rods, it reduces the corrosion rate and is beneficial from that point of view. To determine the effect of thermo-mechanical processing on in-reactor corrosion of Zircaloy-4, oxide thickness measurements at poolside and in the hot cell have been analyzed with the MATPRO corrosion model. A calibrated corrosion parameter in this model provides a measure of the corrosion susceptibility of the Zircaloy-4 cladding. It was found necessary to modify the MATPRO equations with a burnup dependent term to obtain a near constant value of the corrosion parameter over a burnup range of approximately 10 to 45 MWd/kgU. Different calculational tests were performed to confirm that the modified model accurately predicts the corrosion behavior of fuel rods

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)

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.

Corrosion effect of fast reactor fuel claddings on their mechanical properties

International Nuclear Information System (INIS)

Davydov, E.F.; Krykov, F.N.; Shamardin, V.K.

1985-01-01

Fast reactor fuel cladding corrosion effect on its mechanical properties was investigated. UO 2 fuel elements were irradiated in the BOP-60 reactor at the linear heat rate of 42 kw/m. Fuel cladding is made of stainless steel OKh16N15M3BR. Calculated maximum cladding temperature is 920 K. Neutron fluence in the central part of fuel elements is 6.3x10 26 m+H- 2 . To investigate the strength changes temperature dependence of corrossion depth, cladding strength reduction factors was determined. Samples plasticity reduction with corrosion layer increase is considered to be a characteristic feature

Prediction of Agglomeration, Fouling, and Corrosion Tendency of Fuels in CFB Co-Combustion

Science.gov (United States)

Barišć, Vesna; Zabetta, Edgardo Coda; Sarkki, Juha

Prediction of agglomeration, fouling, and corrosion tendency of fuels is essential to the design of any CFB boiler. During the years, tools have been successfully developed at Foster Wheeler to help with such predictions for the most commercial fuels. However, changes in fuel market and the ever-growing demand for co-combustion capabilities pose a continuous need for development. This paper presents results from recently upgraded models used at Foster Wheeler to predict agglomeration, fouling, and corrosion tendency of a variety of fuels and mixtures. The models, subject of this paper, are semi-empirical computer tools that combine the theoretical basics of agglomeration/fouling/corrosion phenomena with empirical correlations. Correlations are derived from Foster Wheeler's experience in fluidized beds, including nearly 10,000 fuel samples and over 1,000 tests in about 150 CFB units. In these models, fuels are evaluated based on their classification, their chemical and physical properties by standard analyses (proximate, ultimate, fuel ash composition, etc.;.) alongside with Foster Wheeler own characterization methods. Mixtures are then evaluated taking into account the component fuels. This paper presents the predictive capabilities of the agglomeration/fouling/corrosion probability models for selected fuels and mixtures fired in full-scale. The selected fuels include coals and different types of biomass. The models are capable to predict the behavior of most fuels and mixtures, but also offer possibilities for further improvements.

Influence of carbide precipitation upon hydrogen fragilization of an AISI 304 steel

International Nuclear Information System (INIS)

Hazarabedian, A.E.; Ovejero Garcia, J.

1991-01-01

The present work deals with austenitic stainless steels for a family of steels that is renowned for its high resistance to hydrogen fragilization. Nevertheless, these steels may suffer hydrogen fragilization under severe working conditions. This fact is strongly dependent on many factors -composition, grain size, other phases present, corrosion sensitivity, etc.-. While there are studies that show how intergranular corrosion is influenced by corrosion sensitivity -mainly due to carbide precipitation in grain boundaries-, there are no reports about the effect of the carbide precipitation itself on hydrogen fragilization for these steels. (Author) [es

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

International Nuclear Information System (INIS)

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

2016-01-01

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

Flue gas corrosion through halogen compounds in fuel gas

Energy Technology Data Exchange (ETDEWEB)

Eisenmann, R

1987-04-01

The halogens of chlorine and fluorine greatly influence the corrosion speed of metal materials. If small quantities of chlorinated and/or fluorinated hydrocarbons are present in fuel gas like in landfill gas, they must not result in enhanced corrosion of gas appliances. Data from literature and the initial results of tests run by the author indicate that quantities at about 10 mg/cbm (in terms of chlorine) can be assumed not to cause any noticeable acceleration of corrosion speed.

Corrosion and protection of spent Al-clad research reactor fuel during extended wet storage

International Nuclear Information System (INIS)

Ramanathan, Lalgudi V.

2009-01-01

A variety of spent research reactor fuel elements with different fuel meats, geometries and 235 U enrichments are presently stored under water in basins throughout the world. More than 90% of these fuels are clad in aluminum (Al) or its alloy and are susceptible to corrosion. This paper presents an overview of the influence of Al alloy composition, galvanic effects (Al alloy/stainless steel), crevice effects, water parameters and synergism between these parameters as well as settled solids on the corrosion of typical Al alloys used as fuel element cladding. Pitting is the main form of corrosion and is affected by water conductivity, chloride ion content, formation of galvanic couples with rack supports and settled solid particles. The extent to which these parameters influence Al corrosion varies. This paper also presents potential conversion coatings to protect the spent fuel cladding. (author)

Corrosion of titanium and titanium alloys in spent fuel repository conditions - literature review

International Nuclear Information System (INIS)

Aho-Mantila, I.; Haenninen, H.; Aaltonen, P.; Taehtinen, S.

1985-03-01

The spent nuclear fuel is planned to be disposed in Finnish bedrock. The canister of spent fuel in waste repository is one barrier to the release of radionuclides. It is possible to choose a canister material with a known, measurable corrosion rate and to make it with thickness allowing corrosion to occur. The other possibility is to use a material which is nearly immune to general corrosion. In this second category there are titanium and titanium alloys which exhibit a very high degree of resistance to general corrosion. In this literature study the corrosion properties of unalloyed titanium, titanium alloyed with palladium and titanium alloyed with molybdenum and nickel are reviewed. The two titanium alloys own in addition to the excellent general corrosion properties outstanding properties against localized corrosion like pitting or crevice corrosion. Stress corrosion cracking and corrosion fatique of titanium seem not to be a problem in the repository conditions, but the possibilities of delayed cracking caused by hydrogen should be carefully appreciated. (author)

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)

Corrosion Studies of Platinum Nano-Particles for Fuel Cells

DEFF Research Database (Denmark)

Shim, Signe Sarah

The main focus of the present thesis is on corrosion and prevention of corrosion of platinum particles supported on carbon. This is important for instance in connection with start up and shutdown of fuel cells. The degradation mechanism of platinum particles supported on carbon has been character......The main focus of the present thesis is on corrosion and prevention of corrosion of platinum particles supported on carbon. This is important for instance in connection with start up and shutdown of fuel cells. The degradation mechanism of platinum particles supported on carbon has been...... characterized during oxygen reduction reaction (ORR) condition using identical location (IL) transmission electron microscopy (TEM). A TEM grid was used as the working electrode in an electrochemical setup allowing a direct correlation between the electrochemical response and the TEM analysis. The main results...... thirds and one monolayer of gold on platinum supported on carbon were synthesized by an inverse micelle method. The results obtained appear independent of the gold coverage. It has been shown that the electrochemical active surface areas of the platinum and platinum gold particles synthesized...

Corrosion testing of uranium silicide fuel specimens

International Nuclear Information System (INIS)

Bourns, W.T.

1968-09-01

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

CANDU reactors with reactor grade plutonium/thorium carbide fuel

Energy Technology Data Exchange (ETDEWEB)

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

2011-08-15

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

Anticipated corrosion in the Vermont Yankee spent fuel pool

International Nuclear Information System (INIS)

Weeks, J.R.

1977-06-01

The report provides additional information relating to a proposed modification to the spent fuel pool at the Vermont Yankee Nuclear Power Station (VYNPS) and addresses corrosion of spent fuel pool storage materials and zircaloy, and provides an analysis of the effectiveness of the Boral sealing

Formation and corrosion of a 410 SS/ceramic composite

Energy Technology Data Exchange (ETDEWEB)

Chen, X., E-mail: xin.chen@anl.gov [Civil and Materials Engineering Department, University of Illinois at Chicago, 842 W. Taylor St., Chicago, IL 60607 (United States); Nuclear Engineering Division, Argonne National Laboratory, 9700 S Cass Ave, Lemont, IL 60439 (United States); Ebert, W.L. [Nuclear Engineering Division, Argonne National Laboratory, 9700 S Cass Ave, Lemont, IL 60439 (United States); Indacochea, J.E. [Civil and Materials Engineering Department, University of Illinois at Chicago, 842 W. Taylor St., Chicago, IL 60607 (United States)

2016-11-15

This study addressed the possible use of alloy/ceramic composite waste forms to immobilize metallic and oxide waste streams generated during the electrochemical reprocessing of spent reactor fuel using a single waste form. A representative composite material was made to evaluate the microstructure and corrosion behavior at alloy/ceramic interfaces by reacting 410 stainless steel with Zr, Mo, and a mixture of lanthanide oxides. Essentially all of the available Zr reacted with lanthanide oxides to generate lanthanide zirconates, which combined with the unreacted lanthanide oxides to form a porous ceramic network that filled with alloy to produce a composite puck. Alloy present in excess of the pore volume of the ceramic generated a metal bead on top of the puck. The alloys in the composite and forming the bead were both mixtures of martensite grains and ferrite grains bearing carbide precipitates; FeCrMo intermetallic phases also precipitated at ferrite grain boundaries within the composite puck. Micrometer-thick regions of ferrite surrounding the carbides were sensitized and corroded preferentially in electrochemical tests. The lanthanide oxides dissolved chemically, but the lanthanide zirconates did not dissolve and are suitable host phases. The presence of oxide phases did not affect corrosion of the neighboring alloy phases. - Highlights: • An alloy/ceramic composite was made to evaluate corrosion at phase boundaries. • Lanthanide oxides and Zr added to 410 steel reacted to form durable zirconates. • Corrosion behavior was evaluated using electrochemical tests and SEM analyses. • Regions of active, passive, galvanic, sensitized, and chemical corrosion observed. • The corrosion current was proportional to relative areas of active alloy phases.

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

International Nuclear Information System (INIS)

Handa, Muneo

1977-11-01

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

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

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

Corrosion testing of uranium silicide fuel specimens

Energy Technology Data Exchange (ETDEWEB)

Bourns, W T

1968-09-15

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

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

International Nuclear Information System (INIS)

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

1980-10-01

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

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.)

Corrosion and alteration of materials from the nuclear industry; La Corrosion et l'alteration des materiaux du nucleaire

Energy Technology Data Exchange (ETDEWEB)

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.

2010-07-01

, testing means, experimental techniques, internal corrosion of zircaloy sheath - the iodine effect, stress corrosion of nickel alloys - hydrogen influence, stress corrosion of stainless steels; C - wear corrosion: a coupled phenomenon, research in the framework of service life extension of the French electronuclear park; 3 - Corrosion in future reactors: A - corrosion in gas reactors: corrosion by helium impurities, oxidation resistance of silicon carbide, corrosion of graphite and carbon-carbon composites; B - corrosion in liquid metal reactors: sodium FBRs, lead and lead alloys reactors; C- corrosion in molten salt reactors: corrosion of Hastelloy N-type nickel alloys by molten fluorides, mass transfer in aniso-thermal fluoride systems, tellurium embrittlement, electrochemical study of pure metals corrosion in molten fluorides; 4 - Materials corrosion and alteration in the back-end of the fuel cycle: A - corrosion in concentrated nitric environment: materials behaviour, self-catalytic mechanism of nitric acid reduction; B - corrosion in unsaturated aqueous environment: metallic corrosion in unsaturated environment - application to the storage of waste containers, bitumens alteration, reinforced concrete behaviour and iron framework corrosion, concrete behaviour in severe thermal environment; C - Corrosion in saturated aqueous environment: metals corrosion in clayey environment, long-term behaviour of glasses, ceramics alteration, underwater concrete durability, clays transformation; D - materials biodegradation: microorganisms and nuclear wastes, biodegradation of bitumen, concretes and steels; 5 - Conclusion, glossary

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

Atmospheric corrosion of uranium-carbon alloys

International Nuclear Information System (INIS)

Rousset, P.; Accary, A.

1965-01-01

The authors study the corrosion of uranium-carbon alloys having compositions close to that of the mono-carbide; they show that the extent of the observed corrosion effects increases with the water vapour content of the surrounding gas and they conclude that the atmospheric corrosion of these alloys is due essentially to the humidity of the air, the effect of the oxygen being very slight at room temperature. They show that the optimum conditions for preserving U-C alloys are either a vacuum or a perfectly dry argon atmosphere. The authors have also established that the type of corrosion involved is a corrosion which 'cracks under stress' and is transgranular (it can also be intergranular in the case of sub-stoichiometric alloys). They propose, finally, two hypotheses for explaining this mechanism, one of which is illustrated by the existence, at the fissure interface, of corrosion products which can play the role of 'corners' in the mono-carbide grains. (authors) [fr

Microstructural, phase evolution and corrosion properties of silicon carbide reinforced pulse electrodeposited nickel–tungsten composite coatings

International Nuclear Information System (INIS)

Singh, Swarnima; Sribalaji, M.; Wasekar, Nitin P.; Joshi, Srikant; Sundararajan, G.; Singh, Raghuvir; Keshri, Anup Kumar

2016-01-01

Graphical abstract: - Highlights: • Pulse electrodeposited Ni–W–SiC coating has been synthesized successfully. • Dome to turtle like structure has been observed on addition of SiC in Ni–W coating. • Formation of W(Ni) solid solution was observed on adding 5 g/l SiC in Ni–W coating. • Corrosion resistance improved for Ni–W–5 g/l SiC coating. • Texture formation and continuous barrier layer enhanced the corrosion resistance. - Abstract: Silicon carbide (SiC) reinforced nickel–tungsten (Ni–W) coatings were successfully fabricated on steel substrate by pulse electrodeposition method (PED) and the amount of SiC was varied as 0 g/l, 2 g/l, and 5 g/l in Ni–W coating. Effect of subsequent addition of SiC on microstructures, phases and on corrosion property of the coating was investigated. Field emission scanning electron microscopy (FE-SEM) image of the surface morphology of the coating showed the transformation from the dome like structure to turtle shell like structure. X-ray diffraction (XRD) of Ni–W–5 g/l SiC showed the disappearance of (220) plane of Ni(W), peak splitting in major peak of Ni(W) and formation of distinct peak of W(Ni) solid solution. Absence of (220) plane, peak splitting and presence of W(Ni) solid solution was explained by the high resolution transmission electron microscopy (HR-TEM) images. Tafel polarization plot was used to study the corrosion property of the coatings in 0.5 M NaCl solution. Ni–W–5 g/l SiC coating was showed higher corrosion resistance (i.e. ∼21% increase in corrosion potential, E_c_o_r_r) compared to Ni–W coating. Two simultaneous phenomena have been identified for the enhanced corrosion resistance of Ni–W–5 g/l SiC coating. (a) Presence of crystallographic texture (b) formation of continuous double barrier layer of NiWO_4 and SiO_2.

Microstructural, phase evolution and corrosion properties of silicon carbide reinforced pulse electrodeposited nickel–tungsten composite coatings

Energy Technology Data Exchange (ETDEWEB)

Singh, Swarnima; Sribalaji, M. [Materials Science and Engineering, Indian Institute of Technology Patna, Navin Government Polytechnic Campus, Patliputra Colony, Patna, Bihar 800013 (India); Wasekar, Nitin P.; Joshi, Srikant; Sundararajan, G. [International Advanced Research Centre for Powder Metallurgy & New Materials (ARCI) Hyderabad, Balapur P.O., Hyderabad, Andhra Pradesh 500005 (India); Singh, Raghuvir [CSIR-National Metallurgical Laboratory, Jamshedpur, Jharkhand 831007 (India); Keshri, Anup Kumar, E-mail: anup@iitp.ac.in [Materials Science and Engineering, Indian Institute of Technology Patna, Navin Government Polytechnic Campus, Patliputra Colony, Patna, Bihar 800013 (India)

2016-02-28

Graphical abstract: - Highlights: • Pulse electrodeposited Ni–W–SiC coating has been synthesized successfully. • Dome to turtle like structure has been observed on addition of SiC in Ni–W coating. • Formation of W(Ni) solid solution was observed on adding 5 g/l SiC in Ni–W coating. • Corrosion resistance improved for Ni–W–5 g/l SiC coating. • Texture formation and continuous barrier layer enhanced the corrosion resistance. - Abstract: Silicon carbide (SiC) reinforced nickel–tungsten (Ni–W) coatings were successfully fabricated on steel substrate by pulse electrodeposition method (PED) and the amount of SiC was varied as 0 g/l, 2 g/l, and 5 g/l in Ni–W coating. Effect of subsequent addition of SiC on microstructures, phases and on corrosion property of the coating was investigated. Field emission scanning electron microscopy (FE-SEM) image of the surface morphology of the coating showed the transformation from the dome like structure to turtle shell like structure. X-ray diffraction (XRD) of Ni–W–5 g/l SiC showed the disappearance of (220) plane of Ni(W), peak splitting in major peak of Ni(W) and formation of distinct peak of W(Ni) solid solution. Absence of (220) plane, peak splitting and presence of W(Ni) solid solution was explained by the high resolution transmission electron microscopy (HR-TEM) images. Tafel polarization plot was used to study the corrosion property of the coatings in 0.5 M NaCl solution. Ni–W–5 g/l SiC coating was showed higher corrosion resistance (i.e. ∼21% increase in corrosion potential, E{sub corr}) compared to Ni–W coating. Two simultaneous phenomena have been identified for the enhanced corrosion resistance of Ni–W–5 g/l SiC coating. (a) Presence of crystallographic texture (b) formation of continuous double barrier layer of NiWO{sub 4} and SiO{sub 2}.

Corrosion database for the nuclear fuel cycle. Sub-project no. 1

International Nuclear Information System (INIS)

Schoenfeld, R.; Wegner, K.

1989-03-01

The aim of the project was to prepare and process data on corrosion in fuel recycling systems of fast breeder reactors and to store them in a test data base designed as an information system. Based on examinations on the nitric acid corrosion of austenitic steels (typical material/corrosive agent combination used in the reprocessing of burned fuel elements of nuclear power plants) and, in coordination with scientist specialized on materials, the most important characteristics were determined and summarized in a catalogue. This catalogue was realized with the help of a relational data base management system as a scientific data base where the adequate information from the original literature is recorded. (orig./MM) [de

1. The application of PIE techniques to the study of the corrosion of spent oxide fuel in deep-rock groundwaters. 2. Spent fuel degradation

International Nuclear Information System (INIS)

Forsyth, R.S.

1991-01-01

During the autumn of 1990, papers summarizing work performed at Studsvik as part of the SKB research programme designed to study the corrosion behaviour of spent nuclear fuel in deep-rock groundwater were presented at two scientific meetings: The first paper presents results and observations of the study of the corrosion of spent oxide fuel in deep-rock ground-waters. The PIE techniques were applied to the detailed study of spent fuel both before and after water contact. The second paper represents an up-dated reporting of results obtained in the Swedish programme relevant to preferential dissolution effects, including interim results from recently stored experiments specifically designed to study possible correlations between corrosion behaviour and fuel properties conditioned by burnup and/or local power variations. Recent observations during the search for corrosion sites in fuel exposed to corrosion for about 4 years are also presented. (KAE)

Drying characteristics of thorium fuel corrosion products

Energy Technology Data Exchange (ETDEWEB)

Smith, R.-E. E-mail: rzl@inel.gov

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 deg. C. Complete removal of chemisorbed water requires heating above 1000 deg. 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.

Assessment of Silicon Carbide Composites for Advanced Salt-Cooled Reactors

Energy Technology Data Exchange (ETDEWEB)

Katoh, Yutai [ORNL; Wilson, Dane F [ORNL; Forsberg, Charles W [ORNL

2007-09-01

The Advanced High-Temperature Reactor (AHTR) is a new reactor concept that uses a liquid fluoride salt coolant and a solid high-temperature fuel. Several alternative fuel types are being considered for this reactor. One set of fuel options is the use of pin-type fuel assemblies with silicon carbide (SiC) cladding. This report provides (1) an initial viability assessment of using SiC as fuel cladding and other in-core components of the AHTR, (2) the current status of SiC technology, and (3) recommendations on the path forward. Based on the analysis of requirements, continuous SiC fiber-reinforced, chemically vapor-infiltrated SiC matrix (CVI SiC/SiC) composites are recommended as the primary option for further study on AHTR fuel cladding among various industrially available forms of SiC. Critical feasibility issues for the SiC-based AHTR fuel cladding are identified to be (1) corrosion of SiC in the candidate liquid salts, (2) high dose neutron radiation effects, (3) static fatigue failure of SiC/SiC, (4) long-term radiation effects including irradiation creep and radiation-enhanced static fatigue, and (5) fabrication technology of hermetic wall and sealing end caps. Considering the results of the issues analysis and the prospects of ongoing SiC research and development in other nuclear programs, recommendations on the path forward is provided in the order or priority as: (1) thermodynamic analysis and experimental examination of SiC corrosion in the candidate liquid salts, (2) assessment of long-term mechanical integrity issues using prototypical component sections, and (3) assessment of high dose radiation effects relevant to the anticipated operating condition.

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.

Influence of LMFBR fuel pin temperature profiles on corrosion rate

International Nuclear Information System (INIS)

Shiels, S.A.; Bagnall, C.; Schrock, S.L.; Orbon, S.J.

1976-01-01

The paper describes the sodium corrosion behavior of 20 percent cold worked Type 316 stainless steel fuel pin cladding under a simulated reactor thermal environment. A temperature gradient, typical of a fuel pin, was generated in a 0.9 m long heater section by direct resistance heating. Specimens were located in an isothermal test section immediately downstream of the heater. A comparison of the measured corrosion rates with available data showed an enhancement factor of between 1.5 and 2 which was attributed to the severe axial temperature gradient through the heater. Differences in structure and surface chemistry were also noted

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

Advanced fuels for gas turbines: Fuel system corrosion, hot path deposit formation and emissions

International Nuclear Information System (INIS)

Seljak, Tine; Širok, Brane; Katrašnik, Tomaž

2016-01-01

Highlights: • Technical feasibility analysis of alternative fuels requires a holistic approach. • Fuel, combustion, corrosion and component functionality are strongly related. • Used approach defines design constraints for microturbines using alternative fuels. - Abstract: To further expand the knowledge base on the use of innovative fuels in the micro gas turbines, this paper provides insight into interrelation between specific fuel properties and their impact on combustion and emission formation phenomena in micro gas turbines for stationary power generation as well as their impact on material corrosion and deposit formation. The objective of this study is to identify potential issues that can be related to specific fuel properties and to propose counter measures for achieving stable, durable, efficient and low emission operation of the micro gas turbine while utilizing advanced/innovative fuels. This is done by coupling combustion and emission formation analyses to analyses of material degradation and degradation of component functionality while interpreting them through fuel-specific properties. To ensure sufficiently broad range of fuel properties to demonstrate the applicability of the method, two different fuels with significantly different properties are analysed, i.e. tire pyrolysis oil and liquefied wood. It is shown that extent of required micro gas turbine adaptations strongly correlates with deviations of the fuel properties from those of the baseline fuel. Through the study, these adaptations are supported by in-depth analyses of impacts of fuel properties on different components, parameters and subsystems and their quantification. This holistic approach is further used to propose methodologies and innovative approaches for constraining a design space of micro gas turbine to successfully utilize wide spectra of alternative/innovative fuels.

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)

Corrosion of aluminium-clad spent fuel at RA research reactor

International Nuclear Information System (INIS)

Pesic, M.; Maksin, T.; Dobrijevic, R.; Idjakovic, Z.

2003-01-01

Almost 95% of all spent fuel elements of the RA research reactor in the Vinca Institute of Nuclear Sciences, Belgrade, Serbia and Montenegro, are stored in 30 aluminium barrels and about 300 stainless steel channel-holders in the temporary spent fuel storage water pool. The first activities of sludge and water samples, taken from the pool, were measured in 1996-1997 and were followed by analysis of chemical composition of samples. Visual inspections of fuel elements in some stainless steel tubes and of the fuel channels stored in the reactor core have shown that some deposits cover aluminium cladding. Stains and surface discoloration are noted on many of the spent fuel elements that were examined visually during the core unloading and inspections carried out in 1979 - 1984. Some of water samples, taken from pool, about a 150 stainless steel tubes and 16 barrels have shown very high 137-Cs activity compared to low activity measured in pool water. It was concluded that aluminium cladding of the fuel elements was penetrated due to corrosion process. Study on influence of water corrosion processes in the RA reactor storage pool was started within the framework of the IAEA CRP 'Corrosion of Research Reactor Aluminium-Clad Spent Fuel in Water' in 2002. The first test rack with various aluminium and stainless steel coupons, supplied by the IAEA, was immersed in the pool already in 1996. New racks were immersed in 2002 and 2003. The rack immersed in 1996 was taken out from the pool in 2002 and the rack immersed in 2002 was taken out in 2003. Results of the examination of these racks, carried out according to the strategy and the protocol, proposed by the IAEA, are described in this paper. (author)

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.

On the carbide formation in high-carbon stainless steel

International Nuclear Information System (INIS)

Mujahid, M.; Qureshi, M.I.

1996-01-01

Stainless steels containing high Cr as well as carbon contents in excess of 1.5 weight percent have been developed for applications which require high resistance erosion and environmental corrosion. Formation of carbides is one of important parameters for controlling properties of these materials especially erosion characteristics. Percent work includes the study of different type of carbides which from during the heat treatment of these materials. It has been found that precipitation of secondary carbides and the nature of matrix transformation plays an important role in determining the hardness characteristics of these materials. (author)

Fuel-cladding chemical interaction

International Nuclear Information System (INIS)

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

2015-01-01

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

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

International Nuclear Information System (INIS)

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

1993-01-01

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

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.

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)

Corrosion issues in the long term storage of aluminum-clad spent nuclear fuels

International Nuclear Information System (INIS)

Louthan, M.R. Jr.; Peacock, H.B. Jr.; Sindelar, R.L.; Iyer, N.C.

1996-01-01

Approximately 8% of the spent nuclear fuel owned by the US Department of Energy is clad with aluminum alloys. The spent fuel must be either reprocessed or temporarily stored in wet or dry storage systems until a decision is made on final disposition in a repository. There are corrosion issues associated with the aluminum cladding regardless of the disposition pathway selected. This paper discusses those issues and provides data and analysis to demonstrate that control of corrosion induced degradation in aluminum clad spent fuels can be achieved through relatively simple engineering practices

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)

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

, testing means, experimental techniques, internal corrosion of zircaloy sheath - the iodine effect, stress corrosion of nickel alloys - hydrogen influence, stress corrosion of stainless steels; C - wear corrosion: a coupled phenomenon, research in the framework of service life extension of the French electronuclear park; 3 - Corrosion in future reactors: A - corrosion in gas reactors: corrosion by helium impurities, oxidation resistance of silicon carbide, corrosion of graphite and carbon-carbon composites; B - corrosion in liquid metal reactors: sodium FBRs, lead and lead alloys reactors; C- corrosion in molten salt reactors: corrosion of Hastelloy N-type nickel alloys by molten fluorides, mass transfer in aniso-thermal fluoride systems, tellurium embrittlement, electrochemical study of pure metals corrosion in molten fluorides; 4 - Materials corrosion and alteration in the back-end of the fuel cycle: A - corrosion in concentrated nitric environment: materials behaviour, self-catalytic mechanism of nitric acid reduction; B - corrosion in unsaturated aqueous environment: metallic corrosion in unsaturated environment - application to the storage of waste containers, bitumens alteration, reinforced concrete behaviour and iron framework corrosion, concrete behaviour in severe thermal environment; C - Corrosion in saturated aqueous environment: metals corrosion in clayey environment, long-term behaviour of glasses, ceramics alteration, underwater concrete durability, clays transformation; D - materials biodegradation: microorganisms and nuclear wastes, biodegradation of bitumen, concretes and steels; 5 - Conclusion, glossary

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

Energy Technology Data Exchange (ETDEWEB)

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

2013-10-15

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

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

Science.gov (United States)

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

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

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

Science.gov (United States)

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

1982-01-01

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

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

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

Corrosion inhibition studies in support of the long term storage of AGR fuel

Energy Technology Data Exchange (ETDEWEB)

Standring, P [Sellafield Limited (United Kingdom)

2012-07-01

Thorp Receipt and Storage (at Sellafield, UK) is currently being investigated as a bridging solution for the storage of AGR fuel pending the out-come of a national review into spent fuel management. AGR spent fuel is known to be susceptible to corrosion through inter-granular attack. To avoid this, the chosen storage regime for AGR fuel is sodium hydroxide dosed pond water to pH 11.4; now 22 years of operating experience. The conversion of TR and S will require a phased transition. During this transition sodium hydroxide cannot be used due to materials compatibility issues. Alternative corrosion inhibitors have been investigated as an interim measure and sodium nitrate has been selected as a suitable candidate. The efficiency of sodium nitrate to inhibit propagating inter-granular attack of active AGR materials has yet to be established. In the longer term sodium hydroxide will be deployed along with a move to a closed loop pond water management system. Given that carbon dioxide is known to be absorbed by sodium hydroxide dosed water and can affect fuel integrity, in the case of Magnox fuel, there is a need to establish its impact on AGR fuel. The objectives are: To establish the impact of carbonate on AGR fuel corrosion; To establish the efficiency of sodium nitrate to inhibit propagating inter-granular attack of irradiated AGR materials.

Factors affecting the corrosion of SiC layer by fission product palladium

International Nuclear Information System (INIS)

Dewita, E.

2000-01-01

HTR is one of the advanced nuclear reactors which has inherent safety system, graphite moderated and helium gas cooled. In general, these reactors are designed with the TRISO coated particle consist of four coating layers that are porous pyrolytic carbon (PyC). inner dense PyC (IPyC), silicon carbide (SiC), and outer dense PyC (OPyC). Among the four coating layers, the SiC plays an important role beside in retaining metallic fission products, it also provides mechanical strength to fuel particle. However, results of post irradiation examination indicate that fission product palladium can react with and corrode SiC layer, This assessment is conducted to get the comprehension about resistance of SiC layer on irradiation effects, especially in order to increase the fuel bum-up. The result of this shows that the corrosion of SiC layer by fission product palladium is beside depend on the material characteristics of SiC, and also there are other factors that affect on the SiC layer corrosion. Fuel enrichment, bum-up, and irradiation time effect on the palladium flux in fuel kernel. While, the fuel density, vapour pressure of palladium (the degree depend on the irradiation temperature and kernel composition) effect on palladium migration in fuel particle. (author)

Corrosion behavior of spent MTR fuel elements in a drowned salt mine repository

International Nuclear Information System (INIS)

Brodda, B.G.; Fachinger, J.

1995-01-01

Spent MTR fuel from German Material Test Reactors will not be reprocessed, but stored in a final salt repository in the deep geologic underground. Fuel elements will be placed in POLLUX containers, which are assumed to resist the corrosive attack of an accidentally formed concentrated salt brine for about 500 years. After a container failure the brine would contact the fuel element, corrode the aluminum plating and possibly leach radionuclides from the fuel. A source term for the calculation of radionuclide mobilization results from the investigation of the behavior of MTR fuel in this scenario, which has to be considered for the long-term safety analysis of a deep mined rock salt repository. Experiments with the different plating materials show that the considered aluminum alloys will not resist the corrosive attack of a brine solution, especially in the presence of iron, under the conditions in a drowned salt mine repository. Although differences in the corrosion rates of about two orders of magnitude were observed when applying different parameter sets, the deterioration must be considered to be almost instantaneous in geological terms. Radionuclides are mobilized from irradiated MTR fuel, when the meat of the fuel element becomes accessible to the brine solution. It seems, however, that the radionuclides are effectively trapped by the aluminum hydroxide formed, as the activity concentrations in the brine solution soon reach a constant level with the progressing corrosion of the cladding aluminum. In the presence of iron a more significant initial release was observed, but also in this case an equilibrium activity seems to be reached as a consequence of radionuclide trapping

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

Advanced Silicon Carbide from Molecular Engineering and Actinide Fuels

International Nuclear Information System (INIS)

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

2008-01-01

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

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....... This results in potassium chloride being converted to potassium sulphate in the combustion chamber and it is sulphate rich deposits that are deposited on the vulnerable metallic surfaces such as high temperature superheaters. Although this removes the problem of chloride corrosion, other 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 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...

Studies on corrosion protection of laser hybrid welded AISI 316 by laser remelting

DEFF Research Database (Denmark)

Olsen, Flemming Ove; Ambat, Rajan; Rasmussen, A.J.

2005-01-01

laser surface melting on microstructure and corrosion behaviour of AISI 316L welds. Welding and laser treatment parameters were varied. General corrosion behaviour of the weld and laser treated surface was characterised using a gel visualization test. The local electrochemistry of the weld and laser......Unlike in autogenous laser welding, hybrid laser welding of stainless steel could introduce grain boundary carbides due to low cooling rates. Formation of grain boundary carbides leads to reduced corrosion properties. Studies have initially been carried out on hybrid laser welding and subsequent...... treated surface was investigated using a novel micro electrochemical technique with a tip resolution of ~1 mm. Results show that hybrid laser welding of 316L has increased corrosion susceptibility probably as a result of grain boundary carbide formation. However a suitable post laser treatment could...

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.

Modelling fireside corrosion of heat exchangers in co-fired pulverised fuel power systems

Energy Technology Data Exchange (ETDEWEB)

Simms, N.J. [Cranfield Univ. (United Kingdom). Energy Technology Centre; Fry, A.T. [National Physical Laboratory, Teddington, Middlesex (United Kingdom)

2010-07-01

As a result of concerns about the effects of CO{sub 2} emissions on the global environment, there is increasing pressure to reduce such emissions from power generation systems. The use of biomass co-firing with coal in conventional pulverised fuel power stations has provided the most immediate route to introduce a class of fuel that is regarded as both sustainable and carbon neutral. In the future it is anticipated that increased levels of biomass will need to be used in such systems to achieve the desired CO{sub 2} emission targets. However there are concerns over the risk of fireside corrosion damage to the various heat exchangers and boiler walls used in such systems. Future pulverised fuel power systems will need to be designed to cope with the effects of using a wide range of coal-biomass mixes. However, such systems will also need to use much higher heat exchanger operating temperatures to increase their conversion efficiencies and counter the effects of the CO{sub 2} capture technologies that will need to be used in them. Higher operating temperatures will also increase the risk of fireside corrosion damage to the critical heat exchangers. This paper reports work that has been carried out to develop quantitative corrosion models for heat exchangers in pulverised fuel power systems. These developments have been particularly targeted at producing models that enable the evaluation of the effects of using different coal-biomass mixtures and of increasing heat exchanger operating conditions. Models have been produced that have been targeted at operating conditions and materials used in (a) superheaters/reheaters and (b) waterwalls. Data used in the development of these models has been produced from full scale and pilot scale plants in the UK using a wide range of coal and biomass mixtures, as well as from carefully targeted series of laboratory corrosion tests. Mechanistic and neural network based models have been investigated during this development process to

Corrosion problems of PWR steam generators

International Nuclear Information System (INIS)

Urbancik, L.; Kostal, M.

Literature data are assessed on corrosion failures of steam generator tubes made of INCONEL 600 or INCOLOY 800. It was found that both alloys with high nickel content showed good stability in a corrosion environment while being sensitive to carbide formation on grain boundaries. The gradual depletion of chromium results from the material and corrosion resistance deteriorates. INCOLOY 800 whose chromium carbide precipitation on grain boundaries in pure water and steam is negligible up to 75O degC and which is not subject to corrosion attacks in the above media and in an oxidizing environment at a temperature to about 700 degC shows the best corrosion resistance. Its favourable properties were tested in long-term operation in the Peach Bottom 1 nuclear power plant where no failures due to corrosion of this material have been recorded since 1967. In view of oxygenic-acid surface corrosion, it is necessary to work in a neutral or slightly basic environment should any one of the two alloys be used for steam generator construction. The results are summed up of an analysis conducted for the Beznau I NOK reactor. Water treatment with ash-free amines can be used as prevention against chemical corrosion mechanisms, although the treatment itself does not ensure corrosion resistance of steam generator key components. (J.B.)

Fire-Side Corrosion: A Case Study of Failed Tubes of a Fossil Fuel Boiler

Directory of Open Access Journals (Sweden)

Majid Asnavandi

2017-01-01

Full Text Available The failures of superheater and reheater boiler tubes operating in a power plant utilizing natural gas or mazut as a fuel have been analysed and the fire-side corrosion has been suggested as the main reason for the failure in boiler tubes. The tubes have been provided by a fossil fuel power plant in Iran and optical and electron microscopy investigations have been performed on the tubes as well as the corrosion products on their surfaces. The results showed that the thickness of the failed tubes is not uniform which suggests that fire-side corrosion has happened on the tubes. Fire-side corrosion is caused by the reaction of combustion products with oxide layers on the tube surface resulting in metal loss and consequently tubes fracture. However, the tubes corrosion behaviour did not follow the conventional models of the fire-side corrosion. Given that, using the corrosion monitoring techniques for these boiler tubes seems essential. As a result, the thickness of the boiler tubes in different parts of the boiler has been recorded and critical points are selected accordingly. Such critical points are selected for installation of corrosion monitoring probes.

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

Formation and corrosion of a 410 SS/ceramic composite

Science.gov (United States)

Chen, X.; Ebert, W. L.; Indacochea, J. E.

2016-11-01

This study addressed the possible use of alloy/ceramic composite waste forms to immobilize metallic and oxide waste streams generated during the electrochemical reprocessing of spent reactor fuel using a single waste form. A representative composite material was made to evaluate the microstructure and corrosion behavior at alloy/ceramic interfaces by reacting 410 stainless steel with Zr, Mo, and a mixture of lanthanide oxides. Essentially all of the available Zr reacted with lanthanide oxides to generate lanthanide zirconates, which combined with the unreacted lanthanide oxides to form a porous ceramic network that filled with alloy to produce a composite puck. Alloy present in excess of the pore volume of the ceramic generated a metal bead on top of the puck. The alloys in the composite and forming the bead were both mixtures of martensite grains and ferrite grains bearing carbide precipitates; FeCrMo intermetallic phases also precipitated at ferrite grain boundaries within the composite puck. Micrometer-thick regions of ferrite surrounding the carbides were sensitized and corroded preferentially in electrochemical tests. The lanthanide oxides dissolved chemically, but the lanthanide zirconates did not dissolve and are suitable host phases. The presence of oxide phases did not affect corrosion of the neighboring alloy phases.

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)

Vanadium carbide coatings: deposition process and properties

International Nuclear Information System (INIS)

Borisova, A.; Borisov, Y.; Shavlovsky, E.; Mits, I.; Castermans, L.; Jongbloed, R.

2001-01-01

Vanadium carbide coatings on carbon and alloyed steels were produced by the method of diffusion saturation from the borax melt. Thickness of the vanadium carbide layer was 5-15 μm, depending upon the steel grade and diffusion saturation parameters. Microhardness was 20000-28000 MPa and wear resistance of the coatings under conditions of end face friction without lubrication against a mating body of WC-2Co was 15-20 times as high as that of boride coatings. Vanadium carbide coatings can operate in air at a temperature of up to 400 o C. They improve fatigue strength of carbon steels and decrease the rate of corrosion in sea and fresh water and in acid solutions. The use of vanadium carbide coatings for hardening of various types of tools, including cutting tools, allows their service life to be extended by a factor of 3 to 30. (author)

Corrosion of aluminum, uranium and plutonium in the presence of water in spent fuel storage tanks

International Nuclear Information System (INIS)

Grzetic, I.

1997-01-01

General problem associated with research reactor exploitation is safe storage of spent nuclear fuel. One of the possible solutions is its storage in aluminum containers filled and cooled with water. With time aluminum starts to corrode. The chemical corrosion of aluminum, as a heterogenous process, could be investigated in two ways. First, is direct investigation of Al corrosion per se, following hydrogen generation during the corrosion of Al in the presence of water. Both ways are based on available physico-chemical and thermodynamical data. Recent measurements of water quality in the Vinca Institute spent fuel pool clearly indicates that the particular case, corrosion is likely to be present. For the particular case, corrosion process could considered in two directions. The first one discusses the corrosion process of reactor fuel aluminum cladding in general. The second consideration is related with theoretically and empirically based calculations of hydrogen pressure in the closed aluminum containers in order to predict their resistance to the increased pressure. Finally, the corrosion of U, Pu and Cd is discussed with respect to solubility and influence of hydrogen on U and UO 2 under wet conditions. (author)

Tantalum and niobium carbides obtention by carbothermic reduction of columbotantalite ores

International Nuclear Information System (INIS)

Gordo, E.; Garcia-Carcedo, F.; Torralba, J.M.

1998-01-01

Tantalum and niobium carbides are characterized by its high hardness and chemical corrosion resistance. Both carbides, but mainly TaC, are used in hard metals (sintered carbides), together with their carbides, to manufacture cutting tools and dies in special machining applications involving mechanical shock at high temperature. Its use as reinforcement of wear resistant materials through powder metallurgy techniques are being investigated. However, the use of TaC is usually limited because of its high cost. Therefore tantalum carbide with niobium content, which is cheaper, is used. In this work the obtention of complex tantalum and niobium carbides from a Spanish columbotantalite ore is studied through relatively cheap and simple process as it is carbothermic reduction. Concentration of the ore, its reduction and the characterization of products are described. (Author) 11 refs

Fuel elements and safety engineering goals

International Nuclear Information System (INIS)

Schulten, R.; Bonnenberg, H.

1990-01-01

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

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...... undertaken. This results in potassium chloride being converted to potassium sulphate in the combustion chamber and it is sulphate rich deposits that are deposited on the vulnerable metallic surfaces such as high temperature superheaters. Although this removes the problem of chloride corrosion, other...... 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...

Corrosion of aluminium alloy test coupons in water of spent fuel storage pool at RA reactor

International Nuclear Information System (INIS)

Pesic, M.; Maksin, T.; Jordanov, G.; Dobrijevic, R.

2004-12-01

Study on corrosion of aluminium cladding, of the TVR-S type of enriched uranium spent fuel elements of the research reactor RA in the storage water pool is examined in the framework nr the International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP) 'Corrosion of Research Reactor Clad-Clad Spent Fuel in Water' since 2002. Standard racks with aluminium coupons are exposed to water in the spent fuel pools of the research reactor RA. After predetermined exposure times along with periodic monitoring of the water parameters, the coupons are examined according to the strategy and the protocol supplied by the IAEA. Description of the standard corrosion racks, experimental protocols, test procedures, water quality monitoring and compilation of results of visual examination of corrosion effects are present in this article. (author)

Evaluation of practicability of aluminosilicate additive fuel. Influence of aluminosilicate for reprocessing and corrosion of pellet

International Nuclear Information System (INIS)

Matsunaga, Junji; Kashibe, Shinji; Kinoshita, Mika; Ishimoto, Shinji; Harada, Kenichi

2014-01-01

Al-Si-O additive fuel is a modified pellet to improve the pellet-cladding interaction (PCI) resistance. This practicability assessment concerns the effect of Al-Si-O addition on the reprocessing and steam corrosion behavior. To address these concerns, a fuel dissolution test in nitric acid and a pellet corrosion test in humidified gas were carried out using the irradiated Al-Si-O additive fuel. Regardless of the Al-Si-O concentration, the dissolution rates of all Al-Si-O additive fuels were faster than that of the standard fuel. The morphology of the insoluble residue obtained from the irradiated Al-Si-O additive fuel could be considered as acceptable for retrieval by the clarification process using a conventional precipitation model. The corrosion resistance of the irradiated Al-Si-O additive fuel to high-temperature (at 1273 K) humidified gas was comparable to or better than that of the standard fuel. The result was interpreted as being due to a large grain size effect by Al-Si-O addition. (author)

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

International Nuclear Information System (INIS)

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

1993-01-01

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

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)

Atmospheric corrosion of uranium-carbon alloys; Corrosion atmospherique des alliages uranium-carbone

Energy Technology Data Exchange (ETDEWEB)

Rousset, P; Accary, A [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1965-07-01

The authors study the corrosion of uranium-carbon alloys having compositions close to that of the mono-carbide; they show that the extent of the observed corrosion effects increases with the water vapour content of the surrounding gas and they conclude that the atmospheric corrosion of these alloys is due essentially to the humidity of the air, the effect of the oxygen being very slight at room temperature. They show that the optimum conditions for preserving U-C alloys are either a vacuum or a perfectly dry argon atmosphere. The authors have also established that the type of corrosion involved is a corrosion which 'cracks under stress' and is transgranular (it can also be intergranular in the case of sub-stoichiometric alloys). They propose, finally, two hypotheses for explaining this mechanism, one of which is illustrated by the existence, at the fissure interface, of corrosion products which can play the role of 'corners' in the mono-carbide grains. (authors) [French] Les auteurs etudient la corrosion des alliages uranium-carbone de composition voisine du monocarbure; ils montrent que l'importance des effets de la corrosion observee augmente avec la teneur en vapeur d'eau du milieu gazeux ambiant et concluent que la corrosion atmospherique de ces alliages est due essentiellement a l'humidite de l'air, l'action de l'oxygene de l'air etant tres faible a la temperature ambiante. Ils indiquent que les conditions optimales de conservation des alliages U-C sont le vide ou une atmosphere d'argon parfaitement desseches. D'autre part, les auteurs etablissent que le type de corrosion mis en jeu est une corrosion 'fissurante sous contrainte', transgranulaire (pouvant egalement etre intergranulaire dans le cas d'alliages sous-stoechiometriques). Ils proposent enfin deux hypotheses pour rendre compte de ce mecanisme, dont l'une est illustree par la mise en evidence, a l'interface des fissures, de produits de corrosion pouvant jouer le role de 'coins' dans les grains de

Material characterization and corrosion control in wet storage of Chilean spent fuel

International Nuclear Information System (INIS)

Lamas, C.; Klein, J.; Escobar, I.

2002-01-01

Chile has two MTR type research reactors and the spent fuel will be stored in water previous to the conditioning for final disposal. One of the serious problem presented during wet storage is the phenomenon of corrosion, which depends on the water quality, the structural materials and the storage conditions. Thus, it is necessary to solve how to guarantee the integrity of the spent fuel during its wet storage. The water quality and fuel assembly materials are being characterized with the purpose to define the criteria of surveillance and control of corrosion as a function of time. The behavior of the 6061 Al and N4 Al alloys is being studied to characterize the susceptibility to pitting corrosion in solutions with chloride and cadmium as aggressive ions. The analyses were performed in a three-electrode electrochemical cell with 6061 Al and N4 Al as working electrodes. Platinum wire was the auxiliary electrode while Ag/AgCl was the reference electrode. To obtain the electrochemical characterization the polarization curves were used and the evolution of the corrosion potential of the aluminum alloys and SS 304 were measured. The electrolyte was deionized water with different concentrations of chloride and cadmium. At present, the results show that 6061 Al and N4 Al alloys are more susceptible to be attacked by pitting due to the presence of chloride than cadmium. (author)

Mechanism of protective action of surface carbide layers on titanium

International Nuclear Information System (INIS)

Chukalovskaya, T.V.; Chebotareva, N.P.; Tomashov, N.D.

1990-01-01

The protective action of surface carbide layer on titanium produced in methane atmosphere at 1000 deg C and under 6.7 kPa pressure in H 2 SO 4 solutions is studied through comparison of microsection metallographic specimens prior to and after corrosion testing (after specimen activation); through comparison of anodic characteristics after partial stripping of the layer up to its complete stripping; through analysis of the behaviour of Ti-TiC galvanic couple, and through investigation of corresponding corrosion diagrams under test conditions. It is shown that screening protective mechanism is primarily got involved in highly agressive media (high temperature and concentration of solution), and in less agressive environment the protection of titanium with carbide layer is primarily ensured by electrochemical mechanism

High-Temperature, Dual-Atmosphere Corrosion of Solid-Oxide Fuel Cell Interconnects

Science.gov (United States)

Gannon, Paul; Amendola, Roberta

2012-12-01

High-temperature corrosion of ferritic stainless steel (FSS) surfaces can be accelerated and anomalous when it is simultaneously subjected to different gaseous environments, e.g., when separating fuel (hydrogen) and oxidant (air) streams, in comparison with single-atmosphere exposures, e.g., air only. This so-called "dual-atmosphere" exposure is realized in many energy-conversion systems including turbines, boilers, gasifiers, heat exchangers, and particularly in intermediate temperature (600-800°C) planar solid-oxide fuel cell (SOFC) stacks. It is generally accepted that hydrogen transport through the FSS (plate or tube) and its subsequent integration into the growing air-side surface oxide layer can promote accelerated and anomalous corrosion—relative to single-atmosphere exposure—via defect chemistry changes, such as increased cation vacancy concentrations, decreased oxygen activity, and steam formation within the growing surface oxide layers. Establishment of a continuous and dense surface oxide layer on the fuel side of the FSS can inhibit hydrogen transport and the associated effects on the air side. Minor differences in FSS composition, microstructure, and surface conditions can all have dramatic influences on dual-atmosphere corrosion behaviors. This article reviews high-temperature, dual-atmosphere corrosion phenomena and discusses implications for SOFC stacks, related applications, and future research.

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.

Advanced Measurements of Silicon Carbide Ceramic Matrix Composites

Energy Technology Data Exchange (ETDEWEB)

Farhad Farzbod; Stephen J. Reese; Zilong Hua; Marat Khafizov; David H. Hurley

2012-08-01

Silicon carbide (SiC) is being considered as a fuel cladding material for accident tolerant fuel under the Light Water Reactor Sustainability (LWRS) Program sponsored by the Nuclear Energy Division of the Department of Energy. Silicon carbide has many potential advantages over traditional zirconium based cladding systems. These include high melting point, low susceptibility to corrosion, and low degradation of mechanical properties under neutron irradiation. In addition, ceramic matrix composites (CMCs) made from SiC have high mechanical toughness enabling these materials to withstand thermal and mechanical shock loading. However, many of the fundamental mechanical and thermal properties of SiC CMCs depend strongly on the fabrication process. As a result, extrapolating current materials science databases for these materials to nuclear applications is not possible. The “Advanced Measurements” work package under the LWRS fuels pathway is tasked with the development of measurement techniques that can characterize fundamental thermal and mechanical properties of SiC CMCs. An emphasis is being placed on development of characterization tools that can used for examination of fresh as well as irradiated samples. The work discuss in this report can be divided into two broad categories. The first involves the development of laser ultrasonic techniques to measure the elastic and yield properties and the second involves the development of laser-based techniques to measurement thermal transport properties. Emphasis has been placed on understanding the anisotropic and heterogeneous nature of SiC CMCs in regards to thermal and mechanical properties. The material properties characterized within this work package will be used as validation of advanced materials physics models of SiC CMCs developed under the LWRS fuels pathway. In addition, it is envisioned that similar measurement techniques can be used to provide process control and quality assurance as well as measurement of

Corrosion behaviour of boiler tube materials during combustion of fuels containing Zn and Pb

Energy Technology Data Exchange (ETDEWEB)

Bankiewicz, D.

2012-11-01

Many power plants burning challenging fuels such as waste-derived fuels experience failures of the superheaters and/or increased waterwall corrosion due to aggressive fuel components already at low temperatures. To minimize corrosion problems in waste-fired boilers, the steam temperature is currently kept at a relatively low level which drastically limits power production efficiency. The elements found in deposits of waste and waste-derived fuels burning boilers that are most frequently associated with high-temperature corrosion are: Cl, S, and there are also indications of Br; alkali metals, mainly K and Na, and heavy metals such as Pb and Zn. The low steam pressure and temperature in waste-fired boilers also influence the temperature of the waterwall steel which is nowadays kept in the range of 300 deg C - 400 deg C. Alkali chloride (KCl, NaCl) induced high-temperature corrosion has not been reported to be particularly relevant at such low material temperatures, but the presence of Zn and Pb compounds in the deposits have been found to induce corrosion already in the 300 deg C - 400 deg C temperature range. Upon combustion, Zn and Pb may react with Cl and S to form chlorides and sulphates in the flue gases. These specific heavy metal compounds are of special concern due to the formation of low melting salt mixtures. These low melting, gaseous or solid compounds are entrained in the flue gases and may stick or condense on colder surfaces of furnace walls and superheaters when passing the convective parts of the boiler, thereby forming an aggressive deposit. A deposit rich in heavy metal (Zn, Pb) chlorides and sulphates increases the risk for corrosion which can be additionally enhanced by the presence of a molten phase. The objective of this study was to obtain better insight into high-temperature corrosion induced by Zn and Pb and to estimate the behaviour and resistance of some boiler superheater and waterwall materials in environments rich in those heavy metals

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

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.

Corrosion of technical ceramics by molten aluminium

NARCIS (Netherlands)

Schwabe, U.; Wolff, L.R.; Loo, van F.J.J.; Ziegler, G.; With, de G.; Terpstra, R.A.; Metselaar, R.

1989-01-01

Corrosion investigations on various types of nonoxide technical ceramic materials, two types of silicon nitride (HIPRBSN and RBSN) and two types of silicon carbide (HIPSIC and SiSiC), were carried out in aluminum (Al99.99) melts. HIPRBSN showed nearly no corrosion attack under the most severe

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.

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.

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.

Corrosion of Spent Nuclear Fuel: The Long-Term Assessment

International Nuclear Information System (INIS)

Ewing, Rodney C.

2003-01-01

This research program is a broadly based effort to understand the long-term behavior of spent nuclear fuel (SNF) and its alteration products in a geologic repository. We have established by experiments and field studies that natural uraninite, UO2+x, and its alteration products are excellent ''natural analogues'' for the study of the corrosion of UO2 in SNF. This on-going research program has addressed the following major issues: (1) What are the long-term corrosion products of natural UO2+x, uraninite, under oxidizing and reducing conditions? (2) What is the paragenesis or the reaction path for the phases that form during alteration? (3) What is the radionuclide content in the corrosion products as compared with the original UO2+x? Do the trace element contents substantiate models developed to predict radionuclide incorporation into the secondary phases? Are the corrosion products accurately predicted from geochemical codes (e.g., EQ3/6 or Geochemist's Workbench) that are used in performance assessments? Can these codes be tested by studies of natural analogue sites (e.g., Oklo, Cigar Lake or Pena Blanca)

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

International Nuclear Information System (INIS)

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

2005-01-01

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

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.

Copper anode corrosion affects power generation in microbial fuel cells

KAUST Repository

Zhu, Xiuping; Logan, Bruce E.

2013-01-01

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.

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%

Calculation of vapour pressures over mixed carbide fuels

International Nuclear Information System (INIS)

Joseph, M.; Mathews, C.K.

1988-01-01

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

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)

Apparatus for surface treatment of U-Pu carbide fuel samples

International Nuclear Information System (INIS)

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

1979-05-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1982-05-19

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

Mechanisms of metal dusting corrosion

DEFF Research Database (Denmark)

Hummelshøj, Thomas Strabo

In this thesis the early stages of metal dusting corrosion is addressed; the development of carbon expanded austenite, C, and the decomposition hereof into carbides. Later stages of metal dusting corrosion are explored by a systematic study of stainless steel foils exposed to metal dusting...... deformed stainless steel flakes is transformed to expanded martensite/austenite during low-temperature carburization. Various experimental procedures to experimentally determine the concentration dependent diffusion coefficient of carbon in expanded austenite are evaluated. The most promising procedure...... powders and flakes. The nature of the decomposition products, carbides of the form M23C6 and M7C3, were evaluated by X-ray diffraction, light optical microscopy, scanning electron microscopy and thermodynamic modelling. The decomposition was found to be dependent on several parameters such as thermal...

Fire-Side Corrosion: A Case Study of Failed Tubes of a Fossil Fuel Boiler

OpenAIRE

Asnavandi, Majid; Kahram, Mohaddeseh; Rezaei, Milad; Rezakhani, Davar

2017-01-01

The failures of superheater and reheater boiler tubes operating in a power plant utilizing natural gas or mazut as a fuel have been analysed and the fire-side corrosion has been suggested as the main reason for the failure in boiler tubes. The tubes have been provided by a fossil fuel power plant in Iran and optical and electron microscopy investigations have been performed on the tubes as well as the corrosion products on their surfaces. The results showed that the thickness of the failed tu...

Corrosion behaviour of WC-VC-Co hardmetals in acidic media

CSIR Research Space (South Africa)

Konadu, DS

2010-09-01

Full Text Available The effect of increasing vanadium carbide (VC) content on the corrosion behaviour of tungsten carbide – 10 wt% cobalt hard metals was investigated in 1 M hydrochloric (HCl), and sulphuric (H2SO4) acids solutions. Increasing VC content makes the open...

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

The SKB spent fuel corrosion programme. An evaluation of results from the experimental programme performed in the Studsvik Hot Cell Laboratory

Energy Technology Data Exchange (ETDEWEB)

Forsyth, R. [Forsyth Consulting, Nykoeping (Sweden)

1997-12-01

During the last few years, many of the specimens in the SKB programme on the corrosion of spent fuel have been analysed by the ICP-MS technique, shortly after conclusion of the corrosion tests, or by the analysis of archive samples. Together with the previous results, this has made available a much more extended analytical data base than that available before, and this has been used in a new evaluation which complements those published earlier. Some of the new analytical data is for tests performed on fuel specimens (from two reference fuel rods, one BWR and one PWR) which have been corrosion tested for over ten years. Most of the data refers to 16 fuel/clad specimens from a short BWR fuel rod, which had burnups over a range of 27.0 to 48.8 MWd/kg U. Detailed examination and characterisation of three other fuel specimens from the rod had shown that the specimens with the higher burnups in this series would have a fuel microstructure and alpha activity content and distribution which, theoretically, may promote enhanced corrosion. These specimens had been exposed to over 5 years of corrosion during nine water contact periods. The corrodants used were a simulated bicarbonate groundwater and de-ionised water, and both oxic and nominally anoxic conditions were included in the test matrix. Most of the emphasis in the evaluation has, therefore, been on the possible effects on corrosion behaviour of the linear heat rating and burnup of the fuels. However, examination of the variation with water contact time of the fractional release rates of selected fission products and their total release over the five years of corrosion, have shown that the corrosion rates during the first few weeks of corrosion of the specimens with the higher burnups were lower than those for specimens with slightly lower burnup. Later, the corrosion rates converged for all specimens. This has been interpreted to be due to burnup-related differences in the fuel microstructure, particularly in the

Pitting corrosion resistance of high alloy OCTG in ferric chloride solution

International Nuclear Information System (INIS)

Masamura, K.; Yamamoto, S.; Matsushima, I.

1986-01-01

The effects of alloying elements and precipitated phases on the corrosion rate of high alloy OCTG in the ferric chloride solution have been evaluated. The corrosion rate of Fe-Cr-Ni-Mo alloys without precipitated phases, e.g. carbides and sigma phase, can be estimated from the composition using the following equation: log(C.R.)=-0.144xPRE-7690/(273+T)+28.6 where C.R. is the corrosion rate in g/m/sup 2//hr; PRE is Cr+3Mo+16N in percent and T is the test temperature in 0 C. The activation energies of the ferric chloride test are almost the same regardless of PRE or Ni content when no detrimental phase precipitates. When carbides or the sigma phase precipitate, the corrosion rate is higher and the activation energy is lowered. This suggests that secondary phases give preferential sites for initiation of pitting corrosion

Mechanistic modelling of the corrosion behaviour of copper nuclear fuel waste containers

Energy Technology Data Exchange (ETDEWEB)

King, F; Kolar, M

1996-10-01

A mechanistic model has been developed to predict the long-term corrosion behaviour of copper nuclear fuel waste containers in a Canadian disposal vault. The model is based on a detailed description of the electrochemical, chemical, adsorption and mass-transport processes involved in the uniform corrosion of copper, developed from the results of an extensive experimental program. Predictions from the model are compared with the results of some of these experiments and with observations from a bronze cannon submerged in seawater saturated clay sediments. Quantitative comparisons are made between the observed and predicted corrosion potential, corrosion rate and copper concentration profiles adjacent to the corroding surface, as a way of validating the long-term model predictions. (author). 12 refs., 5 figs.

Effect of silicon carbide addition on the corrosion behavior of powder metallurgy Cu−30Zn brass in a 3.5 wt% NaCl solution

Energy Technology Data Exchange (ETDEWEB)

Almomani, Mohammed Ali, E-mail: maalmomani7@just.edu.jo [Department of Industrial Engineering, Jordan University of Science and Technology, P. O. Box 3030, Irbid 22110 (Jordan); Tyfour, Wa' il Radwan, E-mail: wrtyfou@just.edu.jo [Department of Industrial Engineering, Jordan University of Science and Technology, P. O. Box 3030, Irbid 22110 (Jordan); Nemrat, Mohammed Hani, E-mail: mohammednemrat@yahoo.com [Department of Mechanical Engineering, Institute of Applied Technology, Abu Dhabi (United Arab Emirates)

2016-09-15

A study was made to evaluate the corrosion behavior when Cu−30Zn alloy is reinforced with different weight fractions of silicon carbide (SiC) particles in a simulated sea solution (3.5 wt% NaCl aqueous solution). The composites were produced via powder metallurgy (PM) route. For the sake of comparison, the corrosion behaviors of unreinforced and reinforced alloy were examined. Electrochemical measurements (potentiodynamic testing) showed that the corrosion rate of the composites decreased with increase of SiC weight percentages, as a result of weak microgalvanic couple between reinforcement particles and Cu−30Zn matrix, and the low possibility of intermetallic phases formation. ANOVA test indicated that the variations of corrosion rate of the composites upon changing weight percentages of SiC particles are statistically significant. Polarization curves showed that the passive film tends to be less stable, and the potential difference between passivation and free corrosion points increased with increase of SiC weight percentages, as SiC cathodically protect the matrix material by sacrificial anodic dissolution of crevice regions about reinforcement particles. Scanning Electron Microscope (SEM) images of the sample's surfaces before and after testing are in agreement with the electrochemical results. - Highlights: • Effect of adding SiC on both uniform and localized corrosion of Cu−30Zn alloy is studied. • Reinforcing Cu−30Zn with nonconductive SiC particles decreases its tendency to uniform and localized corrosion. • Reinforcement particles cathodically protect the matrix material, and retard pit propagation to the matrix.

Carbide precipitation in the heat affected zone of a GTA weld in 21-6-9 stainless steel

International Nuclear Information System (INIS)

Carr, M.J.; Thorvaldson, W.G.

1979-01-01

Grain boundary precipitation was observed in a multipass GTA weld in 21-6-9 stainless steel. The precipitate was identified by electron diffraction as M 23 C 6 -type carbide. The presence of these carbide particles did not cause intergranular attack in standard corrosion tests

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.

Corrosion of copper under Canadian nuclear fuel waste disposal conditions

International Nuclear Information System (INIS)

King, F.; Litke, C.D.

1990-01-01

The corrosion of copper was studied under Canadian nuclear fuel waste disposal conditions. The groundwater in a Canadian waste vault is expected to be saline, with chloride concentrations from 0.1 to 1.0 mol/l. The container would be packed in a sand/clay buffer, and the maximum temperature on the copper surface would be 100C; tests were performed up to 150C. Radiation fields will initially be around 500 rad/h, and conditions will be oxidizing. Sulfides may be present. The minimum design lifetime for the container is 500 years. Most work has been done on uniform corrosion, although pitting has been considered. It was found that the rate of uniform corrosion in aerated NaCl at room temperature is limited by the rate of the anodic reaction, which is controlled mainly by the rate of transport of dissolved metal species away from the copper surface. The rate of corrosion should become controlled by the transport of oxygen to the copper surface only at very low oxygen concentrations. In the presence of gamma radiation the corrosion rate may never become cathodically transport limited. In compacted buffer material, the corrosion rate appears to be limited by the rate of transport of copper species away from the corroding surface. The authors recommend that long-term predictions of container lifetime should be based on the known rate-determining step for the overall corrosion process. 8 refs

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)

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)

The SKB spent fuel corrosion programme. Status report 1988

International Nuclear Information System (INIS)

Werme, L.O.; Forsyth, R.S.

1989-05-01

The results of the Swedish spent fuel corrosion programme from 1982 to 1988 are reviewed. Areas where additional research will be required are indentified. The major findings and conclusions after the first six years of the programme are that uranium attains relatively rapidly a constant solution concentration of about 1 mg/l. This is probably solubility controlled. Also plutonium, after initially higher concentration appear to reach a constant concentration of about 0.3 μg/l in groundwater. In DI water, the normalized Pu release is higher than the U release, indicating ongoing fuel oxidation/alteration after the leachant has been saturated with U. Under reducing conditions, the absence of fuel oxidation and the very low U solubility lead to a stronger tendency towards congruent releases, controlled by the solubility of the fuel matrix. The fission products Cs, Sb, Tc and Mo appear to selectively leached, probably from inclusions or from fuel cracks, fissures and grain boundaries. (authors)

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

International Nuclear Information System (INIS)

Saravia, Alvaro

2015-01-01

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

Effects of applying an external magnetic field during the deep cryogenic heat treatment on the corrosion resistance and wear behavior of 1.2080 tool steel

International Nuclear Information System (INIS)

Akhbarizadeh, Amin; Amini, Kamran; Javadpour, Sirus

2012-01-01

Highlights: ► Deep cryogenic increases the carbide percentage and make a more homogenous distribution. ► Deep cryogenic improve the wear resistance and corrosion behavior of 1.2080 tool steel. ► Applying the magnetic field weaker the carbide distribution and decreases the carbides percentage. ► Magnetized samples showed weaker corrosion and wear behavior. -- Abstract: This work concerns with the effect of applying an external magnetic field on the corrosion behavior, wear resistance and microstructure of 1.2080 (D2) tool steel during the deep cryogenic heat treatment. These analyses were performed via scanning electron microscope (SEM), optical microscope (OM), transmission electron microscope (TEM) and X-ay diffraction (XRD) to study the microstructure, a pin-on-disk wear testing machine to study the wear behavior, and linear sweep voltammetry to study the corrosion behavior of the samples. It was shown that the deep cryogenic heat treatment eliminates retained austenite and makes a more uniform carbide distribution with higher percentage. It was also observed that the deep cryogenic heat treatment improves the wear behavior and corrosion resistance of 1.2080 tool steel. In comparison between the magnetized and non-magnetized samples, the carbide percentage decreases and the carbide distribution weakened in the magnetized samples; subsequently, the wear behavior and corrosion resistance attenuated compared in the magnetized samples.

Long-time corrosion and high-temperature oxidation of zirconium alloys applied on NPP like fuel elements cover

International Nuclear Information System (INIS)

Vrtilkova, V.; Novotny, L.; Lingart, S.; Doukha, R.; Yarosh, Ya.; Kolenchik, Ya.

2007-01-01

Zirconium is applying in nuclear energy since 50-th of last century in capacity of material for cover production for fuel elements, reactor fuel and structural parts, and mainly due to both corrosion stability and low effective cross section for thermal neutrons capture. Impurities in doping elements form and alloy production technology has influence on mechanical and corrosion properties of finite alloy. Long-time corrosion tests for several zirconium alloys in forcing autoclave under different reaction conditions were carried out. After that process kinetics was studied, mass increase, hydrogen formation, zirconium hydride forming morphology, zirconium oxide layer thickness have been determined as well

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

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.

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)

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

Science.gov (United States)

2010-08-01

and B20) In experiments with additions of distilled water, all fuels supported biofilm formation Changes in the water pH did not correlate with...ULSD) and blends of ULSD and B100 (B5 and B20). In experiments with additions of distilled water, all fuels supported biofilm formation . Changes in...corrosion; diesel; biodiesel; biofouling; MIC; carbon steel ; aluminum; stainless steel Introduction Microbial contamination of hydrocarbon fuels

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

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.

The role of fission products (noble metal particles) in spent fuel corrosion process in a failed container

Energy Technology Data Exchange (ETDEWEB)

Wu, L., E-mail: lwu59@uwo.ca [Univ. of Western Ontario, Dept. of Chemistry, London, Ontario (Canada); Shoesmith, D.W. [Univ. of Western Ontario, Dept. of Chemistry, London, Ontario (Canada); Univ. of Western Ontario, Surface Science Western, London, Ontario (Canada)

2013-07-01

The corrosion/dissolution of simulated spent fuel has been studied electrochemically. Fission products within the UO{sub 2} matrix are found to have significant effect on the anodic dissolution behaviour of the fuel. It is observed that H{sub 2}O{sub 2}oxidation is accelerated on the surfaces of doped noble metal (ε) particles existing in the fuel matrix. It is concluded that H{sub 2}O{sub 2} decomposition rather than UO{sub 2} corrosion should be the dominant reaction under high H{sub 2}O{sub 2} concentrations. (author)

Evaluation of aluminum-clad spent fuel corrosion in Argentine basins

International Nuclear Information System (INIS)

Haddad, R.; Loberse, A.N.; Semino, C.J.; Guasp, R.

2001-01-01

An IAEA sponsored Coordinated Research Program was extended to study corrosion effects in several sites. Racks containing Aluminum samples were placed in different positions of each basin and periodic sampling of all the waters was performed to conduct chemical analysis. Different forms of corrosion have been encountered during the programme. In general, the degree of degradation is inversely proportional to the purity of the water. Maximum pit depths after 2 years of exposure are in the range of 100-200 μm. However, sediments deposited on the coupon surfaces seem to be responsible for the developing of large pits (1-2 mm in diameter). In many cases, what appears to be iron oxide particles were found originated by the corrosion of carbon steel components present elsewhere in the basin. These results correlate with observations made on the fuel itself, during exhaustive visual inspection. (author)

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

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

Effect of Zr addition on intergranular corrosion of low-chromium ferritic stainless steel

International Nuclear Information System (INIS)

Park, Jin Ho; Kim, Jeong Kil; Lee, Bong Ho; Seo, Hyung Suk; Kim, Kyoo Young

2014-01-01

Addition of Zr to low-Cr ferritic stainless steel forms a mixture of ZrC and Fe 23 Zr 6 precipitates that can prevent intergranular corrosion. Transmission electron microscopy and three-dimensional atom probe analysis suggest that the ZrC and Fe 23 Zr 6 mixture prevents intergranular corrosion in two ways: by acting as a strong carbide former to suppress the formation of Cr-carbide and by acting as a barrier against the diffusion of the solute Cr towards the grain boundary

FUNDAMENTAL MECHANISMS OF CORROSION OF ADVANCED LIGHT WATER REACTOR FUEL CLADDING ALLOYS AT HIGH BURNUP

International Nuclear Information System (INIS)

Lott, Randy G.

2003-01-01

OAK (B204) The corrosion behavior of nuclear fuel cladding is a key factor limiting the performance of nuclear fuel elements, improved cladding alloys, which resist corrosion and radiation damage, will facilitate higher burnup core designs. The objective of this project is to understand the mechanisms by which alloy composition, heat treatment and microstructure affect corrosion rate. This knowledge can be used to predict the behavior of existing alloys outside the current experience base (for example, at high burn-up) and predict the effects of changes in operation conditions on zirconium alloy behavior. Zirconium alloys corrode by the formation f a highly adherent protective oxide layer. The working hypothesis of this project is that alloy composition, microstructure and heat treatment affect corrosion rates through their effect on the protective oxide structure and ion transport properties. The experimental task in this project is to identify these differences and understand how they affect corrosion behavior. To do this, several microstructural examination techniques including transmission electron microscope (TEM), electrochemical impedance spectroscopy (EIS) and a selection of fluorescence and diffraction techniques using synchrotron radiation at the Advanced Photon Source (APS) were employed

Development of an Alternative Corrosion Inhibitor for the Storage of Advanced Gas-Cooled Reactor Fuel

International Nuclear Information System (INIS)

Standring, P.N.; Hands, B.J.; Morgan, S.; Brooks, A.

2015-01-01

Sellafield Lt. currently stores AGR fuel in sodium hyrodxide dosed pool water to pH 11.5 to prevent susceptible AGR fuel from failing due to inter-granular attack. The exception to the above storage practice is Thorp Receipt and Storage (TR&S) where an AGR reprocessing buffer is stored in demineralised water as the expected storage durations were short term (up to 5 years). With the extended shut-down of Thorp, storage durations have increased and this has prompted a re-evaluation of the AGR storage regime in TR&S. The use of sodium hydroxide is not feasible due to a compatibility issue with aluminum components used in LWR storage furniture. The implementation process adopted by Sellafield Ltd in developing an alternative corrosion inhibitor for spent AGR fuel is outlined. The two stranded approach evaluates the impact of candidate corrosion inhibitors on fuel integrity and on plant and processes. The development studies in support of the fuel integrity strand are reported. Candidate inhibitors were first evaluated inactively in terms of their ability to arrest propagating corrosion, radiation stability, compatibility with aluminium and environmental impact. Sodium Nitrate was concluded to be the most promising inhibitor. Sodium nitrate was subsequently tested with active AGR brace material. These studies involved the use of bespoke test equipment and techniques. The studies demonstrated that propagating corrosion could be arrested using 10 ppm nitrate and showed that the resultant nitrate film required relatively high chloride concentrations to break it down over the study duration of 60 days. The development studies to date have provided the confidence that sodium nitrate has the potential to be an effective inhibitor for AGR fuel. The final phase of the fuel integrity strand involves a Lead Container Study using whole AGR pins. A staged approach is being adopted in the study programme where proceeding to a more onerous study is not progressed until positive

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

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.)

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)

Optimization of uranium carbide fabrication by carbothermic reduction with limited oxygen content

International Nuclear Information System (INIS)

Raveu, Gaelle

2014-01-01

Mixed carbides (U, Pu)C, are good fuel candidate for generation IV reactors because of their high fissile atoms density and excellent thermal properties for economical (more compact and efficient cores) and safety reasons (high melting margin). UC can be imagine as a surrogate material ror R and D studies on (U,Pu)C fuel behavior, because of their similar structures. The carbothermic reaction was used because it is the most studied and now consider for industrial process. However, it involves powders manipulation: in air, carbide can strongly react at room temperature and under controlled atmosphere it can absorb impurities. An inerted installation under Ar, BaGCARA, was therefore used. Process improvements were carried out, including the sintering atmosphere in order to evaluate the impact on the sample purity (about oxygen content). The original method by ion beam analysis was used to determine the surface composition (oxygen in-depth profiles in the first microns and stoichiometry). This oxygen analysis was set for the first time in carbonaceous materials. XRD analysis showed the formation of an intermediate compound during the carbothermic reaction and a better crystallization of the samples fabricated in BaGCARA. They also have a better microstructure, density, and visual appearance if compared to former samples. Vacuum sintering leads to a denser UC with fewer second phases if compared to Ar, Ar/H 2 or controlled PC atmospheres. However, it was not possible to analyze carbides without air contact which may impact their lattice parameter and lead to their deterioration. When the carbide is initially free of oxygen, it oxidizes faster, more intensely and heterogeneously. The mechanical stress induced between the grains lead to fracturing the material, to corrosion cracking and then a de-bonding of the material. A study of oxidation mechanisms would be interesting to validate and understand the evolution of the material in contact with oxygen. A study of the

Corrosion of assemblies in fuel-storage basins at Savannah River Plant

International Nuclear Information System (INIS)

Wollam, C.D.

1980-09-01

Pitting of reactor assemblies has been the major corrosion problem in the Savannah River Plant fuel storage basins. From 1972 to 1976 many reactor assemblies experienced severe pitting corrosion with rates up to 9.3 mm/y. Poor cladding, high concentrations of iron and chloride ions in the water, a galvanic couple between the aluminum clad assemblies and the stainless steel hangers, and scratches in the oxide layer on assemblies have been identified as contributors to the problem. This paper describes the examinations and tests that were conducted and discusses a theory that explains the observed phenomena

Sulfide Production and Corrosion in Seawater During Exposure to FAME Alternative Fuel

Science.gov (United States)

2012-06-01

transporting fatty acid methyl ester ( FAME ] alternative diesel fuel in contact with natural seawater under anaerobic conditions. Coastal Key West...Glycerol Fatty Acid Methyl Ester Exposure Chamber Anaerobic Chamber - bal. N2,10% H2, 0.1% C02 - maintain pH ~8 Polarization Resistance (Rp...and Corrosion in Seawater During Exposure to FAME Alternative Fuel Jason 5. Lee Richard I. Ray BrendaJ. Little Naval Research Laboratory Stennis

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.

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

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

National Research Council Canada - National Science Library

Parker, Donald S

2003-01-01

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

A study on the corrosion rate for metal nuclear fuel by the soxhlet

International Nuclear Information System (INIS)

Oh, S. J.; Lee, Y. R.; Lee, D. B.; Park, J. M.; Kim, K. H.; Lee, Y. S.; Park, H. D.; Kim, C. K.

2002-01-01

In order to compare in-pile performance of nuclear fuel candidates for HANARO, corrosion test with the Soxhlet apparatus for rare-earth-oxide added U-Mo alloy fuels has been carried out by measuring a leaching rate. It appeared from the result that the leaching rate of the U-Mo fuel specimen became decreased as a rare-earth-oxide added, and there was a little difference in the leaching rate depending on the kind of the rare-earth-oxide

Intergranular Corrosion of 316L Stainless Steel by Aging and UNSM (Ultrasonic Nano-crystal Surface Modification) treatment

International Nuclear Information System (INIS)

Lee, J. H.; Kim, Y. S.

2015-01-01

Austenitic stainless steels have been widely used in many engineering fields because of their high corrosion resistance and good mechanical properties. However, welding or aging treatment may induce intergranular corrosion, stress corrosion cracking, pitting, etc. Since these types of corrosion are closely related to the formation of chromium carbide in grain boundaries, the alloys are controlled using methods such as lowering the carbon content, solution heat treatment, alloying of stabilization elements, and grain boundary engineering. This work focused on the effects of aging and UNSM (Ultrasonic Nano-crystal Surface Modification) on the intergranular corrosion of commercial 316L stainless steel and the results are discussed on the basis of the sensitization by chromium carbide formation and carbon segregation, residual stress, grain refinement, and grain boundary engineering

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

International Nuclear Information System (INIS)

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

1997-04-01

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

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)

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

Corrosion resistance characteristics of stamped and hydroformed proton exchange membrane fuel cell metallic bipolar plates

Energy Technology Data Exchange (ETDEWEB)

Dundar, F. [NSF I/UCRC Center for Precision Forming (CPF), Virginia Commonwealth University, Richmond, VA (United States); Department of Materials Science and Engineering, Gebze Institute of Technology (Turkey); Dur, Ender; Koc, M. [NSF I/UCRC Center for Precision Forming (CPF), Virginia Commonwealth University, Richmond, VA (United States); Mahabunphachai, S. [NSF I/UCRC Center for Precision Forming (CPF), Virginia Commonwealth University, Richmond, VA (United States); National Metal and Materials Technology Center (MTEC), Pathumthani (Thailand)

2010-06-01

Metallic bipolar plates have several advantages over bipolar plates made from graphite and composites due to their high conductivity, low material and production costs. Moreover, thin bipolar plates are possible with metallic alloys, and hence low fuel cell stack volume and mass are. Among existing fabrication methods for metallic bipolar plates, stamping and hydroforming are seen as prominent approaches for mass production scales. In this study, the effects of important process parameters of these manufacturing processes on the corrosion resistance of metallic bipolar plates made of SS304 were investigated. Specifically, the effects of punch speed, pressure rate, stamping force and hydroforming pressure were studied as they were considered to inevitably affect the bipolar plate micro-channel dimensions, surface topography, and hence the corrosion resistance. Corrosion resistance under real fuel cell conditions was examined using both potentiodynamic and potentiostatic experiments. The majority of the results exhibited a reduction in the corrosion resistance for both stamped and hydroformed plates when compared with non-deformed blank plates of SS304. In addition, it was observed that there exist an optimal process window for punch speed in stamping and the pressure rate in hydroforming to achieve improved corrosion resistance at a faster production rate. (author)

Tribological behaviors of graphite sliding against cemented carbide in CaCl2 solution

International Nuclear Information System (INIS)

Guo, Fei; Tian, Yu; Liu, Ying; Wang, Yuming

2015-01-01

The tribological behaviors of graphite sliding against cemented carbide were investigated using a standard tribological tester Plint TE92 in a ring-on-ring contact configuration in both CaCl 2 solution and deionized water. An interesting phenomenon occurred: as the CaCl 2 solution concentration increased, the friction coefficient firstly decreased and was lower than that in the deionized water, and then gradually increased, exceeding the friction coefficient in the deionized water. The wear rate of the ,graphite also presented the same variation trend. According to the polarization curves of cemented carbide, contact angle measurements, Raman spectrum analysis and scanning electron microscope (SEM) images analysis, the above friction and wear behaviors of graphite sliding against cemented carbide were attributed to the graphite surface wettability and the cemented carbide surface corrosion property. (paper)

Inhibition Effect of Deanol on Mild Steel Corrosion in Dilute ...

African Journals Online (AJOL)

NICOLAAS

2014-06-23

Jun 23, 2014 ... The influence of deanol on the corrosion behaviour of mild steel in dilute sulphuric acid with sodium ... the formation of a complex precipitate of protective film, which ... silicon carbide abrasive papers of 80, 120, 220, 800 and 1000 grit ...... ions in sulphuric acid on the corrosion behaviour of stainless steel,.

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

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

International Nuclear Information System (INIS)

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

2011-01-01

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

UK experience on fuel and cladding interaction in oxide fuels

Energy Technology Data Exchange (ETDEWEB)

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

1977-04-01

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

UK experience on fuel and cladding interaction in oxide fuels

International Nuclear Information System (INIS)

Batey, W.; Findlay, J.R.

1977-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2008-07-01

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

Corrosion product deposition on fuel element surfaces of a boiling water reactor

International Nuclear Information System (INIS)

Orlov, A.

2011-01-01

Over the last decade the problem of corrosion products deposition on light water reactor fuel elements has been extensively investigated in relation to the possibility of failures caused by them. The goal of the present study is to understand in a quantitative way the formation of such kind of deposits and to analytically understand the mechanism of formation and deposition with help of the quasi-steady state concentrations of a number of 3d metals in reactor water. Recent investigations on the complex corrosion product deposits on a Boiling Water Reactor (BWR) fuel cladding have shown that the observed layer locally presents unexpected magnetic properties. The buildup of magnetic corrosion product deposits (crud) on the fuel cladding of the BWR, Kernkraftwerk Leibstadt (KKL) Switzerland has hampered the Eddy-current based measurements of ZrO 2 layer thickness. The magnetic behavior of this layer and its axial variation on BWR fuel cladding is of interest with respect to non-destructive cladding characterization. Consequently, a cladding from a BWR was cut at elevations of 810 mm, where the layer was observed to be magnetic, and of 1810 mm where it was less magnetic. The samples were subsequently analyzed using electron probe microanalysis (EPMA), magnetic analysis and X-ray techniques (μXRF, μXRD and μXAFS). Both EPMA and μXRF have shown that the observed corrosion deposit layer which is situated on the Zircaloy corrosion layer consists mostly of 3-d elements’ oxides (Fe, Zn, Ni and Mn). The distribution of these elements within the investigated layer is rather complex and not homogeneous. The main components identified by 2D μXRD mapping inside the layer were hematite and spinel phases with the common formula (M x Fe y )[M (1-x) Fe (2-y) ]O 4 , where M = Zn, Ni, Mn. With μXRD it was clearly shown that the cell parameter of analyzed spinel is different from the one of the pure endmembers (ZnFe 2 O 4 , NiFe 2 O 4 and MnFe 2 O 4 ) proving the existence of

Testing Systems and Results for Advanced Nuclear Fuel Materials

International Nuclear Information System (INIS)

Rooyen, I.J. van; Griffith, G.W.; Garnier, J.E.

2012-01-01

Light Water Reactor Sustainability (LWRS) Program Advanced LWR Nuclear Fuel Development (ALFD) Pathway. Development and testing of high performance fuel cladding identified as high priority to support: enhancement of fuel performance, reliability, and reactor safety. One of the technologies being examined is an advanced fuel cladding made from ceramic matrix composites (CMC) utilizing silicon carbide (SiC) as a structural material supplementing a commercial Zircaloy-4 (Zr-4) tube. A series of out-of-pile tests to fully characterize the SiC CMC hybrid design to produce baseline data. The planned tests are intended to either produce quantitative data or to demonstrate the properties required to achieve two initial performance conditions relative to standard zircaloybased cladding: decreased hydrogen uptake (corrosion) and decreased fretting of the cladding tube under normal operating and postulated accident conditions. These two failure mechanisms account for approximately 70% of all in-pile failures of LWR commercial fuel assemblies

Iodine-induced stress corrosion cracking of fixed deflection stressed slotted rings of Zircaloy fuel cladding

International Nuclear Information System (INIS)

Sejnoha, R.; Wood, J.C.

1978-01-01

Stress corrosion cracking of Zircaloy fuel cladding by fission products is thought to be an important mechanism influencing power ramping defects of water-reactor fuels. We have used the fixed-deflection stressed slotted-ring technique to demonstrate cracking. The results show both the sensitivity and limitations of the stressed slotted-ring method in determining the responses of tubing to stress corrosion cracking. They are interpreted in terms of stress relaxation behavior, both on a microscopic scale for hydrogen-induced stress-relief and on a macroscopic scale for stress-time characteristics. Analysis also takes account of nonuniform plastic deformation during loading and residual stress buildup on unloading. 27 refs

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.)

Corrosion resistant composite materials

International Nuclear Information System (INIS)

Ul'yanin, E.A.

1986-01-01

Foundations for corrosion-resistant composite materials design are considered with account of components compatibility. Fibrous and lamellar composites with metal matrix, dispersion-hardened steels and alloys, refractory metal carbides-, borides-, nitrides-, silicides-based composites are described. Cermet compositions and fields of their application, such as protective coatings for operation in agressive media at high temperatures, are presented

Modelling of iodine-induced stress corrosion cracking in CANDU fuel

International Nuclear Information System (INIS)

Lewis, B.J.; Thompson, W.T.; Kleczek, M.R.; Shaheen, K.; Juhas, M.; Iglesias, F.C.

2011-01-01

Iodine-induced stress corrosion cracking (I-SCC) is a recognized factor for fuel-element failure in the operation of nuclear reactors requiring the implementation of mitigation measures. I-SCC is believed to depend on certain factors such as iodine concentration, oxide layer type and thickness on the fuel sheath, irradiation history, metallurgical parameters related to sheath like texture and microstructure, and the mechanical properties of zirconium alloys. This work details the development of a thermodynamics and mechanistic treatment accounting for the iodine chemistry and kinetics in the fuel-to-sheath gap and its influence on I-SCC phenomena. The governing transport equations for the model are solved with a finite-element technique using the COMSOL Multiphysics (registered) commercial software platform. Based on this analysis, this study also proposes potential remedies for I-SCC.

Improvements to the corrosion resistance of stainless steels for fuel cell applications : supplementary report for phase 2

Energy Technology Data Exchange (ETDEWEB)

Kuyucak, S.; Li, J.; Liu, P.; Shehata, M.; Kruszewski, J.; Lo, J.; Guertsman, V.Y.; Gu, G.P. [Natural Resources Canada, Ottawa, ON (Canada). CANMET Materials Technology Laboratory

2007-07-15

This paper reported on a newly developed method of making bipolar electrodes from type 304 stainless steel. Two stainless steels were cast, hot-rolled and heat treated. The microstructures were then examined to determine the chromium carbide formation. Plain and mechanically polished samples were sent to General Motors for conductivity measurements to investigate the thermo-mechanical treatment as a means of improving the contact resistance of stainless steel bipolar plates subject to the operating conditions in a proton-exchange membrane (PEM) fuel cell. The treatment induces precipitation of conducive particles. The surface of the stainless steel is etched so that particles protrude from the surface. When the bipolar plates are stacked with sufficient load, the protruding surface precipitates indent into adjacent graphite electrodes, making direct electrical contact. The most common precipitate is M{sub 23}C{sub 6} carbide. This paper described the carbide precipitation required for electrical conductivity and presented a model for electrical conductance across a bipolar plate. It included a description of inter-particle distance and carbide size; carbide formation in type 304 stainless steels; heat-treatment processing of 304 steel for electrical conductance and desensitization; and the effect of steel composition on carbide growth. The experimental work was outlined in terms of casting, hot rolling, cold rolling, heat treatment, aging treatment for carbide growth, and desensitization treatment. Both alloys that were subjected to the thermo-mechanical treatment in this study showed a uniform distribution of carbide precipitates. Their size varied from very small to about 0.8{mu}m. Scanning electron microscopy (SEM) analysis did not detect a change in particle size and population density of these particles with prolonged annealing at 800 degrees C. 4 refs., 6 tabs., 14 figs.

Corrosion evaluation technology

International Nuclear Information System (INIS)

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

A Study On The Metal Carbide Composite Diffusion Bonding For Mechanical Seal

Directory of Open Access Journals (Sweden)

Kim D.-K.

2015-06-01

Full Text Available Mechanical Seal use highly efficient alternative water having a great quantity of an aqueous solution and has an advantage no corrosion brine. Metal Carbide composites have been investigated as potential materials for high temperature structural applications and for application in the processing industry. The existing Mechanical seal material is a highly expensive carbide alloy, and it is difficult to take a price advantage. Therefore the study of replacing body area with inexpensive steel material excluding O-ring and contact area which demands high characteristics is needed.

Influence and role of ethanol minor constituents of fuel grade ethanol on corrosion behavior of carbon steel

International Nuclear Information System (INIS)

Samusawa, Itaru; Shiotani, Kazuhiko

2015-01-01

Highlights: • The pitting factors of the minor contents of ethanol are acetic acid, Cl and H 2 O. • Formic acid in ethanol promotes general corrosion. • The H 2 O content in fuel-grade-ethanol (FGE) affects the corrosion morphology. • Acetic acid generates iron acetate, which has high solubility in FGE environments. • A pitting mechanism based on the rupture of passive film is proposed. - Abstract: The influences of organic acids, chloride and water on the corrosion behavior of carbon steel in fuel grade ethanol (FGE) environments were investigated by immersion testing in simulated FGE. The roles of acetic acid, chloride and water in pitting corrosion were studied by using X-ray photoelectron spectroscopy (XPS), auger electron spectroscopy (AES) and electrochemical experiments. The results indicated that iron acetate is generated on oxide film. Iron(II) acetate shows high solubility in FGE environments. The sites where iron(II) acetate is existed become preferential anodic sites, and chloride promotes anodic dissolution at such sites

Point defects and transport properties in carbides

International Nuclear Information System (INIS)

Matzke, Hj.

1984-01-01

Carbides of transition metals and of actinides are interesting and technologically important. The transition-metal carbides (or carbonitrides) are extensively being used as hard materials and some of them are of great interest because of the high transition temperature for superconductivity, e.g. 17 K for Nb(C,N). Actinide carbides and carbonitrides, (U,Pu)C and (U,Pu)(C,N) are being considered as promising advanced fuels for liquid metal cooled fast breeder nuclear reactors. Basic interest exists in all these materials because of their high melting points (e.g. 4250 K for TaC) and the unusually broad range of homogeneity of nonstoichiometric compositions (e.g. from UCsub(0.9) to UCsub(1.9) at 2500 K). Interaction of point defects to clusters and short-range ordering have recently been studied with elastic neutron diffraction and diffuse scattering techniques, and calculations of energies of formation and interaction of point defects became available for selected carbides. Diffusion measurements also exist for a number of carbides, in particular for the actinide carbides. The existing knowledge is discussed and summarized with emphasis on informative examples of particular technological relevance. (Auth.)

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)

Influence of alloying elements and density on aqueous corrosion behaviour of some sintered low alloy steels

International Nuclear Information System (INIS)

Kandavel, T.K.; Chandramouli, R.; Karthikeyan, P.

2012-01-01

Highlights: ► Corrosion of low alloy P/M steels under HCl acid pickling environment has been studied. ► Influence of density, strain and alloying elements on the rate of corrosion of the steels has been investigated. ► Residual porosity has significant effect on acid corrosion. ► Addition of the alloying elements Cu, Mo and Ti reduces the corrosion rate significantly. ► Carbide forming elements Mo and Ti improve further the resistance of the steels to aqueous corrosion. -- Abstract: Low alloy steels produced through powder metallurgy route of sintering followed by forging are promising candidate materials for high strength small components. Porosity in such steels poses a real challenge during acid pickling treatment, which is one of the processing steps during manufacturing. The present research work attempts to investigate the mechanism underlying the acid corrosion behaviour of some sintered low alloy steels under induced acid pickling conditions. Sintered-forged low alloy steel samples containing molybdenum (Mo), copper (Cu) and titanium (Ti) were subjected to aqueous corrosion attack by immersing the samples in 18% HCl (Hydrochloric acid) solution for 25 h. Sample weight loss and Fe (Iron) loss were estimated for the corroded samples. The morphology of the corroded surfaces was studied through metallography and scanning electron microscopy. Higher porosity alloys underwent enhanced corrosion rates. Both corrosion rate and iron loss are found to decrease linearly with reduction in porosity in all cases of the alloys. The alloying elements Mo, Ti and Cu, when added in combination, have played a complementary role in the reduction of corrosion rate by almost one order of magnitude compared to unalloyed steel. Presence of carbides of the carbide forming elements Mo and Ti played a positive role on the corrosion behaviour of the low alloy steels.

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?

Plan of development of ZrC-TRISO coated fuel particle and construction of ZrC coater

Energy Technology Data Exchange (ETDEWEB)

Ueta, Shohei; Ino, Hiroichi; Sawa, Kazuhiro [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Tobita, Tsutomu [Nuclear Engineering Company, Ltd., Tokai, Ibaraki (Japan); Takahashi, Masashi [Nuclear Fuel Industries, Ltd., Tokyo (Japan)

2002-11-01

In order to use coated fuel particle under higher temperature condition, more refractory coating material, which is more refractory than conventional silicon carbide (SiC), should be applied. Zirconium carbide (ZrC) is considered to be one of the promising materials, which is proposed as candidate for VHTR fuel material in GENERATION-IV, because of its intactness under high temperature of around 2000degC and its higher stability against kernel migration (amoeba effect) and fission product corrosion under normal operating condition. In order to develop ZrC coated particle for commercial use, research and development items were extracted based on review of the previous works. Research and development plan was determined. Based on the plan, a new ZrC coater of 100g batch size, which applies bromine process, was constructed. This report describes the review of precious works, extracted research and develop items and plan, and specifications of the ZrC coater. (author)

In-pile Hydrothermal Corrosion Evaluation of Coated SiC Ceramics and Composites

Energy Technology Data Exchange (ETDEWEB)

Carpenter, David [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Ang, Caen [Univ. of Tennessee, Knoxville, TN (United States); Katoh, Yutai [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Linton, Kory D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrani, Kurt A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-09-01

Hydrothermal corrosion accelerated by water radiolysis during normal operation is among the most critical technical feasibility issues remaining for silicon carbide (SiC) composite-based cladding that could provide enhanced accident-tolerance fuel technology for light water reactors. An integrated in-pile test was developed and performed to determine the synergistic effects of neutron irradiation, radiolysis, and pressurized water flow, all of which are relevant to a typical pressurized water reactor (PWR). The test specimens were chosen to cover a range of SiC materials and a variety of potential options for environmental barrier coatings. This document provides a summary of the irradiation vehicle design, operations of the experiment, and the specimen loading into the irradiation vehicle.

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)

Effects of cold work, sensitization treatment, and the combination on corrosion behavior of stainless steels in nitric acid

International Nuclear Information System (INIS)

Mayuzumi, M.; Ohta, J.; Arai, T.

1998-01-01

In a reprocessing process, spent nuclear fuels from light-water reactors are dissolved in nitric acid (HNO 3 ) to separate and recover the fissile materials such as uranium and plutonium from the radioactive fission products. Corrosion behavior of two stainless steels (SS) was investigated in nitric acid (HNO 3 ) for the effect of cold work (CW), sensitization heat treatment (Sens.), and a combination (CW + Sens.). The corrosion rate of the solution-treated type 304 SS (UNS S30400) with extra-low carbon (type 304ELC SS (UNS S30403)) increased with time and reached constant values after 1,000 h of immersion. However, constant corrosion rates were obtained for 25% Cr-20% Ni-Nb (type 310Nb SS [UNS S31040]) from the initial stage of immersion. CW mitigated corrosion of the solution-treated SS. The effect of CW was different on the two types of SS, with the sensitization heat-treated type 304 ELC SS showing higher corrosion rates and type 310Nb SS lower corrosion rates by CW. Corrosion resistance of type 310Nb SS was superior to type 304 ELC SS after all treatments. Chromium concentration of the sensitization-treated type 304 ELC SS was lower in the grain-boundary region than of the solution-treated one, although no chromium carbide precipitation was observed. This may have been the cause of intergranular corrosion enhancement by sensitization treatment

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.

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

Fabrication and characterization of fully ceramic microencapsulated fuels

Energy Technology Data Exchange (ETDEWEB)

Terrani, K.A., E-mail: kurt.terrani@gmail.com [Fuel Cycle and Isotopes Division, Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Kiggans, J.O.; Katoh, Y. [Materials Science and Technology Division, Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Shimoda, K. [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Montgomery, F.C.; Armstrong, B.L.; Parish, C.M. [Materials Science and Technology Division, Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Hinoki, T. [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Hunn, J.D. [Fuel Cycle and Isotopes Division, Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Snead, L.L. [Materials Science and Technology Division, Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States)

2012-07-15

The current generation of fully ceramic microencapsulated fuels, consisting of Tristructural Isotropic fuel particles embedded in a silicon carbide matrix, is fabricated by hot pressing. Matrix powder feedstock is comprised of alumina-yttria additives thoroughly mixed with silicon carbide nanopowder using polyethyleneimine as a dispersing agent. Fuel compacts are fabricated by hot pressing the powder-fuel particle mixture at a temperature of 1800-1900 Degree-Sign C using compaction pressures of 10-20 MPa. Detailed microstructural characterization of the final fuel compacts shows that oxide additives are limited in extent and are distributed uniformly at silicon carbide grain boundaries, at triple joints between silicon carbide grains, and at the fuel particle-matrix interface.

Influence of carbides and microstructure of CoCrMo alloys on their metallic dissolution resistance.

Science.gov (United States)

Valero-Vidal, C; Casabán-Julián, L; Herraiz-Cardona, I; Igual-Muñoz, A

2013-12-01

CoCrMo alloys are passive and biocompatible materials widely used as joint replacements due to their good mechanical properties and corrosion resistance. Electrochemical behaviour of thermal treated CoCrMo alloys with different carbon content in their bulk alloy composition has been analysed. Both the amount of carbides in the CoCrMo alloys and the chemical composition of the simulated body fluid affect the electrochemical properties of these biomedical alloys, thus passive dissolution rate was influenced by the mentioned parameters. Lower percentage of carbon in the chemical composition of the bulk alloy and thermal treatments favour the homogenization of the surface (less amount of carbides), thus increasing the availability of Cr to form the oxide film and improving the corrosion resistance of the alloy. © 2013.

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

Reactive-transport model for the prediction of the uniform corrosion behaviour of copper used fuel containers

International Nuclear Information System (INIS)

King, F.; Kolar, M.; Maak, P.

2008-01-01

Used fuel containers in a deep geological repository will be subject to various forms of corrosion. For containers made from oxygen-free, phosphorus-doped copper, the most likely corrosion processes are uniform corrosion, underdeposit corrosion, stress corrosion cracking, and microbiologically influenced corrosion. The environmental conditions within the repository are expected to evolve with time, changing from warm and oxidizing initially to cool and anoxic in the long-term. In response, the corrosion behaviour of the containers will also change with time as the repository environment evolve. A reactive-transport model has been developed to predict the time-dependent uniform corrosion behaviour of the container. The model is based on an experimentally-based reaction scheme that accounts for the various chemical, microbiological, electrochemical, precipitation/dissolution, adsorption/desorption, redox, and mass-transport processes at the container surface and in the compacted bentonite-based sealing materials within the repository. Coupling of the electrochemical interfacial reactions with processes in the bentonite buffer material allows the effect of the evolution of the repository environment on the corrosion behaviour of the container to be taken into account. The Copper Corrosion Model for Uniform Corrosion predicts the time-dependent corrosion rate and corrosion potential of the container, as well as the evolution of the near-field environment

High temperature oxidation of carbide-carbon materials of NbC-C, NbC-TiC-C systems

International Nuclear Information System (INIS)

Afonin, Yu.D.; Shalaginov, V.N.; Beketov, A.R.

1981-01-01

The effect of titanium carbide additions on the oxidation of carbide - carbon composition NbC-TiC-C in oxygen under the pressure of 10 mm Hg and in the air at atmospheric pressure in the temperature range 800-1300 deg is studied. It is shown that the region of negative temperature coefficient during oxidation in the system NbC+C is determined by the processes of sintering and polymorphous transformation. The specific character of the oxide film, formed during oxidation of Nbsub(x)Tisub(y)C+C composites is connected with non-equilibrium nature of carbide grain in its composition. Carbon gasification takes place with the formation of carbon dioxide. Composite materials, containing titanium carbide in complex carbide up to 50-83 mol. %, are the most corrosion resisting ones [ru

Overview of chemical characterization of FBTR fuel

International Nuclear Information System (INIS)

Venkatesan, V.; Nandi, C.; Patil, A.B.; Prakash, Amrit; Khan, K.B.; Arun Kumar

2015-01-01

Uranium Plutonium mixed carbide fuel is the driver fuel for Fast Breeder Test Reactor (FBTR) at IGCAR. The fuel is being fabricated at Radiometallurgy Division, BARC by conventional powder metallurgy route. During the fabrication of fuel, chemical quality control of process intermediates is very important to reach stringent specification of the final fuel product. Different steps are involved in the fabrication of uranium-plutonium carbide (MC) for FBTR. The main steps in the fabrication of MC fuel pellets are carbothermic reduction (CR) of mixture of uranium oxide, plutonium oxide and graphite powder to prepare MC clinkers, crushing and milling of MC clinkers and consolidation of MC powders into fuel pellets and sintering. As a part of process control, analysis of uranium (U), plutonium (Pu), carbon in oxide graphite mixture and U, Pu, carbon, oxygen, nitrogen, MC, M 2 C 3 contents in mixed carbide powder (MC clinkers) are carried out at our laboratory. Analysis of U, Pu, carbon, oxygen, nitrogen, MC and M 2 C 3 contents in mixed carbide sintered pellets are carried out as a part of quality control. This paper describes an overview of analytical instruments used during chemical quality control of mixed carbide fuel

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

International Nuclear Information System (INIS)

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

2007-01-01

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

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

Electrochemical Corrosion Behaviour of Alumina-Al 6061 and Silicon Carbide-Al 6061 Metal-Matrix Composites

International Nuclear Information System (INIS)

Mohamed, K.E.; Gad, M.M.A.; El-Sayed, A.A.; Moustafa, O.H.

2001-01-01

The electrochemical corrosion behaviour of powder metallurgy-processed metal-matrix composites (MMCs)based on Al alloy 6061 reinforced with particulate Al 2 O 3 or Sic has been studied in chloride-containing environment. Also, the corrosion behaviour of the unrein forced Al 6061 produced by the same route investigated. Electrochemical tests were conducted on composites containing 10 and 20 vo l% of both reinforced particulates. Potentiodynamic polarization tests have been carried out in neutral as well as acidic and alkaline de-aerated 10 -3 M Na CI solution. In the neutral environment, the addition of Al 2 O 3 particulates was found to shift both the corrosion potential (E corr ) and the break down potential (E b ) slightly into the positive direction irrespective of the volume fraction added (10 and 20 vo l%). On the other hand , Sic caused a shift of E corr into the active site while the E b value was slightly ennobled. For both composites, the corrosion current values at the break down potentials were almost the same as the unrein forced alloy. In an attempt to further clarify the role of both particulate addition, cathodic polarization runs were conducted in both acidic (ph 3) and alkaline (ph 9)solutions for 20 vo l% of Al 2 O 3 and 20 vo l% Sic composite specimens. This indicated that cathodic current values for Sic composites were higher than those corresponding to the unrein forced alloy 6061, and those for the Al 2 O 3 composites were lower

Corrosion resistance and electrochemical potentiokinetic reactivation testing of some iron-base hardfacing alloys

International Nuclear Information System (INIS)

Cockeram, B.V.

1999-01-01

Hardfacing alloys are weld deposited on a base material to provide a wear resistant surface. Commercially available iron-base hardfacing alloys are being evaluated for replacement of cobalt-base alloys to reduce nuclear plant activation levels. Corrosion testing was used to evaluate the corrosion resistance of several iron-base hardfacing alloys in highly oxygenated environments. The corrosion test results indicate that iron-base hardfacing alloys in the as-deposited condition have acceptable corrosion resistance when the chromium to carbon ratio is greater than 4. Tristelle 5183, with a high niobium (stabilizer) content, did not follow this trend due to precipitation of niobium-rich carbides instead of chromium-rich carbides. This result indicates that iron-base hardfacing alloys containing high stabilizer contents may possess good corrosion resistance with Cr:C < 4. NOREM 02, NOREM 01, and NoCo-M2 hardfacing alloys had acceptable corrosion resistance in the as-deposited and 885 C/4 hour heat treated condition, but rusting from sensitization was observed in the 621 C/6 hour heat treated condition. The feasibility of using an Electrochemical Potentiokinetic Reactivation (EPR) test method, such as used for stainless steel, to detect sensitization in iron-base hardfacing alloys was evaluated. A single loop-EPR method was found to provide a more consistent measurement of sensitization than a double loop-EPR method. The high carbon content that is needed for a wear resistant hardfacing alloy produces a high volume fraction of chromium-rich carbides that are attacked during EPR testing. This results in inherently lower sensitivity for detection of a sensitized iron-base hardfacing alloy than stainless steel using conventional EPR test methods

A fundamental study on stress corrosion cracking of SUS 304 steel in high temperature water

International Nuclear Information System (INIS)

Mukai, Yoshihiko; Murata, Masato

1985-01-01

SCC susceptibility of sensitized SUS 304 stainless steel in high temperature water was studied. The results obtained are as follows. SCC susceptibility was increased by adding crevices to the tensile specimen surface, for the corrodent became acidified by hydrolysis in crevices. SCC susceptibility was best fit to TTS curve obtained by EPR test, not by other corrosion tests such as Strauss test or the grain boundary corrosion test in high temperature water. In addition, by giving a simulated weld thermal cycle before the sensitizing heat treatment, the sensitization was clearly promoted. This seemed to be caused by the reason that nucleation of carbide occured in the simulated weld thermal cycle process and it promoted the carbide growth and the formation of Cr poor layer around carbide in the subsequent sensitization process. (author)

Long-term effects of neutron absorber and fuel matrix corrosion on criticality

International Nuclear Information System (INIS)

Culbreth, W.G.; Zielinski, P.R.

1994-01-01

Proposed waste package designs will require the addition of neutron absorbing material to prevent the possibility of a sustained chain reaction occurring in the fuel in the event of water intrusion. Due to the low corrosion rates of the fuel matrix and the Zircaloy cladding, there is a possibility that the neutron absorbing material will corrode and leak from the waste container long before the subsequent release of fuel matrix material. An analysis of the release of fuel matrix and neutron absorber material based on a probabilistic model was conducted and the results were used to prepare input to KENO-V, an neutron criticality code. The results demonstrate that, in the presence of water, the computed values of k eff exceeded the maximum of 0.95 for an extended period of time

Present status and further objectives of SNR fuel element development

International Nuclear Information System (INIS)

Karsten, G.

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

Benzotriazole (BTA), A Promising Corrosion Inhibitor for WC-Co Hardmetal

Energy Technology Data Exchange (ETDEWEB)

Schnyder, B.; Stoessel-Sittig, C.; Koetz, R.; Hochstrasser-Kurz, S. [ETH Zuerich (Switzerland); Virtanen, S. [ETH Zuerich (Switzerland); Jaeggi, Ch. [University of Bern (Switzerland); Eichenberger, N. [University of Bern (Switzerland); Szoecs, E. [University of Bern (Switzerland); Siegenthaler, H. [University of Bern (Switzerland); Ziegler, P. [AGIE SA (Switzerland); Beltrami, I. [AGIE SA (Switzerland)

2004-03-01

Wire Electro-Discharge Machining (W-EDM) of tungsten carbide with Co-binder may lead to corrosion and discolouration at the surface. The corrosion behaviour of WC-Co based hardmetal was investigated in different aqueous solutions (acidic, neutral, and alkaline solutions). At open-circuit potential WC-Co based hardmetals show rather high dissolution rates in all types of electrolyte. An efficient corrosion inhibitor (benzotriazole, C{sub 6}H{sub 5}N{sub 3}) could be found for a borate buffer solution, pH = 8.4. (author)

Dependence of corrosion properties of AISI 304L stainless steel on the austenite grain size

Energy Technology Data Exchange (ETDEWEB)

Sabooni, Soheil; Rashtchi, Hamed; Eslami, Abdoulmajid; Karimzadeh, Fathallah; Enayati, Mohammad Hossein; Raeissi, Keyvan; Imani, Reihane Faghih [Isfahan Univ. of Technology, Isfahan (Iran, Islamic Republic of). Dept. of Materials Engineering; Ngan, Alfonso Hing Wan [The Univ. of Hong Kong (China). Dept. of Mechanical Engineering

2017-07-15

The corrosion resistance of austenitic stainless steels is known to be hampered by the loss of chromium available for passive surface layer formation as a result of chromium carbide precipitation at austenite grain boundaries during annealing treatments. Although high-temperature annealing can promote carbide dissolution leading to better corrosion resistance, grain coarsening also results, which would lead to poorer mechanical properties. Processing methods to achieve both good corrosion resistance and mechanical properties are thus highly desirable for austenitic stainless steels. In the present study, we show that the corrosion resistance of AISI 304L stainless steel can be improved by grain refinement into the ultrafine-grained regime. Specifically, samples with different austenite grain sizes in the range of 0.65-12 μm were studied by potentiodynamic polarization and electrochemical impedance spectroscopy tests in a 3.5 wt.% NaCl solution. All samples showed a typical passive behavior with similar corrosion potential, but the corrosion current density decreased significantly with decreasing grain size. The results show that the sample with the finest grain size had the best corrosion resistance due to a higher resistance of the passive layer to pitting attacks. This study indicates that grain refinement which improves mechanical properties can also significantly improve the corrosion resistance of AISI 304L stainless steel.

Aluminum Corrosion and Turbidity

International Nuclear Information System (INIS)

Longtin, F.B.

2003-01-01

Aluminum corrosion and turbidity formation in reactors correlate with fuel sheath temperature. To further substantiate this correlation, discharged fuel elements from R-3, P-2 and K-2 cycles were examined for extent of corrosion and evidence of breaking off of the oxide film. This report discusses this study

Study through potentiodynamic techniques of the corrosion resistance of different aluminium base MMC's with boron additions

International Nuclear Information System (INIS)

Abenojar, J.; Bautista, A.; Guzman, S.; Velasco, F.; Martinez, M.A.

2009-01-01

This paper compares a wrought aluminium with a PM aluminium and PM aluminium alloys with boron-base additions, containing boron carbide and Fe/B (obtained by mechanical alloying during 36 hours from a Fe-B 50% mixture by weight). The effect of sintering temperature for the Fe/B containing material and the effect of mechanical alloying for the boron carbide containing aluminium alloy on the corrosion resistance of those materials have been studied. Their behaviour is followed through cyclic anodic polarization curves in chloride media. In the Al+20%Fe/B composite, low sintering temperatures (650- 950 deg C) exert a negative effect. However, when the material was sintered at high temperature (1000-1100 deg C) its behaviour was very similar to the PM pure aluminium. The effect of mechanical alloying studied in aluminium with boron carbide was also important in corrosion resistance, finding a lower corrosion rate in the mechanically alloyed material. (author)

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

International Nuclear Information System (INIS)

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

1977-10-01

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

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

International Nuclear Information System (INIS)

Haefner, H.E.

1982-03-01

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

An overview on dry reprocessing of irradiated nuclear fuels

International Nuclear Information System (INIS)

Ouyang Yinggen

2002-01-01

Although spent nuclear fuels have been reprocessed successfully for many years by the well-know Purex process based on solvent extraction, other reprocessing method which do not depend upon the use of organic solvents and aqueous media appear to have important potential advantage. There are two main non-aqueous methods for the reprocessing of spent fuel: fluoride-volatility process and pyro-electrochemical process. The presence of a poser in the process is that PuF 6 is obviously thermodynamically stable only in the presence of a large excess of fluorine. Pyro-electrochemical process is suited to processing metallic, oxide and carbide fuels. First, the fuel is dissolved in fresh salts, then, electrodes are introduced into the bath, U and Pu are deposited on the cathode, third, separation and refinement U and Pu are deposited on the cathode. There is a couple of contradictions in the process that are not in harmonious proportion in the fields on the nuclear fuel is dissolved the ability in the molten salt and corrosiveness of the molten salt for equipment used in the process

Combined Photoemission Spectroscopy and Electrochemical Study of a Mixture of (Oxy)carbides as Potential Innovative Supports and Electrocatalysts.

Science.gov (United States)

Calvillo, Laura; Valero-Vidal, Carlos; Agnoli, Stefano; Sezen, Hikmet; Rüdiger, Celine; Kunze-Liebhäuser, Julia; Granozzi, Gaetano

2016-08-03

Active and stable non-noble metal materials, able to substitute Pt as catalyst or to reduce the Pt amount, are vitally important for the extended commercialization of energy conversion technologies, such as fuel cells and electrolyzers. Here, we report a fundamental study of nonstoichiometric tungsten carbide (WxC) and its interaction with titanium oxycarbide (TiOxCy) under electrochemical working conditions. In particular, the electrochemical activity and stability of the WxC/TiOxCy system toward the ethanol electrooxidation reaction (EOR) and hydrogen evolution reaction (HER) are investigated. The chemical changes caused by the applied potential are established by combining photoemission spectroscopy and electrochemistry. WxC is not active toward the ethanol electrooxidation reaction at room temperature but it is highly stable under these conditions thanks to the formation of a passive thin film on the surface, consisting mainly of WO2 and W2O5, which prevents the full oxidation of WxC. In addition, WxC is able to adsorb ethanol, forming ethoxy groups on the surface, which constitutes the first step for the ethanol oxidation. The interaction between WxC and TiOxCy plays an important role in the electrochemical stability of WxC since specific orientations of the substrate are able to stabilize WxC and prevent its corrosion. The beneficial interaction with the substrate and the specific surface chemistry makes tungsten carbide a good electrocatalyst support or cocatalyst for direct ethanol fuel cells. However, WxC is active toward the HER and chemically stable under hydrogen reduction conditions, since no changes in the chemical composition or dissolution of the film are observed. This makes tungsten carbide a good candidate as electrocatalyst support or cocatalyst for the electrochemical production of hydrogen.

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-11-29

Fuel cladding chemical interactions (FCCI) have been acknowledged as a critical issue in a metallic fuel/steel cladding system due to the formation of low melting intermetallic eutectic compounds between the fuel and cladding steel, resulting in reduction in cladding wall thickness as well as a formation of eutectic compounds that can initiate melting in the fuel at lower temperature. In order to mitigate FCCI, diffusion barrier coatings on the cladding inner surface have been considered. In order to generate the required coating techniques, pack cementation, electroplating, and electrophoretic deposition have been investigated. However, these methods require a high processing temperature of above 700 oC, resulting in decarburization and decomposition of the martensites in a ferritic/martensitic (F/M) cladding steel. Alternatively, organometallic chemical vapor deposition (OMCVD) can be a promising process due to its low processing temperature of below 600 oC. The aim of the project is to conduct applied and fundamental research towards the development of diffusion barrier coatings on the inner surface of F/M fuel cladding tubes. Advanced cladding steels such as T91, HT9 and NF616 have been developed and extensively studied as advanced cladding materials due to their excellent irradiation and corrosion resistance. However, the FCCI accelerated by the elevated temperature and high neutron exposure anticipated in fast reactors, can have severe detrimental effects on the cladding steels through the diffusion of Fe into fuel and lanthanides towards into the claddings. To test the functionality of developed coating layer, the diffusion couple experiments were focused on using T91 as cladding and Ce as a surrogate lanthanum fission product. By using the customized OMCVD coating equipment, thin and compact layers with a few micron between 1.5 µm and 8 µm thick and average grain size of 200 nm and 5 µm were successfully obtained at the specimen coated between 300oC and

Influence of Ti, C and N concentration on the intergranular corrosion behaviour of AISI 316Ti and 321 stainless steels

International Nuclear Information System (INIS)

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

2007-01-01

Intergranular corrosion behaviour of 316Ti and 321 austenitic stainless steels has been evaluated in relation to the influence exerted by modification of Ti, C and N concentrations. For this evaluation, electrochemical measurements - double loop electrochemical potentiokinetic reactivation (DL-EPR) - were performed to produce time-temperature-sensitization (TTS) diagrams for tested materials. Transmission (TEM) and scanning electron microscopy (SEM) were used to determine the composition and nature of precipitates. The addition of Ti promotes better intergranular corrosion resistance in stainless steels. The precipitation of titanium carbides reduces the formation of chromium-rich carbides, which occurs at lower concentrations. Also, the reduction of carbon content to below 0.03 wt.% improves sensitization resistance more than does Ti content. The presence of Mo in AISI 316Ti stainless steel reduces chromium-rich carbide precipitation; the reason is that Mo increases the stability of titanium carbides and tends to replace chromium in the formation of carbides and intermetallic compounds, thus reducing the risks of chromium-depletion

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

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

Science.gov (United States)

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

2018-01-01

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

Novel fabrication of silicon carbide based ceramics for nuclear applications

Science.gov (United States)

Singh, Abhishek Kumar

Advances in nuclear reactor technology and the use of gas-cooled fast reactors require the development of new materials that can operate at the higher temperatures expected in these systems. These materials include refractory alloys based on Nb, Zr, Ta, Mo, W, and Re; ceramics and composites such as SiC--SiCf; carbon--carbon composites; and advanced coatings. Besides the ability to handle higher expected temperatures, effective heat transfer between reactor components is necessary for improved efficiency. Improving thermal conductivity of the fuel can lower the center-line temperature and, thereby, enhance power production capabilities and reduce the risk of premature fuel pellet failure. Crystalline silicon carbide has superior characteristics as a structural material from the viewpoint of its thermal and mechanical properties, thermal shock resistance, chemical stability, and low radioactivation. Therefore, there have been many efforts to develop SiC based composites in various forms for use in advanced energy systems. In recent years, with the development of high yield preceramic precursors, the polymer infiltration and pyrolysis (PIP) method has aroused interest for the fabrication of ceramic based materials, for various applications ranging from disc brakes to nuclear reactor fuels. The pyrolysis of preceramic polymers allow new types of ceramic materials to be processed at relatively low temperatures. The raw materials are element-organic polymers whose composition and architecture can be tailored and varied. The primary focus of this study is to use a pyrolysis based process to fabricate a host of novel silicon carbide-metal carbide or oxide composites, and to synthesize new materials based on mixed-metal silicocarbides that cannot be processed using conventional techniques. Allylhydridopolycarbosilane (AHPCS), which is an organometal polymer, was used as the precursor for silicon carbide. Inert gas pyrolysis of AHPCS produces near-stoichiometric amorphous

Iron Carbides in Fischer–Tropsch Synthesis: Theoretical and Experimental Understanding in Epsilon-Iron Carbide Phase Assignment

International Nuclear Information System (INIS)

Liu, Xing-Wu; Cao, Zhi; Zhao, Shu; Gao, Rui

2017-01-01

As active phases in low-temperature Fischer–Tropsch synthesis for liquid fuel production, epsilon iron carbides are critically important industrial materials. However, the precise atomic structure of epsilon iron carbides remains unclear, leading to a half-century of debate on the phase assignment of the ε-Fe 2 C and ε’-Fe 2.2 C. Here, we resolve this decades-long question by a combining theoretical and experimental investigation to assign the phases unambiguously. First, we have investigated the equilibrium structures and thermal stabilities of ε-Fe x C, (x = 1, 2, 2.2, 3, 4, 6, 8) by first-principles calculations. We have also acquired X-ray diffraction patterns and Mössbauer spectra for these epsilon iron carbides, and compared them with the simulated results. These analyses indicate that the unit cell of ε-Fe 2 C contains only one type of chemical environment for Fe atoms, while ε’-Fe 2.2 C has six sets of chemically distinct Fe atoms.

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)

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)

A new corrosion resistant, martensitic stainless steel for improved performance in miniature bearings

Energy Technology Data Exchange (ETDEWEB)

Tomasello, C.M.; Maloney, J.L.; Materkowski, J.P. [Latrobe Steel Co., Latrobe, PA (United States); Ward, P.C. [MPB Corp., Keene, NH (United States)

1998-12-31

A new alloy, 440 N-DUR{trademark} has been developed which will provide the corrosion resistance of 440C with improved carbide size and distribution for noiseless miniature precision bearing operation. The alloy may be through hardened to achieve a minimum hardness of 60 HRC. Its nominal composition is 0.65 wt.% C, 14.5 wt.% Cr, 0.30 wt.% Si, 0.45 wt.% Mn and 0.10 wt.% N{sub 2}. The development of the alloy is a result of a factorial experimental design including 17 alloy variants. The optimum alloy provides a combination of the best carbide structure, corrosion resistance and heat treat response. The addition of nitrogen combined with this carbon and chromium content improves the alloy`s hardenability and corrosion resistance. The alloy successfully withstands copper sulfate exposure and is currently being tested in several bearing applications. It also has great potential to outperform 440C and other corrosion resistant alloys for other ambient and low temperature applications because of its improved microstructure and heat treat response.

Effect of cerium addition on the corrosion behaviour of carbon-alloyed iron aluminides

International Nuclear Information System (INIS)

Sriram, S.; Balasubramaniam, R.; Mungole, M.N.; Bharagava, S.; Baligidad, R.G.

2006-01-01

The effect of Ce addition on the microstructure and corrosion behavior of carbon-alloyed iron aluminides Fe-20.0Al-2.0C, Fe-18.5Al-3.6C and Fe-19.2Al-3.3C-0.07Ce (in at.%) has been studied. The potentiodynamic polarization behaviour of the alloys was evaluated in freely aerated 0.25 mol/l H 2 SO 4 . A 0.05% C steel was used for comparison purposes. All the alloys exhibited active-passive behaviour in the acidic solution. The addition of Ce destroyed passivity as indicated by lower breakdown potentials in polarization studies. This has been related to the finer distribution of the carbides in the microstructure. Corrosion rates were evaluated by immersion testing. The iron aluminide with Ce addition exhibited a lower corrosion rate compared to the aluminides without Ce addition. This has been attributed to modifications in surface film with Ce addition. Scanning electron microscopy of corroded surfaces indicated that the carbon-alloyed intermetallics were susceptible to localized galvanic corrosion due to the presence of carbides in the microstructure

Effect of the reinforced boron carbide particulate content of AA6061 alloy on formation of the passive film in seawater

International Nuclear Information System (INIS)

Katkar, V.A.; Gunasekaran, G.; Rao, A.G.; Koli, P.M.

2011-01-01

Highlights: → Presence of boron carbide increases the corrosion rate of A6061 alloy in seawater. → Increasing the B 4 C content decreases passive layer thickness. → Passive films formed on A6061 and its B 4 C composites are n-type semiconductors. - Abstract: The effect of boron carbide (B 4 C) reinforcement on the corrosion of AA6061 alloy was studied by investigating passive films formed in seawater. The higher passive current and its potential-dependence for these composites indicated formation of porous passive film. Electrochemical impedance spectroscopy (EIS) graph suggests that the alloy surface is partly or totally active. The formed passive film is n-type semiconductor junction in nature. The difference between corrosion potential (E corr ) and potential at zero charge (PZC) suggests that the chloride ions responsible for film breakdown exist within the passive film. A suitable mechanism is proposed for the passive film breakdown.

Hardened over-coating fuel particle and manufacture of nuclear fuel using its fuel particle

International Nuclear Information System (INIS)

Yoshimuda, Hideharu.

1990-01-01

Coated-fuel particles comprise a coating layer formed by coating ceramics such as silicon carbide or zirconium carbide and carbons, etc. to a fuel core made of nuclear fuel materials. The fuel core generally includes oxide particles such as uranium, thorium and plutonium, having 400 to 600 μm of average grain size. The average grain size of the coated-fuel particle is usually from 800 to 900 μm. The thickness of the coating layer is usually from 150 to 250 μm. Matrix material comprising a powdery graphite and a thermosetting resin such as phenol resin, etc. is overcoated to the surface of the coated-fuel particle and hardened under heating to form a hardened overcoating layer to the coated-fuel particle. If such coated-fuel particles are used, cracks, etc. are less caused to the coating layer of the coated-fuel particles upon production, thereby enabling to prevent the damages to the coating layer. (T.M.)

Effect of Low-Temperature Sensitization on the Corrosion Behavior of AISI Type 304L SS Weld Metal in Simulated Groundwater

Science.gov (United States)

Suresh, Girija; Nandakumar, T.; Viswanath, A.

2018-05-01

The manuscript presents the investigations carried out on the effect of low-temperature sensitization (LTS) of 304L SS weld metal on its corrosion behavior in simulated groundwater, for its application as a canister material for long-term storage of nuclear vitrified high-level waste in geological repositories. AISI type 304L SS weld pad was fabricated by multipass gas tungsten arc welding process using 308L SS filler wire. The as-welded specimens were subsequently subjected to carbide nucleation and further to LTS at 500 °C for 11 days to simulate a temperature of 300 °C for 100-year life of the canister in geological repositories. Delta ferrite ( δ-ferrite) content of the 304L SS weld metal substantially decreased on carbide nucleation treatment and further only a marginal decrease occurred on LTS treatment. The microstructure of the as-welded consisted of δ-ferrite as a minor phase distributed in austenite matrix. The δ-ferrite appeared fragmented in the carbide-nucleated and LTS-treated weld metal. The degree of sensitization measured by double-loop electrochemical potentokinetic reactivation method indicated an increase in carbide nucleation treatment when compared to the as-welded specimens, and further increase occurred on LTS treatment. Potentiodynamic anodic polarization investigations in simulated groundwater indicated a substantial decrease in the localized corrosion resistance of the carbide-nucleated and LTS 304L SS weld metals, when compared to the as-welded specimens. Post-experimental micrographs indicated pitting as the primary mode of attack in the as-welded, while pitting and intergranular corrosion (IGC) occurred in the carbide-nucleated weld metal. LTS-treated weld metal predominantly underwent IGC attack. The decrease in the localized corrosion resistance of the weld metal after LTS treatment was found to have a direct correlation with the degree of sensitization and the weld microstructure. The results are detailed in the manuscript.

Effect of Low-Temperature Sensitization on the Corrosion Behavior of AISI Type 304L SS Weld Metal in Simulated Groundwater

Science.gov (United States)

Suresh, Girija; Nandakumar, T.; Viswanath, A.

2018-04-01

The manuscript presents the investigations carried out on the effect of low-temperature sensitization (LTS) of 304L SS weld metal on its corrosion behavior in simulated groundwater, for its application as a canister material for long-term storage of nuclear vitrified high-level waste in geological repositories. AISI type 304L SS weld pad was fabricated by multipass gas tungsten arc welding process using 308L SS filler wire. The as-welded specimens were subsequently subjected to carbide nucleation and further to LTS at 500 °C for 11 days to simulate a temperature of 300 °C for 100-year life of the canister in geological repositories. Delta ferrite (δ-ferrite) content of the 304L SS weld metal substantially decreased on carbide nucleation treatment and further only a marginal decrease occurred on LTS treatment. The microstructure of the as-welded consisted of δ-ferrite as a minor phase distributed in austenite matrix. The δ-ferrite appeared fragmented in the carbide-nucleated and LTS-treated weld metal. The degree of sensitization measured by double-loop electrochemical potentokinetic reactivation method indicated an increase in carbide nucleation treatment when compared to the as-welded specimens, and further increase occurred on LTS treatment. Potentiodynamic anodic polarization investigations in simulated groundwater indicated a substantial decrease in the localized corrosion resistance of the carbide-nucleated and LTS 304L SS weld metals, when compared to the as-welded specimens. Post-experimental micrographs indicated pitting as the primary mode of attack in the as-welded, while pitting and intergranular corrosion (IGC) occurred in the carbide-nucleated weld metal. LTS-treated weld metal predominantly underwent IGC attack. The decrease in the localized corrosion resistance of the weld metal after LTS treatment was found to have a direct correlation with the degree of sensitization and the weld microstructure. The results are detailed in the manuscript.

Laser cladding in-situ carbide particle reinforced Fe-based composite coatings with rare earth oxide addition

Institute of Scientific and Technical Information of China (English)

吴朝锋; 马明星; 刘文今; 钟敏霖; 张红军; 张伟明

2009-01-01

Particulate reinforced metal matrix composite(PR-MMC) has excellent properties such as good wear resistance,corrosion resistance and high temperature properties.Laser cladding is usually used to form PR-MMC on metal surface with various volume fractions of ceramic particles.Recent literatures showed that laser melting of powder mixture containing carbon and carbide-forming elements,was favorable for the formation of in-situ synthesized carbide particles.In this paper,rare earth oxide(RE2O3) was added into t...

On the enhancement of corrosion in HTR graphite due to machining

International Nuclear Information System (INIS)

Hatton, D.; Harris, A.M.; Hall, J.A.

1975-09-01

A correlation is reported between bands of corrosion on the coolant surface of an inner sleeve from a Dragon reactor tubular interacting pin, and peaks in the axial profile of 60 Co along the pin. Tungsten has been observed at the corroded positions and it is postulated that these materials, deposited from a tungsten carbide tipped tool during machining, cause catalytic corrosion of the graphite. (author)

Reactor fuel cladding tube with excellent corrosion resistance and method of manufacturing the same

International Nuclear Information System (INIS)

Okuda, Takanari; Kanehara, Mitsuo; Abe, Katsuhiro; Nishimura, Takashi.

1995-01-01

The present invention provides a fuel cladding tube having an excellent corrosion resistance and thus a long life, and a suitable manufacturing method therefor. Namely, in the fuel cladding tube, the outer circumference of an inner layer made of a zirconium base alloy is coated with an outer layer made of a metal more corrosion resistant than the zirconium base alloy. Ti or a titanium alloy is suitable for the corrosion resistant metal. In addition, the outer layer can be coated by a method such as vapor deposition or plating, not limited to joining of the inner layer material and the outer layer material. Specifically, a composite material having an inner layer made of a zirconium alloy coated by the outer material made of a titanium alloy is applied with hot fabrication at a temperature within a range of from 500 to 850degC and at a fabrication rate of not less than 5%. The fabrication method includes any of extrusion, rolling, drawing, and casting. As the titanium-base alloy, a Ti-Al alloy or a Ti-Nb alloy containing Al of not more than 20wt%, or Nb of not more than 20wt% is preferred. (I.S.)

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

plants. The type of corrosion attack can be directly ascribed to the composition of the deposit and the metal surface temperature. A series of field tests have been undertaken in the various straw-fired power plants in Denmark, namely Masnedø, Rudkøbing and Ensted. Three types of exposure were undertaken......In Denmark, straw and other types of biomass are used for generating energy in power plants. Straw has the advantage that it is a "carbon dioxide neutral fuel" and therefore environmentally acceptable. Straw combustion is associated with corrosion problems which are not encountered in coal-fired...... 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...

Corrosion product balances for the Ringhals PWR plants based on extensive fuel crud and water chemistry measurements

International Nuclear Information System (INIS)

Lundgren, K.; Wikmark, G.; Bengtsson, B.

2010-01-01

The corrosion product balance in a PWR plant is of great importance for the fuel performance as well as for the radiation field buildup. This balance is of special concern in connection to steam generator replacement (SGR) and power uprate projects. The Ringhals PWRs are all of Westinghouse design. Two of the plants have performed Steam Generator Replacement (SGR) to I-690 SG tubes and such a replacement is being planned in the third and last unit in 2011. Two of the units are in different phases of power uprate projects. The plants are all on 10-14-months cycles operating with medium to high fuel duty. Water chemistry is controlled by a pH300 in the range ∼7.2 to 7.4 from beginning of cycle to end of cycle (BOC-EOC) in the units with new SGs while kept at a coordinated pH of 7.2 in the one still using I-600. The maximum Li content has recently been increased to about 4.5 to 5 ppm in all units. In order to be able to improve the assessment of corrosion product balances in the plants, comprehensive fuel crud measurements were performed in 2007. Improved integrated reactor water sampling techniques have also been introduced in order to make accurate mass balances possible. The corrosion products covered in the study are the main constituents, Ni, Fe and Cr in the primary circuit Inconel and stainless steel, together with Co. The activated corrosion products, Co-58, Co-60, Cr-51, Fe-59 and Mn-54, are all mainly produced through neutron irradiation of the covered corrosion products. The main results of the corrosion product balances are presented. Observed differences between the plants, indicating significant impact of pH control and SG tube materials, are presented and discussed. The importance of accurate sampling techniques is especially addressed in this paper. (author)

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

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

International Nuclear Information System (INIS)

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

1976-02-01

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

Advanced technologies of production of cemented carbides and composite materials based on them

International Nuclear Information System (INIS)

Bondarenko, V.; Pavlotskaya, E.; Martynova, L.; Epik, I.

2001-01-01

The paper presents new technological processes of production of W, WC and (Ti, W)C powders, cemented carbides having a controlled carbon content, high-strength nonmagnetic nickel-bonded cemented carbides, cemented carbide-based composites having a wear-resistant antifriction working layer as well as processes of regeneration of cemented carbide waste. It is shown that these technological processes permit radical changes in the production of carbide powders and products of VK, TK, VN and KKhN cemented carbides. The processes of cemented carbide production become ecologically acceptable and free of carbon black, the use of cumbersome mixers is excluded, the power expenditure is reduced and the efficiency of labor increases. It becomes possible to control precisely the carbon content within a two-phase region -carbide-metal. A high wear resistance of parts of friction couples which are lubricated with water, benzine, kerosene, diesel fuel and other low-viscosity liquids, is ensured with increased strength and shock resistance. (author)

High performance nuclear fuel element

International Nuclear Information System (INIS)

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

1980-01-01

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

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

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

Science.gov (United States)

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

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2008-04-15

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

Instant release fraction corrosion studies of commercial UO2 BWR spent nuclear fuel

Science.gov (United States)

Martínez-Torrents, Albert; Serrano-Purroy, Daniel; Sureda, Rosa; Casas, Ignasi; de Pablo, Joan

2017-05-01

The instant release fraction of a spent nuclear fuel is a matter of concern in the performance assessment of a deep geological repository since it increases the radiological risk. Corrosion studies of two different spent nuclear fuels were performed using bicarbonate water under oxidizing conditions to study their instant release fraction. From each fuel, cladded segments and powder samples obtained at different radial positions were used. The results were normalised using the specific surface area to permit a comparison between fuels and samples. Different radionuclide dissolution patterns were studied in terms of water contact availability and radial distribution in the spent nuclear fuel. The relationship between the results of this work and morphological parameters like the grain size or irradiation parameters such as the burn-up or the linear power density was studied in order to increase the understanding of the instant release fraction formation.

Pilot production of 325 kg of uranium carbide

International Nuclear Information System (INIS)

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

1969-01-01

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

Capabilities to improve corrosion resistance of fuel claddings by using powerful laser and plasma sources

Energy Technology Data Exchange (ETDEWEB)

Borisov, V. M., E-mail: borisov@triniti.ru; Trofimov, V. N.; Sapozhkov, A. Yu.; Kuzmenko, V. A.; Mikhaylov, V. B.; Cherkovets, V. Ye.; Yakushkin, A. A. [Troitsk Institute for Innovation and Fusion Research (Russian Federation); Yakushin, V. L.; Dzhumayev, P. S. [National Research Nuclear University MEPhI (Moscow Engineering Physics Institute) (Russian Federation)

2016-12-15

The treatment conditions of fuel claddings of the E110 alloy by using powerful UV or IR laser radiation, which lead to the increase in the corrosion resistance at the high-temperature (T = 1100°C) oxidation simulating a loss-of-coolant accident, are determined. The possibility of the complete suppression of corrosion under these conditions by using pulsed laser deposition of a Cr layer is demonstrated. The behavior of protective coatings of Al, Al{sub 2}O{sub 3}, and Cr planted on steel EP823 by pulsed laser deposition, which is planned to be used in the BREST-OD-300, is studied. The methods of the almost complete suppression of corrosion in liquid lead to the temperature of 720°C are shown.

Corrosion characteristics of K-claddings

International Nuclear Information System (INIS)

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

2004-01-01

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

Fuel rod D07/B15 from Ringhals 2 PWR: Source material for corrosion/leach tests in groundwater. Fuel rod/pellet characterization program. Pt. 1

International Nuclear Information System (INIS)

Forsyth, R.

1987-03-01

A joint SKB/STUDSVIK experimental program to determine the corrosion rates and to establish the corrosion mechanisms of spent UO 2 fuel in groundwater under both oxidizing and reducing conditions is in progress in the Hot Cell Laboratory of Studsvik Energiteknik AB. High burnup fuel of both BWR and PWR type are studied. Characterization of the spent fuel at both rod and pellet level is an important part of the experimental program. Experiments on PWR fuel have been concentrated so far on specimens from one rod, manufacturer's number 03688, which had occupied position B15 in assembly D07. This assembly had been irradiated for 5 cycles in the Ringhals 2 reactor between 1977 and 1983. The calculated assembly burnup was 41.3 MWd/kg U. The present report is a collection of separate reports describing those items in the characterization program which have been performed so far. No overall summary of the experimental results is given here, and the report should be viewed as a collection of reference data. (orig.)

Evaluation of seawater corrosion of SSCs

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-08-15

In the unit 1 to unit 4 of the Fukushima Daiichi Nuclear Power Plant, seawater was injected in reactor pressure vessels and spent fuel pools in order to cool nuclear fuel after the disaster of the 2011 off the Pacific coast of Tohoku Earthquake and Tsunami. In fiscal 2012, overall plan of this project has been developed in consideration of corrosion events that might be assumed reactor pressure vessels, spent fuel pools and primary containment vessels of Fukushima Daiichi Nuclear Power Station that was designated to be as the 'Specified Nuclear Power Facilities'. In this project, crevice corrosion susceptibility of stainless steel, galvanic corrosion of aluminum alloy, and uniform corrosion of carbon steel piping will be evaluated. (author)

High-temperature effect of hydrogen on sintered alpha-silicon carbide

Science.gov (United States)

Hallum, G. W.; Herbell, T. P.

1986-01-01

Sintered alpha-silicon carbide was exposed to pure, dry hydrogen at high temperatures for times up to 500 hr. Weight loss and corrosion were seen after 50 hr at temperatures as low as 1000 C. Corrosion of SiC by hydrogen produced grain boundary deterioration at 1100 C and a mixture of grain and grain boundary deterioration at 1300 C. Statistically significant strength reductions were seen in samples exposed to hydrogen for times greater than 50 hr and temperatures above 1100 C. Critical fracture origins were identified by fractography as either general grain boundary corrision at 1100 C or as corrosion pits at 1300 C. A maximum strength decrease of approximately 33 percent was seen at 1100 and 1300 C after 500 hr exposure to hydrogen. A computer assisted thermodynamic program was also used to predict possible reaction species of SiC and hydrogen.

Carbon potential measurement on some actinide carbides

International Nuclear Information System (INIS)

Anthonysamy, S.; Ananthasivan, K.; Kaliappan, I.; Chandramouli, V.; Vasudeva Rao, P.R.; Mathews, C.K.; Jacob, K.T.

1994-01-01

Uranium-Plutonium mixed carbides with a Pu/(U+Pu) ratio of 0.55 are to be used as the fuel in the Fast Breeder Test Reactor (FBTR) at Kalpakkam, India. Carburization of the stainless steel clad by this fuel is determined by its carbon potential. Because the carbon potential of this fuel composition is not available in the literature, it was measured by the methane-hydrogen gas equilibration technique. The sample was equilibrated with purified hydrogen and the equilibrium methane-to-hydrogen ratio in the gas phase was measured with a flame ionization detector. The carbon potential of the ThC-ThC 2 as well as Mo-Mo 2 C system, which is an important binary in the actinide-fission product-carbon systems, were also measured by this technique in the temperature range 973 to 1,173 K. The data for the Mo-Mo 2 C system are in agreement with values reported in the literature. The results for the ThC-ThC 2 system are different from estimated values with large uncertainty limits given in the literature. The data on (U, Pu) mixed carbides indicates the possibility of stainless steel clad attack under isothermal equilibrium conditions

Low temperature CVD deposition of silicon carbide

International Nuclear Information System (INIS)

Dariel, M.; Yeheskel, J.; Agam, S.; Edelstein, D.; Lebovits, O.; Ron, Y.

1991-04-01

The coating of graphite on silicon carbide from the gaseous phase in a hot-well, open flow reactor at 1150degC is described. This study constitutes the first part of an investigation of the process for the coating of nuclear fuel by chemical vapor deposition (CVD)

The German carbide program: Performance, experimental findings, and evaluation of irradiation results

International Nuclear Information System (INIS)

Steiner, H.; Freund, D.; Geithoff, D.

1982-09-01

In this report a synopsis of the German carbide program is presented. The program comprises the irradiation of about 100 carbide pins equipped with pelletted fuel. Most of these fuel pins were He-bonded, the sodium bonding concept taken as a back-up solution. The main design parameters such as smear and pellet density, gap size, pin diameter and wall thickness as well as the irradiation conditions were varied mostly within wide ranges. Based on a compilation of relevant pin parameters, irradiation conditions, and the results of various irradiation experiments conclusions on the optimum ranges of the main design parameters are drawn. Furthermore, some important aspects of fuel pin behaviour are discussed based on quantitative results from post irradiation examinations. (orig.) [de

Some proposed mechanisms for internal cladding corrosion

International Nuclear Information System (INIS)

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

1977-01-01

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

Some proposed mechanisms for internal cladding corrosion

Energy Technology Data Exchange (ETDEWEB)

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

1977-04-01

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

High temperature corrosion of metallic interconnects in solid oxide fuel cells

International Nuclear Information System (INIS)

Bastidas, D. M.

2006-01-01

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. (Author) 66 refs

Alloy Microstructure Dictates Corrosion Modes in THA Modular Junctions.

Science.gov (United States)

Pourzal, Robin; Hall, Deborah J; Ehrich, Jonas; McCarthy, Stephanie M; Mathew, Mathew T; Jacobs, Joshua J; Urban, Robert M

2017-12-01

Adverse local tissue reactions (ALTRs) triggered by corrosion products from modular taper junctions are a known cause of premature THA failure. CoCrMo devices are of particular concern because cobalt ions and chromium-orthophosphates were shown to be linked to ALTRs, even in metal-on-polyethylene THAs. The most common categories of CoCrMo alloy are cast and wrought alloy, which exhibit fundamental microstructural differences in terms of grain size and hard phases. The impact of implant alloy microstructure on the occurring modes of corrosion and subsequent metal ion release is not well understood. The purpose of this study was to determine whether (1) the microstructure of cast CoCrMo alloy varies broadly between manufacturers and can dictate specific corrosion modes; and whether (2) the microstructure of wrought CoCrMo alloy is more consistent between manufacturers and has low implications on the alloy's corrosion behavior. The alloy microstructure of four femoral-stem and three femoral-head designs from four manufacturers was metallographically and electrochemically characterized. Three stem designs were made from cast alloy; all three head designs and one stem design were made from wrought alloy. Alloy samples were sectioned from retrieved components and then polished and etched to visualize grain structure and hard phases such as carbides (eg, M 23 C 6 ) or intermetallic phases (eg, σ phase). Potentiodynamic polarization (PDP) tests were conducted to determine the corrosion potential (E corr ), corrosion current density (I corr ), and pitting potential (E pit ) for each alloy. Four devices were tested within each group, and each measurement was repeated three times to ensure repeatable results. Differences in PDP metrics between manufacturers and between alloys with different hard phase contents were compared using one-way analysis of variance and independent-sample t-tests. Microstructural features such as twin boundaries and slip bands as well as corrosion

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.

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

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

Science.gov (United States)

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

2011-03-01

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

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)

Carbide Coatings for Nickel Alloys, Graphite and Carbon/Carbon Composites to be used in Fluoride Salt Valves

Energy Technology Data Exchange (ETDEWEB)

Nagle, Denis [Johns Hopkins Univ., Baltimore, MD (United States); Zhang, Dajie [Johns Hopkins Univ., Baltimore, MD (United States)

2015-10-22

The focus of this research was concerned with developing materials technology that supports the evolution of Generation IV Advanced High Temperature Reactor (AHTR) concepts. Specifically, we investigate refractory carbide coatings for 1) nickel alloys, and 2) commercial carbon-carbon composites (CCCs). Numerous compelling reasons have driven us to focus on carbon and carbide materials. First, unlike metals, the strength and modulus of CCCs increase with rising temperature. Secondly, graphite and carbon composites have been proven effective for resisting highly corrosive fluoride melts such as molten cryolite [Na₃AlF₆] at ~1000°C in aluminum reduction cells. Thirdly, graphite and carbide materials exhibit extraordinary radiation damage tolerance and stability up to 2000°C. Finally, carbides are thermodynamically more stable in liquid fluoride salt than the corresponding metals (i.e. Cr and Zr) found in nickel based alloys.

Corrosion properties of HLW and spent fuel overpacks in highly alkaline environments

International Nuclear Information System (INIS)

Kursten, B.

2009-01-01

Throughout the world, deep geological disposal in stable rocks with low groundwater flow is considered for the long-term management of long-lived radioactive waste (vitrified high-level waste - VHLW - and spent fuel - SF).The main advantage of the SC design, with respect to corrosion, is that under the predicted conditions (i.e. highly alkaline concrete buffer), the carbon steel overpack is expected to undergo uniform corrosion (passive dissolution). The key objective of this study is to demonstrate that the carbon steel overpack will be able to ensure complete containment of the radioactivity at least during the thermal phase, this is the period during which the temperature of the host rock is expected to lie above the range of temperatures within which nominal radionuclide migration properties can be relied upon

Corrosion of immersed ceramic heat exchanger tubes in aluminium foundry baths

Energy Technology Data Exchange (ETDEWEB)

Bracho-Troconis, C.B.; Frot, G.; Bienvenu, Y. [Ecole des Mines de Paris, Evry (France). Centre des Materiaux; Frety, N. [Ecole des Mines d`Albi-Carmaux (France); Alliat, I. [CERSTA-Gaz de France, Saint-Denis (France)

1997-12-31

The corrosion of three non-oxide ceramics by Al-9Si-3Cu baths and by fluxes (mixtures of chlorides and fluorides of sodium and potassium) at about 750 C was studied in a foundry environment. Comparison of results of the metallurgical examination of A, a silicon-nitride-bonded silicon carbide and of B, a reaction-bonded silicon nitride, surface treated to fill all the external porosity provides some insight into the role of the bonding phase and the porosity. Grade C is a graphite bonded silicon carbide with an external protection by a ceramic glazing. The SiC phase in the tubes is inert to the corrosive liquids (attributed to the silicon content in the metal). A and C ceramics react only in the presence of a flux. Sodium and chlorine were identified in the corrosion products as well as AlN (A) and Al{sub 4}C{sub 3} (C), resulting from reaction of the silicon nitride or of the graphite bonding phase with aluminium. This suggests that the fluxes are responsible for the corrosive process, by causing the formation of gaseous aluminium halides which penetrate the porous bonding phase and react with it to form AlN or Al{sub 4}C{sub 3}. (orig.) 13 refs.

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)

The oxidative corrosion of carbide inclusions at the surface of uranium metal during exposure to water vapour

International Nuclear Information System (INIS)

Scott, T.B.; Petherbridge, J.R.; Harker, N.J.; Ball, R.J.; Heard, P.J.; Glascott, J.; Allen, G.C.

2011-01-01

Highlights: → High resolution imagery (FIB, SEM and SIMS) of carbide inclusions in uranium metal. → Real time images following the reaction of the carbide inclusions with water vapour. → Shown preferential consumption of carbide over that of the bulk metal. → Quantity of impurities in the metal therefore seriously influence reaction rate. → Metal purity must be considered when storing uranium in air or moist conditions. - Abstract: The reaction between uranium and water vapour has been well investigated, however discrepancies exist between the described kinetic laws, pressure dependence of the reaction rate constant and activation energies. Here this problem is looked at by examining the influence of impurities in the form of carbide inclusions on the reaction. Samples of uranium containing 600 ppm carbon were analysed during and after exposure to water vapour at 19 mbar pressure, in an environmental scanning electron microscope (ESEM) system. After water exposure, samples were analysed using secondary ion mass spectrometry (SIMS), focused ion beam (FIB) imaging and sectioning and transmission electron microscopy (TEM) with X-ray diffraction (micro-XRD). The results of the current study indicate that carbide particles on the surface of uranium readily react with water vapour to form voluminous UO 3 .xH 2 O growths at rates significantly faster than that of the metal. The observation may also have implications for previous experimental studies of uranium-water interactions, where the presence of differing levels of undetected carbide may partly account for the discrepancies observed between datasets.

Tungsten carbide encapsulated in nitrogen-doped carbon with iron/cobalt carbides electrocatalyst for oxygen reduction reaction

Energy Technology Data Exchange (ETDEWEB)

Zhang, Jie; Chen, Jinwei, E-mail: jwchen@scu.edu.cn; Jiang, Yiwu; Zhou, Feilong; Wang, Gang; Wang, Ruilin, E-mail: rl.wang@scu.edu.cn

2016-12-15

Graphical abstract: A hybrid catalyst was prepared via a quite green and simple method to achieve an one-pot synthesis of the N-doping carbon, tungsten carbides, and iron/cobalt carbides. It exhibited comparable electrocatalytic activity, higher durability and ability to methanol tolerance compared with commercial Pt/C to ORR. - Highlights: • A novel type of hybrid Fe/Co/WC@NC catalysts have been successfully synthesized. • The hybrid catalyst also exhibited better durability and methanol tolerance. • Multiple effective active sites of Fe{sub 3}C, Co{sub 3}C, WC, and NC help to improve catalytic performance. - Abstract: This work presents a type of hybrid catalyst prepared through an environmental and simple method, combining a pyrolysis of transition metal precursors, a nitrogen-containing material, and a tungsten source to achieve a one-pot synthesis of N-doping carbon, tungsten carbides, and iron/cobalt carbides (Fe/Co/WC@NC). The obtained Fe/Co/WC@NC consists of uniform Fe{sub 3}C and Co{sub 3}C nanoparticles encapsulated in graphitized carbon with surface nitrogen doping, closely wrapped around a plate-like tungsten carbide (WC) that functions as an efficient oxygen reduction reaction (ORR) catalyst. The introduction of WC is found to promote the ORR activity of Fe/Co-based carbide electrocatalysts, which is attributed to the synergistic catalysts of WC, Fe{sub 3}C, and Co{sub 3}C. Results suggest that the composite exhibits comparable electrocatalytic activity, higher durability, and ability for methanol tolerance compared with commercial Pt/C for ORR in alkaline electrolyte. These advantages make Fe/Co/WC@NC a promising ORR electrocatalyst and a cost-effective alternative to Pt/C for practical application as fuel cell.

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.

Corrosion studies of thermally sensitised AGR fuel element brace in pH7 and pH9.2 borate solutions

International Nuclear Information System (INIS)

Tyfield, S.P.; Smith, C.A.

1987-04-01

Brace and cladding of AGR fuel elements sensitised in reactor are susceptible to intergranular and crevice corrosion, which may initiate in the pH7 borate pond storage environment of CEGB/SSEB stations. This report considers the benefit in corrosion control that is provided by raising the pond solution pH to 9.2, whilst maintaining the boron level at 1250 gm -3 . The greater corrosion protection provided by pH9.2 solution compared to the pH7 borate solution is demonstrated by a series of tests with non-active laboratory sensitised brace samples exposed to solutions dosed with chloride or sulphate in order to promote localised corrosion. The corrosion tests undertaken consisted of 5000 hour immersions at 32 0 C and shorter term electrochemically monitored experiments (rest potential, impedance, anodic current) generally conducted at 22 0 C. The pH9.2 solution effectively inhibited the initiation of crevice and intergranular corrosion in the presence of low levels of chloride and sulphate, whereas the pH7 solution did not always do so. However, the pH9.2 solution, dosed with 40 gm -3 chloride, failed to suppress fully crevice corrosion initiated in unborated 40 gm -3 chloride solution at 22 0 C. Fluoride is not deleterious at low levels ∼ 10 gm -3 in the borate solutions. The significant improvement in corrosion control demonstrated for the change from pH7 to pH9.2 borate solution on laboratory sensitised brace samples should ideally be confirmed using complete irradiated AGR fuel elements. (U.K.)

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

Corrosion resistance of Ultra-Low-Carbon 19% Cr-11% Ni stainless steel for nuclear fuel reprocessing plants in nitric acid

International Nuclear Information System (INIS)

Ariga, Tamako; Takagi, Yoshio; Inazumi, Toru; Masamura, Katsumi; Sukekawa, M.

1995-01-01

An Ultra-Low-Carbon 19% Cr-11% Ni Stainless Steels used in nuclear fuel reprocessing plants where highly corrosion resistance in nitric acid is required has been developed. This steel has optimized the chemistry composition to decrease inclusions and deformation-induced martensitic transformation. The formation of deformation-induced martensite has the potential danger of accelerating corrosion in nitric acid. In this paper, effects of cold reduction and martensitic transformation on corrosion resistance of Ultra-Low-Carbon Stainless Steels in nitric acid are discussed. The developed steel showed excellent corrosion resistance during long-term exposure to nitric acid. (author)

Molten fuel-coolant interaction behaviours of various fast reactor fuels (Paper No. HMT-45-87)

International Nuclear Information System (INIS)

Doshi, J.B.

1987-01-01

A parametric computational model of molten fuel-coolant interaction (MFCI) including a particle size distribution is developed and employed to analyse behaviours of various possible reactor fuels, such as oxide, carbide and metal in MFCI scenario. It is observed that while higher thermal conductivity and lower specific heat of carbide compared to oxide is responsible for higher peak pressure and work done per unit mass, the trend is not observed in the metal fuel. The reason for this is the lower operation temperature and latent heat of metallic fuel. (author). 9 refs., 1 fig

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)

Stainless steel corrosion in conditions simulating WWER-1000 primary coolant. Corrosion behaviour in mixed core

International Nuclear Information System (INIS)

Krasnorutskij, V.S.; Petel'guzov, I.A.; Gritsina, V.M.; Zuek, V.A.; Tret'yakov, M.V.; Rud', R.A.; Svichkar', N.V.; Slabospitskaya, E.A.; Ishchenko, N.I.

2011-01-01

Research into corrosion kinetics of austenitic stainless steels (06Cr18Ni10Ti, 08Cr18Ni10Ti, 12Cr18Ni10Ti) in medium which corresponds to composition and parameters of WWER-1000 primary coolant with different pH values in autoclave out-pile conditions during 14000 hours is given. Surface of oxide films on stainless steels is investigated. Visual inspection of Westinghouse and TVEL fuel was carried out after 4 cycles in WWER-1000 primary water chemistry conditions at South Ukraine NPP. Westinghouse and TVEL fuel cladding materials possess high corrosion resistance. Blushing of weldments was observed. No visual corrosion defects or deposits were observed on fuel rods.

Thermal-hydraulics and neutronics studies on the FP7 CP-ESFR oxide and carbide cores

Energy Technology Data Exchange (ETDEWEB)

Ammirabile, L.; Tsige-Tamirat, H. [European Commission, JRC, Inst. for Energy, Petten (Netherlands)

2011-07-01

In the framework of the the Collaborative Project on European Sodium Fast Reactor (CP-ESFR) two core designs that are currently being proposed for the 3600 MWth sodium-cooled reactor concept: one is based on oxide fuel and the other on carbide fuel. Using the European Safety Assessment Platform (ESAP), JRC-IE has conducted static calculation on neutronics (incl. reactivity coefficients) and thermal-hydraulic characteristics for both oxide and carbide reference cores. The quantities evaluated include: keff, coolant heat-up, void, and Doppler reactivity coefficients, axial and radial expansion reactivity coefficients, pin-by-pin calculated power profiles, average and peak channel temperatures. This paper presents the ESAP models applied in the study together with the relevant results for the oxide and carbide core. (author)

Thermal-hydraulics and neutronics studies on the FP7 CP-ESFR oxide and carbide cores

International Nuclear Information System (INIS)

Ammirabile, L.; Tsige-Tamirat, H.

2011-01-01

In the framework of the the Collaborative Project on European Sodium Fast Reactor (CP-ESFR) two core designs that are currently being proposed for the 3600 MWth sodium-cooled reactor concept: one is based on oxide fuel and the other on carbide fuel. Using the European Safety Assessment Platform (ESAP), JRC-IE has conducted static calculation on neutronics (incl. reactivity coefficients) and thermal-hydraulic characteristics for both oxide and carbide reference cores. The quantities evaluated include: keff, coolant heat-up, void, and Doppler reactivity coefficients, axial and radial expansion reactivity coefficients, pin-by-pin calculated power profiles, average and peak channel temperatures. This paper presents the ESAP models applied in the study together with the relevant results for the oxide and carbide core. (author)

Zircaloy-4 corrosion in PWR's

International Nuclear Information System (INIS)

Fyfitch, S.; Smalley, W.R.; Roberts, E.

1985-01-01

Zircaloy-4 waterside corrosion has been studied extensively in the nuclear industry for a number of years. Following the early crud-related corrosion failures in the Saxton test reactor, Westinghouse undertook numerous programs to minimize crud deposition on fuel rods in power reactors through primary coolant chemistry control. Modern plants today are operating with improved coolant chemistry guidelines, and crud deposition levels are very low in proportion to earlier experience. Zircaloy-4 corrosion under a variety of coolant chemistry, heat flux and exposure conditions has been studied extensively. Experience to date, even in relatively high coolant temperature plants, has indicated that -for both fuel cladding and structural components- Zircaloy-4 waterside corrosion performance has been excellent. Recognizing future industry trends, however, which will result in Zircaloy-4 being subjected to ever increasing corrosion duties, Westinghouse will continue accumulating Zircaloy-4 corrosion experience in large power plants. 13 refs.

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

Instant release fraction corrosion studies of commercial UO{sub 2} BWR spent nuclear fuel

Energy Technology Data Exchange (ETDEWEB)

Martínez-Torrents, Albert, E-mail: albert.martinez@ctm.com.es [Fundació CTM Centre Tecnològic, Plaça de la Ciència 2, 08243 Manresa (Spain); Serrano-Purroy, Daniel [European Commission, DG Joint Research Centre - JRC, Directorate G - Nuclear Safety & Security, Department G.III, P.O. Box 2340, D-76125 Karlsruhe (Germany); Sureda, Rosa [Fundació CTM Centre Tecnològic, Plaça de la Ciència 2, 08243 Manresa (Spain); Casas, Ignasi [Department of Chemical Engineering, Universitat Politècnica de Catalunya – Barcelona Tech, Eduard Maristany 14, 08019 Barcelona (Spain); Pablo, Joan de [Fundació CTM Centre Tecnològic, Plaça de la Ciència 2, 08243 Manresa (Spain); Department of Chemical Engineering, Universitat Politècnica de Catalunya – Barcelona Tech, Eduard Maristany 14, 08019 Barcelona (Spain)

2017-05-15

The instant release fraction of a spent nuclear fuel is a matter of concern in the performance assessment of a deep geological repository since it increases the radiological risk. Corrosion studies of two different spent nuclear fuels were performed using bicarbonate water under oxidizing conditions to study their instant release fraction. From each fuel, cladded segments and powder samples obtained at different radial positions were used. The results were normalised using the specific surface area to permit a comparison between fuels and samples. Different radionuclide dissolution patterns were studied in terms of water contact availability and radial distribution in the spent nuclear fuel. The relationship between the results of this work and morphological parameters like the grain size or irradiation parameters such as the burn-up or the linear power density was studied in order to increase the understanding of the instant release fraction formation.

V and Nb Influence on the Austenitic Stainless Steel Corrosion in 0.1 M HCl

Directory of Open Access Journals (Sweden)

Amel GHARBI

2014-05-01

Full Text Available Vanadium and niobium were added in AISI309 austenitic stainless steel composition to modify their structure and pitting corrosion resistance in 0.1 M HCl. The structural characterization was carried out by X-rays diffraction and optical microscopy. Corrosion behavior was investigated using potentiodynamic tests and electrochemical impedance measurements (EIS .Results showed that vanadium and niobium addition precipitated stable carbides (VC, NbC to chromium carbides’ detriment and improved austenitic stainless steel corrosion resistance.

Electrochemical corrosion behaviour of nickel chromium-chromium carbide coating by HVOF process

Science.gov (United States)

Amudha, A.; Nagaraja, H. S.; Shashikala, H. D.

2018-04-01

To overcome the corrosion problem in marine industry, coatings are one of the most economical solutions. In this paper, the corrosion behaviour of 25(NiCr)-75Cr3C2 cermet coating on low carbon steel substrate by HVOF process is studied. Different phases such as Cr7C3 and Cr3C2, along with Ni and chromium oxide(Cr3O2) constituents present in the coating were revealed by X-Ray Diffraction (XRD) analysis. The morphology of the coating obtained by scanning electron microscope (SEM) gave confirmation for the XRD analysis. Electrochemical corrosion techniques such as Linear Polarization Resistance (LPR) and Electrochemical Impedance Spectroscopy (EIS) were used to study the corrosion behaviour of the cermet in 3.5wt% NaCl electrolyte solution. The corrosion current density of the coated sample and substrate were found to be 6.878µA/cm-2 and 21.091µA/cm-2 respectively. The Nyquist Impedance spectra were used to derive an equivalent circuit to analyze the interaction between the coating and electrolyte. The Bode Impedance plots obtained by EIS for the coating showed a typical passive material capacitive behaviour, indicated by medium to low frequency with phase angle approaching -60o, suggesting that a stable film is formed on the tested material in the electrolyte used.

ENTIRELY AQUEOUS SOLUTION-GEL ROUTE FOR THE PREPARATION OF ZIRCONIUM CARBIDE, HAFNIUM CARBIDE AND THEIR TERNARY CARBIDE POWDERS

Directory of Open Access Journals (Sweden)

Zhang Changrui

2016-07-01

Full Text Available An entirely aqueous solution-gel route has been developed for the synthesis of zirconium carbide, hafnium carbide and their ternary carbide powders. Zirconium oxychloride (ZrOCl₂.8H₂O, malic acid (MA and ethylene glycol (EG were dissolved in water to form the aqueous zirconium carbide precursor. Afterwards, this aqueous precursor was gelled and transformed into zirconium carbide at a relatively low temperature (1200 °C for achieving an intimate mixing of the intermediate products. Hafnium and the ternary carbide powders were also synthesized via the same aqueous route. All the zirconium, hafnium and ternary carbide powders exhibited a particle size of ∼100 nm.

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

Irradiation and examination results of the AC-3 mixed-carbide test

International Nuclear Information System (INIS)

Mason, R.E.; Hoth, C.W.; Stratton, R.W.; Botta, F.

1992-01-01

The AC-3 test was a cooperative Swiss/US irradiation test of mixed-carbide, (U,Pr)C, fuel pins in the Fast Flux Test Facility. The test included 25 Swiss-fabricated sphere-pac-type fuel pins and 66 U.S. fabricated pellet-type fuel pins. The test was designed to operate at prototypical fast reactor conditions to provide a direct comparison of the irradiation performance of the two fuel types. The test design and fuel fabrication processes used for the AC-3 test are presented

Rapid Tempering of Martensitic Stainless Steel AISI420: Microstructure, Mechanical and Corrosion Properties

Science.gov (United States)

Abbasi-Khazaei, Bijan; Mollaahmadi, Akbar

2017-04-01

In this research, the effect of rapid tempering on the microstructure, mechanical properties and corrosion resistance of AISI 420 martensitic stainless steel has been investigated. At first, all test specimens were austenitized at 1050 °C for 1 h and tempered at 200 °C for 1 h. Then, the samples were rapidly reheated by a salt bath furnace in a temperature range from 300 to 1050 °C for 2 min and cooled in air. The tensile tests, impact, hardness and electrochemical corrosion were carried out on the reheated samples. Scanning electron microscopy was used to study the microstructure and fracture surface. To investigate carbides, transmission electron microscopy and also scanning electron microscopy were used. X-ray diffraction was used for determination of the retained austenite. The results showed that the minimum properties such as the tensile strength, impact energy, hardness and corrosion resistance were obtained at reheating temperature of 700 °C. Semi-continuous carbides in the grain boundaries were seen in this temperature. Secondary hardening phenomenon was occurred at reheating temperature of 500 °C.

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

Corrosion behaviors and contact resistances of the low-carbon steel bipolar plate with a chromized coating containing carbides and nitrides

Energy Technology Data Exchange (ETDEWEB)

Bai, Ching-Yuan; Ger, Ming-Der [Department of Applied Chemistry and Materials Science, Chung Cheng Institute of Technology, National Defense University, Ta-His, Tao-Yuan, 335 (China); Wu, Min-Sheng [Department of Weapon System Engineering, Chung Cheng Institute of Technology, National Defense University, Ta-His, Tao-Yuan, 335 (China)

2009-08-15

This work improved the surface performance of low-carbon steel AISI 1020 by a reforming pack chromization process at low temperature (700 C) and investigated the possibility that the modified steels are used as metal bipolar plates (BPP) of PEMFCs. The steel surface was activated by electrical discharge machining (EDM) with different currents before the chromizing procedure. Experimental results indicate that a dense and homogenous Cr-rich layer is formed on the EDM carbon steels by pack chromization. The chromized coating pretreated with electrical discharge currents of 2 A has the lowest corrosion current density, 5.78 x 10{sup -8} Acm{sup -2}, evaluated by potentiodynamic polarization in a 0.5 M H{sub 2}SO{sub 4} solution and the smallest interfacial contact resistance (ICR), 11.8 m{omega}-cm{sup 2}, at 140 N/cm{sup 2}. The carbon steel with a coating containing carbides and nitrides is promising for application as metal BPPs, and this report presents the first research in producing BPPs with carbon steels. (author)

Initial report on stress-corrosion-cracking experiments using Zircaloy-4 spent fuel cladding C-rings

International Nuclear Information System (INIS)

Smith, H.D.

1988-09-01

The Nevada Nuclear Waste Storage Investigations (NNWSI) Project is sponsoring C-ring stress corrosion cracking scoping experiments as a first step in evaluating the potential for stress corrosion cracking of spent fuel cladding in a potential tuff repository environment. The objective is to scope the approximate behavior so that more precise pressurized tube testing can be performed over an appropriate range of stress, without expanding the long-term effort needlessly. The experiment consists of stressing, by compression with a dead weight load, C-rings fabricated from spent fuel cladding exposed to an environment of Well J-13 water held at 90/degree/C. The results indicate that stress corrosion cracking occurs at the high stress levels employed in the experiments. The cladding C-rings, tested at 90% of the stress at which elastic behavior is obtained in these specimens, broke in 25 to 64 d when tested in water. This was about one third of the time required for control tests to break in air. This is apparently the first observation of stress corrosion under the test conditions of relatively low temperature, benign environment but very high stress. The 150 ksi test stress could be applied as a result of the particular specimen geometry. By comparison, the uniaxial tensile yield stress is about 100 to 120 ksi and the ultimate stress is about 150 ksi. When a general model that fits the high stress results is extrapolated to lower stress levels, it indicates that the C-rings in experiments now running at /approximately/80% of the yield strength should take 200 to 225 d to break. 21 refs., 24 figs., 5 tabs

Dependence of silicon carbide coating properties on deposition parameters: preliminary report

International Nuclear Information System (INIS)

Lauf, R.J.; Braski, D.N.

1980-05-01

Fuel particles for the High-Temperature Gas-Cooled Reactor (HTGR) contain a layer of pyrolytic silicon carbide, which acts as a pressure vessel and provides containment of metallic fission products. The silicon carbide (SiC) is deposited by the thermal decomposition of methyltrichlorosilane (CH 3 SiCl 3 or MTS) in an excess of hydrogen. The purpose of the current study is to determine how the deposition variables affect the structure and properties of the SiC layer

Technology of the production of breeder fuel elements

International Nuclear Information System (INIS)

Funke, P.

1976-01-01

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

Chemical compatibility between cladding alloys and advanced fuels

International Nuclear Information System (INIS)

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

1975-05-01

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

Storage of spent fuels: implementation of a research program on the risk of waste container rupture due to stress corrosion induced by fission products

International Nuclear Information System (INIS)

Parise, M.; Walle, E.; Foct, J.

2001-01-01

The following topics were dealt with: research programm on stress corrosion of spent fuel casks materials due to fission products, such as iodine, chemical interactions with zirconium, chemical aspects of stress corrosion, rupture risk assessment

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

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

Hydrotreatment activities of supported molybdenum nitrides and carbides

Energy Technology Data Exchange (ETDEWEB)

Dolce, G.M.; Savage, P.E.; Thompson, L.T. [University of Michigan, Ann Arbor, MI (United States). Dept. of Chemical Engineering

1997-05-01

The growing need for alternative sources of transportation fuels encourages the development of new hydrotreatment catalysts. These catalysts must be active and more hydrogen efficient than the current commercial hydrotreatment catalysts. Molybdenum nitrides and carbides are attractive candidate materials possessing properties that are comparable or superior to those of commercial sulfide catalysts. This research investigated the catalytic properties of {gamma}-Al{sub 2}O{sub 3}-supported molybdenum nitrides and carbides. These catalysts were synthesized via temperature-programmed reaction of supported molybdenum oxides with ammonia or methane/hydrogen mixtures. Phase constituents and compositions were determined by X-ray diffraction, elemental analysis, and neutral activation analysis. Oxygen chemisorption was used to probe the surface properties of the catalysts. Specific activities of the molybdenum nitrides and carbides were competitive with those of a commercial sulfide catalyst for hydrodenitrogenation (HDN), hydrodesulfurization (HDS), and hydrodeoxygenation (HDO). For HDN and HDS, the catalytic activity on a molybdenum basis was a strong inverse function of the molybdenum loading. Product distributions of the HDN, HDO and HDS of a variety of heteroatom compounds indicated that several of the nitrides and carbides were more hydrogen efficient than the sulfide catalyst. 35 refs., 8 figs., 7 tabs.

Dissolution of nuclear fuel samples for analytical purposes. I

International Nuclear Information System (INIS)

Krtil, J.

1983-01-01

Main attention is devoted to procedures for dissolving fuels based on uranium metal and its alloys, uranium oxides and carbides, plutonium metal, plutonium dioxide, plutonium carbides, mixed PuC-UC carbides and mixed oxides (PuU)O 2 . Data from the literature and experience gained with the dissolution of nuclear fuel samples at the Central Control Laboratory of the Nuclear Research Institute at Rez are given. (B.S.)

Microstructure and intergranular corrosion resistance of UNS S17400 (17-4PH) stainless steel

Energy Technology Data Exchange (ETDEWEB)

Tavares, S.S.M., E-mail: ssmtavares@terra.com.b [Universidade Federal Fluminense - Programa de Pos-graduacao em Engenharia Mecanica (PGMEC), Rua Passo da Patria, 156 - CEP 24210-240 - Niteroi/RJ (Brazil); Silva, F.J. da; Scandian, C. [Universidade Federal do Espirito Santo - Departamento de Engenharia Mecanica - Av. Fernando Ferrrari, 514 - CEP 29075-910 - Vitoria/ES (Brazil); Silva, G.F. da [Universidade Federal Fluminense - Programa de Pos-graduacao em Engenharia Mecanica (PGMEC), Rua Passo da Patria, 156 - CEP 24210-240 - Niteroi/RJ (Brazil); Abreu, H.F.G. de [Universidade Federal do Ceara - Departamento de Engenharia Metalurgica e Materiais - Campus do Pici, Bloco 702 - CEP 60455-760 - Fortaleza/CE (Brazil)

2010-11-15

UNS S17400 or 17-4PH is a precipitation hardening martensitic stainless steel with many industrial applications. Quite different mechanical properties can be produced in this material by varying the aging temperature. In this work, the influence of aging temperature on the intergranular corrosion susceptibility was evaluated by electrochemical and metallographic tests. The microstructural features were investigated by X-ray diffraction, optical and scanning electron microscopy. Intergranular chromium carbide precipitation occurs in specimens aged at high temperatures, although NbC carbides were also observed. The results obtained by double loop electrochemical potentiodynamic reactivation tests (DL-EPR) show that the susceptibility to intergranular corrosion resistance increases with the increase of aging temperature. Healing due to Cr diffusion in the 600-650 {sup o}C range was not observed by DL-EPR tests.

Microstructure and intergranular corrosion resistance of UNS S17400 (17-4PH) stainless steel

International Nuclear Information System (INIS)

Tavares, S.S.M.; Silva, F.J. da; Scandian, C.; Silva, G.F. da; Abreu, H.F.G. de

2010-01-01

UNS S17400 or 17-4PH is a precipitation hardening martensitic stainless steel with many industrial applications. Quite different mechanical properties can be produced in this material by varying the aging temperature. In this work, the influence of aging temperature on the intergranular corrosion susceptibility was evaluated by electrochemical and metallographic tests. The microstructural features were investigated by X-ray diffraction, optical and scanning electron microscopy. Intergranular chromium carbide precipitation occurs in specimens aged at high temperatures, although NbC carbides were also observed. The results obtained by double loop electrochemical potentiodynamic reactivation tests (DL-EPR) show that the susceptibility to intergranular corrosion resistance increases with the increase of aging temperature. Healing due to Cr diffusion in the 600-650 o C range was not observed by DL-EPR tests.

Thermochemical Analysis of Gas-Cooled Reactor Fuels Containing Am and Pu Oxides

International Nuclear Information System (INIS)

Lindemer, T.B.

2002-01-01

Literature values and estimated data for the thermodynamics of the actinide oxides and fission products are applied to explain the chemical behavior in gas-cooled-reactor fuels. Emphasis is placed on the Am-O-C and Pu-O-C systems and the data are used to plot the oxygen chemical potential versus temperature of solid-solid and solid-gas equilibria. These results help explain observations of vaporization in Am oxides, nitrides, and carbides and provide guidance for the ceramic processing of the fuels. The thermodynamic analysis is then extended to the fission product systems and the Si-C-O system. Existing data on oxygen release (primarily as CO) as a function of burnup in the thoria-urania fuel system is reviewed and compared to values calculated from thermodynamic data. The calculations of oxygen release are then extended to the plutonia and americia fuels. Use of ZrC not only as a particle coating that may be more resistant to corrosion by Pd and other noble-metal fission products, but also as a means to getter oxygen released by fission is discussed

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

Energy Technology Data Exchange (ETDEWEB)

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

1996-05-01

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

Fission product Pd-SiC interaction in irradiated coated particle fuels

International Nuclear Information System (INIS)

Tiegs, T.N.

1980-04-01

Silicon carbide is the main barrier to fission product release from coated particle fuels. Consequently, degradation of the SiC must be minimized. Electron microprobe analysis has identified that palladium causes corrosion of the SiC in irradiated coated particles. Further ceramographic and electron microprobe examinations on irradiated particles with kernels ranging in composition from UO 2 to UC 2 , including PuO/sub 2 -x/ and mixed (Th, Pu) oxides, and in enrichment from 0.7 to 93.0% 235 U revealed that temperature is the major factor affecting the penetration rate of SiC by Pd. The effects of kernel composition, Pd concentration, other fission products, and SiC properties are secondary

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

Science.gov (United States)

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

2013-09-03

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

A systematic approach for development of a PWR cladding corrosion model

International Nuclear Information System (INIS)

Quecedo, M.; Serna, J.J.; Weiner, R.A.; Kersting, P.J.

2001-01-01

A new model for the in-reactor corrosion of Improved (low-tin) Zircaloy-4 cladding irradiated in commercial pressurized water reactors (PWRs) is described. The model is based on an extensive database of PWR fuel cladding corrosion data from fuel irradiated in commercial reactors, with a range of fuel duty and coolant chemistry control strategies which bracket current PWR fuel management practices. The fuel thermal duty with these current fuel management practices is characterized by a significant amount of sub-cooled nucleate boiling (SNB) during the fuel's residence in-core, and the cladding corrosion model is very sensitive to the coolant heat transfer models used to calculate the coolant temperature at the oxide surface. The systematic approach to developing the new corrosion model therefore began with a review and evaluation of several alternative models for the forced convection and SNB coolant heat transfer. The heat transfer literature is not sufficient to determine which of these heat transfer models is most appropriate for PWR fuel rod operating conditions, and the selection of the coolant heat transfer model used in the new cladding corrosion model has been coupled with a statistical analysis of the in-reactor corrosion enhancement factors and their impact on obtaining the best fit to the cladding corrosion data. The in-reactor corrosion enhancement factors considered in this statistical analysis are based on a review of the current literature for PWR cladding corrosion phenomenology and models. Fuel operating condition factors which this literature review indicated could have a significant effect on the cladding corrosion performance were also evaluated in detail in developing the corrosion model. An iterative least squares fitting procedure was used to obtain the model coefficients and select the coolant heat transfer models and in-reactor corrosion enhancement factors. This statistical procedure was completed with an exhaustive analysis of the model

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

Semi-empirical corrosion model for Zircaloy-4 cladding

International Nuclear Information System (INIS)

Nadeem Elahi, Waseem; Atif Rana, Muhammad

2015-01-01

The Zircaloy-4 cladding tube in Pressurize Water Reactors (PWRs) bears corrosion due to fast neutron flux, coolant temperature, and water chemistry. The thickness of Zircaloy-4 cladding tube may be decreased due to the increase in corrosion penetration which may affect the integrity of the fuel rod. The tin content and inter-metallic particles sizes has been found significantly in the magnitude of oxide thickness. In present study we have developed a Semiempirical corrosion model by modifying the Arrhenius equation for corrosion as a function of acceleration factor for tin content and accumulative annealing. This developed model has been incorporated into fuel performance computer code. The cladding oxide thickness data obtained from the Semi-empirical corrosion model has been compared with the experimental results i.e., numerous cases of measured cladding oxide thickness from UO 2 fuel rods, irradiated in various PWRs. The results of the both studies lie within the error band of 20μm, which confirms the validity of the developed Semi-empirical corrosion model. Key words: Corrosion, Zircaloy-4, tin content, accumulative annealing factor, Semi-empirical, PWR. (author)

Corrosion behavior of Fe-Si metallic coatings added with NiCrAlY in an environment of fuel oil ashes at 700 C

Energy Technology Data Exchange (ETDEWEB)

Salinas-Bravo, V.M.; Porcayo-Calderon, J.; Romero-Castanon, T. [Instituto de Investigaciones Electricas, Gerencia de Procesos Termicos., Av. Reforma 113, C.P. 62490 Col. Palmira. Temixco. Morelos (Mexico); Dominguez-Patino, G.; Gonzalez-Rodriguez, J.G. [U.A.E.M. Centro de Investigaciones en Ingenieria y Ciencias Aplicadas., Av. Universidad 1001, C.P. 62210, Col. Chamilpa. Cuernavaca, Morelos (Mexico)

2005-07-01

Electrochemical potentiodynamic polarization curves and immersion tests for 300 h at 700 C in a furnace have been used to evaluate the corrosion resistance of Fe-Si metallic coatings added with up to 50 wt.% of NiCrAIY. The corrosive environment was fuel oil ashes from a steam generator. The composition of fuel oil ashes includes high content of vanadium, sodium and sulfur. The results obtained show that only the addition of 20 wt.% NiCrAlY to the Fe-Si coating improves its corrosion resistance. The behavior of all tested coatings is explained by the results obtained from the analysis of every coating using electron microscopy and energy dispersive X-ray analysis. (Abstract Copyright [2005], Wiley Periodicals, Inc.)

The effects of applying silicon carbide coating on core reactivity of pebble-bed HTR in water ingress accident

Energy Technology Data Exchange (ETDEWEB)

Zuhair, S.; Setiadipura, Topan [National Nuclear Energy Agency of Indonesia, Serpong Tagerang Selatan (Indonesia). Center for Nuclear Reactor Technology and Safety; Su' ud, Zaki [Bandung Institute of Technology (Indonesia). Dept. of Physics

2017-03-15

Graphite is used as the moderator, fuel barrier material, and core structure in High Temperature Reactors (HTRs). However, despite its good thermal and mechanical properties below the radiation and high temperatures, it cannot avoid corrosion as a consequence of an accident of water/air ingress. Degradation of graphite as a main HTR material and the formation of dangerous CO gas is a serious problem in HTR safety. One of the several steps that can be adopted to avoid or prevent the corrosion of graphite by the water/air ingress is the application of a thin layer of silicon carbide (SiC) on the surface of the fuel element. This study investigates the effect of applying SiC coating on the fuel surfaces of pebble-bed HTR in water ingress accident from the reactivity points of view. A series of reactivity calculations were done with the Monte Carlo transport code MCNPX and continuous energy nuclear data library ENDF/B-VII at temperature of 1200 K. Three options of UO{sub 2}, PuO{sub 2}, and ThO{sub 2}/UO{sub 2} fuel kernel were considered to obtain the inter comparison of the core reactivity of pebble-bed HTR in conditions of water/air ingress accident. The calculation results indicated that the UO{sub 2}-fueled pebble-bed HTR reactivity was slightly reduced and relatively more decreased when the thickness of the SiC coating increased. The reactivity characteristic of ThO{sub 2}/UO{sub 2}-fueled pebble-bed HTR showed a similar trend to that of UO{sub 2}, but did not show reactivity peak caused by water ingress. In contrast with UO{sub 2}- and ThO{sub 2}-fueled pebble-bed HTR, although the reactivity of PuO{sub 2}-fueled pebble-bed HTR was the lowest, its characteristics showed a very high reactivity peak (0.33 Δk/k) and this introduction of positive reactivity is difficult to control. SiC coating on the surface of the plutonium fuel pebble has no significant impact. From the comparison between reactivity characteristics of uranium, thorium and plutonium cores with 0

Corrosion of an austenite and ferrite stainless steel weld

Directory of Open Access Journals (Sweden)

BRANIMIR N. GRGUR

2011-07-01

Full Text Available Dissimilar metal connections are prone to frequent failures. These failures are attributed to the difference in the mechanical properties across the weld, the coefficients of thermal expansion of the two types of steels and the resulting creep at the interface. For the weld analyzed in this research, it was shown that corrosion measurements can be used for a proper evaluation of the quality of weld material and for the prediction of whether or not the material, after the applied welding process, can be in service without failures. It was found that the corrosion of the weld analyzed in this research resulted from the simultaneous activity of different types of corrosion. In this study, electrochemical techniques including polarization and metallographic analysis were used to analyze the corrosion of a weld material of ferrite and austenitic stainless steels. Based on surface, chemical and electrochemical analyses, it was concluded that corrosion occurrence was the result of the simultaneous activity of contact corrosion (ferrite and austenitic material conjuction, stress corrosion (originating from deformed ferrite structure and inter-granular corrosion (due to chromium carbide precipitation. The value of corrosion potential of –0.53 V shows that this weld, after the thermal treatment, is not able to repassivate a protective oxide film.

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

Energy Technology Data Exchange (ETDEWEB)

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

Wear and corrosion performance of metallurgical coatings in sodium

International Nuclear Information System (INIS)

Johnson, R.N.; Farwick, D.G.

1980-01-01

The friction, wear, and corrosion performance of several metallurgical coatings in 200 to 650 0 C sodium are reviewed. Emphasis is placed on those coatings which have successfully passed the qualification tests necessary for acceptance in breeder reactor environments. Tests include friction, wear, corrosion, thermal cycling, self-welding, and irradiation exposure under as-prototypic-as-possible service conditions. Materials tested were coatings of various refractory metal carbides in metallic binders, nickel-base and cobalt-base alloys and intermetallic compounds such as the aluminides and borides. Coating processes evaluated included plasma spray, detonation gun, sputtering, spark-deposition, and solid-state diffusion

Corrosion of Ceramic Materials

Science.gov (United States)

Opila, Elizabeth J.; Jacobson, Nathan S.

1999-01-01

Non-oxide ceramics are promising materials for a range of high temperature applications. Selected current and future applications are listed. In all such applications, the ceramics are exposed to high temperature gases. Therefore it is critical to understand the response of these materials to their environment. The variables to be considered here include both the type of ceramic and the environment to which it is exposed. Non-oxide ceramics include borides, nitrides, and carbides. Most high temperature corrosion environments contain oxygen and hence the emphasis of this chapter will be on oxidation processes.

1982 Annual Status Report Plutonium Fuels and Actinide Programme

International Nuclear Information System (INIS)

Lindner, R.

1983-01-01

The programme of the Transuranium Institute has long included work on advanced fuels for fast breeder reactors. Study of the swelling of carbide and nitride fuels is now nearing completion, the retention of fission gases in bubbles of different sizes in the fuel having been quantified as function of burn-up and temperature. An important step forward has been achieved in the studies of the Equation of State of Nuclear Fuels up to 5000 K. Formation of some of the less abundant isotopes in PWR fuel has been determined experimentally. Aerosol formation during the fabrication of plutonium containing fuels, part of the activity Safe Handling of Plutonium Fuel has been studied. Head-End Processing of carbide fuels has continued experiments with high burn up mixed carbides. In the field of actinide research the preparation and characterisation of pure specimens is carried out. Effect of actinides on the properties of waste glasses is investigated

Observations of crud deposits, corrosion and erosion of BWR and PWR fuel

International Nuclear Information System (INIS)

Bairiot, H.

1983-01-01

The BWR experience is limited to one reactor but the PWR experience covers a wide range of successive generations of power plants (7 in total). The systems are described and their water chemistry briefly commented. Some R and D performed on the effects of the operating regimes (steady state and transients) are summarized. Observations made by pool-side inspections and postirradiation examinations of fuel are outlined concerning water chemistry effects (crud deposits and corrosion) and ''mechanical'' coolant-cladding interaction (chip deposits and baffle jetting). (author)

Fighting corrosion in India

Energy Technology Data Exchange (ETDEWEB)

Rajagopalan, K S; Rangaswamy, N S

1979-03-01

A survey covers the cost of corrosion in India; methods of preventing corrosion in industrial plants; some case histories, including the prevention of corrosion in pipes through which fuels are pumped to storage and the stress-corrosion cracking of evaporators in fertilizer plants; estimates of the increase in demand in 1979-89 for anticorrosion products and processes developed by the Central Electrochemical Research Institute (CECRI) at Karaikudi, India; industries that may face corrosion problems requiring assistance from CECRI, including the light and heavy engineering structural, and transport industries and the chemical industry; and some areas identified for major efforts, including the establishment of a Corrosion Advisory Board with regional centers and the expansion of the Tropical Corrosion Testing Station at Mandapam Camp, Tamil Nadu.

Fabrication of carbide and nitride pellets and the nitride irradiations Niloc 1 and Niloc 2

International Nuclear Information System (INIS)

Blank, H.

1991-01-01

Besides the relatively well-known advanced LMFBR mixed carbide fuel an advanced mixed nitride is also an attractive candidate for the optimised fuel cycle of the European Fast Reactor, but the present knowledge about the nitride is still insufficient and should be raised to the level of the carbide. For such an optimised fuel cycle the following general conditions have been set up for the fuel: (i) the burnup of the optimised MN and MC should be at least 15 a/o or even beyond, at moderate linear ratings of less than 75 kW/m (ii) the fuel will be used in a He-bonding pin concept and (iii) as far as available an advanced economic pellet fabrication method should be employed. (iv) The fuel structure must contain 15 - 20% porosity in order to accomodate the fission product swelling at high burnup. This report gives a comprehensive description of fuel and pellet fabrication and characterization, irradiation, and post-irradiation examination. From the results important conclusions can be drawn about future work on nitrides

Integrity: A semi-mechanistic model for stress corrosion cracking of fuel

Energy Technology Data Exchange (ETDEWEB)

Tayal, M; Hallgrimson, K; Macquarrie, J; Alavi, P [Atomic Energy of Canada Ltd., Mississauga, ON (Canada); Sato, S; Kinoshita, Y; Nishimura, T [Electric Power Development Co. Ltd., Tokyo (Japan)

1997-08-01

In this paper we describe the features, validation, and illustrative applications of a semi-mechanistic model, INTEGRITY, which calculates the probability of fuel defects due to stress corrosion cracking. The model expresses the defect probability in terms of fundamental parameters such as local stresses, local strains, and fission product concentration. The assessments of defect probability continue to reflect the influence of conventional parameters like ramped power, power-ramp, burnup and Canlub coating. In addition, the INTEGRITY model provides a mechanism to account for the impacts of additional factors involving detailed fuel design and reactor operation. Some examples of the latter include pellet density, pellet shape and size, sheath diameter and thickness, pellet/sheath clearance, coolant temperature and pressure, etc. The model has been fitted to a database of 554 power-ramp irradiations of CANDU fuel with and without Canlub. For this database the INTEGRITY model calculates 75 defects vs 75 actual defects. Similarly good agreements were noted in the different sub-groups of the data involving non-Canlub, thin-Canlub, and thick-Canlub fuel. Moreover, the shapes and the locations of the defect thresholds were consistent with all the above defects as well as with additional 14 ripple defects that were not in the above database. Two illustrative examples demonstrate how the defect thresholds are influenced by changes in the internal design of the fuel element and by extended burnup. (author). 19 refs, 7 figs.