WorldWideScience

Sample records for fuel matrix corrosion

  1. 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

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

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

    International Nuclear Information System (INIS)

    Krautwasser, P.; Eatherly, W.P.

    1981-01-01

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

  4. 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

  5. Spent fuel UO{sub 2} matrix corrosion behaviour studies through alpha-doped UO{sub 2} pellets leaching

    Energy Technology Data Exchange (ETDEWEB)

    Muzeau, B.; Jegou, C.; Broudic, V. [CEA-Valrho DEN/DTCD/SECM Laboratoire des Materiaux et Procedes Actifs BP 17171 F-30207 Bagnols-sur-Ceze cedex (France)

    2005-07-01

    Full text of publication follows: The option of direct disposal of spent nuclear fuel in a deep geological formation raises the need to investigate the long-term behaviour of the UO{sub 2} matrix in aqueous media subjected to {alpha}-{beta}-{gamma} radiations. The {beta}-{gamma} emitters account for the most of the activity of spent fuel at the moment it is removed from the reactor, but diminish within a millennial time frame by over three orders of magnitude to less than the long-term activity. The latter persist over much longer time periods and must therefore be taken into account over geological disposal scale. In the present investigation the UO{sub 2} matrix corrosion under alpha radiation is studied as a function of different parameters such as: the alpha activity, the carbonates and hydrogen concentrations,.. In order to study the effect of alpha radiolysis of water on the UO{sub 2} matrix, {sup 238/239}Pu doped UO{sub 2} pellets (0.22 %wt. Pu total) were fabricated with different {sup 238}Pu/{sup 239}Pu ratio to reproduce the alpha activity of a 47 GWd.t{sub HMi}{sup -1} UOX spent fuel at different milestones in time (15, 50, 1500, 10000 and 40000 years). Undoped UO{sub 2} pellets were also available as reference sample. Leaching experiments were conducted in deionized or carbonated water (NaHCO{sub 3} 1 mM), under Argon (O{sub 2} < 0.1 ppm), or Ar/H{sub 2} 30% gas mixture. Previous experiments conducted in deionized water under argon atmosphere, have shown a good correlation between alpha activity and uranium release for the 15-, 1500- and 40000-years alpha doped UO{sub 2} batches. Besides, uranium release in the leachate is controlled either by the kinetics, or by the thermodynamics. Provided the solubility limit of uranium is not achieved, uranium concentration increases and is only limited by the kinetics, unless precipitation occurs and the uranium concentration remains constant over time. These controls are highly dependant on the solution chemistry

  6. Carbonate fuel cell matrix

    Science.gov (United States)

    Farooque, Mohammad; Yuh, Chao-Yi

    1996-01-01

    A carbonate fuel cell matrix comprising support particles and crack attenuator particles which are made platelet in shape to increase the resistance of the matrix to through cracking. Also disclosed is a matrix having porous crack attenuator particles and a matrix whose crack attenuator particles have a thermal coefficient of expansion which is significantly different from that of the support particles, and a method of making platelet-shaped crack attenuator particles.

  7. 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

  8. 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)

  9. 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

  10. 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

  11. 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.

  12. 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

  13. 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

  14. 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.

  15. 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

  16. 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)

  17. 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.)

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

    International Nuclear Information System (INIS)

    BD Hanson; J Abrefah; SC Marschman; SG Prussin

    2000-01-01

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

  19. Prevention of nuclear fuel cladding materials corrosion

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  20. 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

  1. 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.

  2. 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)

  3. 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.

  4. 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)

  5. 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

  6. 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.

  7. 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.

  8. Corrosion of Graphite Aluminum Metal Matrix Composites

    Science.gov (United States)

    1991-02-01

    cathodic protection of G/AI MMCs resulted in overprotection 13. Overprotection resulted from a local increase in pH near cathodic sites during...34Cathodic Overprotection of SiC/6061-T6 and G/6061- T6 Aluminum Alloy Metal Matrix Composites," Scripta Metallurgica, 22 (1988) 413-418. 14. R

  9. 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

  10. METMET fuel with Zirconium matrix alloys

    International Nuclear Information System (INIS)

    Savchenko, A.; Konovalov, I.; Totev, T.

    2008-01-01

    The novel type of WWER-1000 fuel has been designed at A.A. Bochvar Institute. Instead of WWER-1000 UO 2 pelletized fuel rod we apply dispersion type fuel element with uniformly distributed high uranium content granules of U9Mo, U5Nb5Zr, U3Si alloys metallurgically bonded between themselves and to cladding by a specially developed Zr-base matrix alloy. The fuel meat retains a controllable porosity to accommodate fuel swelling. The optimal volume ratios between the components are: 64% fuel, 18% matrix, 18% pores. Properties of novel materials as well as fuel compositions on their base have been investigated. Method of fuel elements fabrication by capillary impregnation has been developed. The primary advantages of novel fuel are high uranium content (more than 15% in comparison with the standard UO 2 pelletized fuel rod), low temperature of fuel ( * d/tU) and serviceability under transient conditions. The use of the novel fuel might lead to natural uranium saving and reduced amounts of spent fuel as well as to optimization of Nuclear Plant operation conditions and improvements of their operation reliability and safety. As a result the economic efficiency shall increase and the cost of electric power shall decrease. (authors)

  11. 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)

  12. 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.

  13. Inert matrix fuel in dispersion type fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Savchenko, A.M. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation)]. E-mail: sav@bochvar.ru; Vatulin, A.V. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation); Morozov, A.V. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation); Sirotin, V.L. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation); Dobrikova, I.V. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation); Kulakov, G.V. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation); Ershov, S.A. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation); Kostomarov, V.P. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation); Stelyuk, Y.I. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation)

    2006-06-30

    The advantages of using inert matrix fuel (IMF) as a dispersion fuel in an aluminium alloy matrix are considered, in particular, low temperatures in the fuel centre, achievable high burn-ups, serviceability in transients and an environmentally friendly process of fuel rod fabrication. Two main versions of IMF are under development at A.A. Bochvar Institute, i.e. heterogeneous or isolated distribution of plutonium. The out-of-pile results on IMF loaded with uranium dioxide as plutonium simulator are presented. Fuel elements with uranium dioxide composition fabricated at A.A. Bochvar Institute are currently under MIR tests (RIAR, Dimitrovgrad). The fuel elements reached a burn-up of 88 MW d kg{sup -1} (equivalent to the burn up of the standard uranium dioxide pelletized fuel) without loss of leak-tightness of the cladding. The feasibility of fabricating IMF of these particular types with plutonium dioxide is considered with a view to in-pile irradiation.

  14. Inert matrix fuel in dispersion type fuel elements

    Science.gov (United States)

    Savchenko, A. M.; Vatulin, A. V.; Morozov, A. V.; Sirotin, V. L.; Dobrikova, I. V.; Kulakov, G. V.; Ershov, S. A.; Kostomarov, V. P.; Stelyuk, Y. I.

    2006-06-01

    The advantages of using inert matrix fuel (IMF) as a dispersion fuel in an aluminium alloy matrix are considered, in particular, low temperatures in the fuel centre, achievable high burn-ups, serviceability in transients and an environmentally friendly process of fuel rod fabrication. Two main versions of IMF are under development at A.A. Bochvar Institute, i.e. heterogeneous or isolated distribution of plutonium. The out-of-pile results on IMF loaded with uranium dioxide as plutonium simulator are presented. Fuel elements with uranium dioxide composition fabricated at A.A. Bochvar Institute are currently under MIR tests (RIAR, Dimitrovgrad). The fuel elements reached a burn-up of 88 MW d kg-1 (equivalent to the burn up of the standard uranium dioxide pelletized fuel) without loss of leak-tightness of the cladding. The feasibility of fabricating IMF of these particular types with plutonium dioxide is considered with a view to in-pile irradiation.

  15. Electrometallurgical treatment of aluminum-matrix fuels

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  16. 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.)

  17. 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

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

    International Nuclear Information System (INIS)

    2009-12-01

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

  19. 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

  20. 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

  1. Reducing Actinide Production Using Inert Matrix Fuels

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-08-23

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

  2. 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)

  3. 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

  4. 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.

  5. 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

  6. 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

  7. 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)

  8. 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)

  9. 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

  10. Corrosion behaviour of 2124 aluminium alloy-silicon carbide metal matrix composites in sodium chloride environment

    International Nuclear Information System (INIS)

    Singh, Nirbhay; Vadera, K.K.; Ramesh Kumar, A.V.; Singh, R.S.; Monga, S.S.; Mathur, G.N.

    1999-01-01

    Aluminium alloy based particle reinforced metal matrix composites (MMCs) are being considered for a range of applications. Their mechanical properties have been investigated in detail, but more information about their corrosion resistance is needed. In this investigation, the corrosion behaviour of silicon carbide particulates (SiC p )-2124 aluminium metal matrix composites was studied in 3 wt% sodium chloride solution by means of electrochemical technique and optical microscope. The effects of weight percentages and particle size of silicon carbide particulates on corrosion behaviour of the composite were studied in NaCl and it was observed that corrosion rate increases linearly with the increasing weight percentage of SiC p . The corrosion rate of the MMC increases by increasing the size of SiC particles. Anodization improved corrosion resistance of the composites. (author)

  11. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-05-01

    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.

  13. Molten carbonate fuel cell integral matrix tape and bubble barrier

    International Nuclear Information System (INIS)

    Reiser, C.A.; Maricle, D.L.

    1983-01-01

    A molten carbonate fuel cell matrix material is described made up of a matrix tape portion and a bubble barrier portion. The matrix tape portion comprises particles inert to molten carbonate electrolyte, ceramic particles and a polymeric binder, the matrix tape being flexible, pliable and having rubber-like compliance at room temperature. The bubble barrier is a solid material having fine porosity preferably being bonded to the matrix tape. In operation in a fuel cell, the polymer binder burns off leaving the matrix and bubble barrier providing superior sealing, stability and performance properties to the fuel cell stack

  14. Corrosion Tests of LWR Fuels - Nuclide Release

    International Nuclear Information System (INIS)

    P.A. Finn; Y. Tsai; J.C. Cunnane

    2001-01-01

    Two BWR fuels [64 and 71 (MWd)/kgU], one of which contained 2% Gd, and two PWR fuels [30 and 45 (MWd)/kgU], are tested by dripping groundwater on the fuels under oxidizing and hydrologically unsaturated conditions for times ranging from 2.4 to 8.2 yr at 90 C. The 99 Tc, 129 I, 137 Cs, 97 Mo, and 90 Sr releases are presented to show the effects of long reaction times and of gadolinium on nuclide release. This investigation showed that the five nuclides at long reaction times have similar fractional release rates and that the presence of 2% Gd reduced the 99 Tc cumulative release fraction by about an order of magnitude over that of a fuel with a similar burnup

  15. Helium in inert matrix dispersion fuels

    International Nuclear Information System (INIS)

    Veen, A. van; Konings, R.J.M.; Fedorov, A.V.

    2003-01-01

    The behaviour of helium, an important decay product in the transmutation chains of actinides, in dispersion-type inert matrix fuels is discussed. A phenomenological description of its accumulation and release in CERCER and CERMET fuel is given. A summary of recent He-implantation studies with inert matrix metal oxides (ZrO 2 , MgAl 2 O 4 , MgO and Al 2 O 3 ) is presented. A general picture is that for high helium concentrations helium and vacancy defects form helium clusters which convert into over-pressurized bubbles. At elevated temperature helium is released from the bubbles. On some occasions thermal stable nano-cavities or nano-pores remain. On the basis of these results the consequences for helium induced swelling and helium storage in oxide matrices kept at 800-1000 deg. C will be discussed. In addition, results of He-implantation studies for metal matrices (W, Mo, Nb and V alloys) will be presented. Introduction of helium in metals at elevated temperatures leads to clustering of helium to bubbles. When operational temperatures are higher than 0.5 melting temperature, swelling and helium embrittlement might occur

  16. Carbon fuel cells with carbon corrosion suppression

    Science.gov (United States)

    Cooper, John F [Oakland, CA

    2012-04-10

    An electrochemical cell apparatus that can operate as either a fuel cell or a battery includes a cathode compartment, an anode compartment operatively connected to the cathode compartment, and a carbon fuel cell section connected to the anode compartment and the cathode compartment. An effusion plate is operatively positioned adjacent the anode compartment or the cathode compartment. The effusion plate allows passage of carbon dioxide. Carbon dioxide exhaust channels are operatively positioned in the electrochemical cell to direct the carbon dioxide from the electrochemical cell.

  17. 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.

  18. 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...

  19. 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.

  20. 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.

  1. 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.

  2. 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

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

    Science.gov (United States)

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

    2018-03-01

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

  4. 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)

  5. 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)

  6. 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)

  7. 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

  8. Corrosion performance of SiCsubp/6061 Al metal matrix composites in sodium chloride solution

    International Nuclear Information System (INIS)

    Mohmad Soib bin Selamat

    1995-01-01

    The corrosion performance of silicon carbide particle/aluminium metal matrix composites (SiCsubp/Al) were studied in sodium chloride solution by means of electrochemical, microscopic, gravimetric and analytical techniques. The materials under investigation were compocasting processed 6061 Al reinforced with increasing amounts of SiC particles. Potentiostatic polarization tests were done in 0.1M NaCl solutions that were aerated or deaerated to observe overall corrosion behaviour. It was seen that the corrosion potentials did not vary greatly in relation to the amounts of SiCsubp reinforcement. Corrosion tests showed that the degree of corrosion increased with increasing SiCsubp content. SEM analysis technique was used to study the corroded samples and the pitting morphology. By TEM, no intermetallic layer was found at SiC/Al interface. A model for pitting process was proposed

  9. 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

  10. 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.

  11. 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)

  12. 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.

  13. 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

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  15. Advances in HTR fuel matrix technology

    International Nuclear Information System (INIS)

    Voice, E.H.; Sturge, D.W.

    1974-02-01

    Progress in the materials and technology of matrix consolidation in recent years is summarised, noting especially the development of an improved resin and the introduction of a new graphite powder. An earlier irradiation programme, the Matrix Test Series, is recalled and the fabrication of the most recent experiment, the directly-cooled homogeneous Met. VI, is described. (author)

  16. Vascular Canals in Permanent Hyaline Cartilage: Development, Corrosion of Nonmineralized Cartilage Matrix, and Removal of Matrix Degradation Products.

    Science.gov (United States)

    Gabner, Simone; Häusler, Gabriele; Böck, Peter

    2017-06-01

    Core areas in voluminous pieces of permanent cartilage are metabolically supplied via vascular canals (VCs). We studied cartilage corrosion and removal of matrix degradation products during the development of VCs in nose and rib cartilage of piglets. Conventional staining methods were used for glycosaminoglycans, immunohistochemistry was performed to demonstrate collagens types I and II, laminin, Ki-67, von Willebrand factor, VEGF, macrophage marker MAC387, S-100 protein, MMPs -2,-9,-13,-14, and their inhibitors TIMP1 and TIMP2. VCs derived from connective tissue buds that bulged into cartilage matrix ("perichondrial papillae", PPs). Matrix was corroded at the tips of PPs or resulting VCs. Connective tissue stromata in PPs and VCs comprised an axial afferent blood vessel, peripherally located wide capillaries, fibroblasts, newly synthesized matrix, and residues of corroded cartilage matrix (collagen type II, acidic proteoglycans). Multinucleated chondroclasts were absent, and monocytes/macrophages were not seen outside the blood vessels. Vanishing acidity characterized areas of extracellular matrix degradation ("preresorptive layers"), from where the dismantled matrix components diffused out. Leached-out material stained in an identical manner to intact cartilage matrix. It was detected in the stroma and inside capillaries and associated downstream veins. We conclude that the delicate VCs are excavated by endothelial sprouts and fibroblasts, whilst chondroclasts are specialized to remove high volumes of mineralized cartilage. VCs leading into permanent cartilage can be formed by corrosion or inclusion, but most VCs comprise segments that have developed in either of these ways. Anat Rec, 300:1067-1082, 2017. © 2016 Wiley Periodicals, Inc. © 2016 Wiley Periodicals, Inc.

  17. 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)

  18. 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

  19. Improved graphite matrix for coated-particle fuel

    International Nuclear Information System (INIS)

    Schell, D.H.; Davidson, K.V.

    1978-10-01

    An experimental process was developed to incorporate coated fuel particles in an extruded graphite matrix. This structure, containing 41 vol% particles, had a high matrix density, >1.6 g/cm 3 , and a matrix conductivity three to four times that of a pitch-injected fuel rod at 1775 K. Experiments were conducted to determine the uniformity of particle loadings in extrusions. Irradiation specimens were supplied for five tests in the High-Fluence Isotope Reactor at the Oak Ridge National Laboratory

  20. INERT-MATRIX FUEL: ACTINIDE ''BURNING'' AND DIRECT DISPOSAL

    International Nuclear Information System (INIS)

    Rodney C. Ewing; Lumin Wang

    2002-01-01

    Excess actinides result from the dismantlement of nuclear weapons (Pu) and the reprocessing of commercial spent nuclear fuel (mainly 241 Am, 244 Cm and 237 Np). In Europe, Canada and Japan studies have determined much improved efficiencies for burnup of actinides using inert-matrix fuels. This innovative approach also considers the properties of the inert-matrix fuel as a nuclear waste form for direct disposal after one-cycle of burn-up. Direct disposal can considerably reduce cost, processing requirements, and radiation exposure to workers

  1. 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

  2. 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.

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

    Science.gov (United States)

    Zhu, Jun

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

  4. 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)

  5. 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

  6. Electrolyte matrix for molten carbonate fuel cells

    Science.gov (United States)

    Huang, C.M.; Yuh, C.Y.

    1999-02-09

    A matrix is described for a carbonate electrolyte including a support material and an additive constituent having a relatively low melting temperature and a relatively high coefficient of thermal expansion. The additive constituent is from 3 to 45 weight percent of the matrix and is formed from raw particles whose diameter is in a range of 0.1 {micro}m to 20 {micro}m and whose aspect ratio is in a range of 1 to 50. High energy intensive milling is used to mix the support material and additive constituent during matrix formation. Also disclosed is the use of a further additive constituent comprising an alkaline earth containing material. The further additive is mixed with the support material using high energy intensive milling. 5 figs.

  7. Electrolyte matrix for molten carbonate fuel cells

    Science.gov (United States)

    Huang, Chao M.; Yuh, Chao-Yi

    1999-01-01

    A matrix for a carbonate electrolyte including a support material and an additive constituent having a relatively low melting temperature and a relatively high coefficient of thermal expansion. The additive constituent is from 3 to 45 weight percent of the matrix and is formed from raw particles whose diameter is in a range of 0.1 .mu.m to 20 .mu.m and whose aspect ratio is in a range of 1 to 50. High energy intensive milling is used to mix the support material and additive constituent during matrix formation. Also disclosed is the use of a further additive constituent comprising an alkaline earth containing material. The further additive is mixed with the support material using high energy intensive milling.

  8. 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.

  9. Corrosion of Continuous Fiber Reinforced Aluminum Metal Matrix Composites (CF-AMCs)

    Science.gov (United States)

    Tiwari, Shruti

    The first objective of this research is to study the atmospheric corrosion behavior of continuous reinforced aluminum matrix composites (CF-AMCs). The materials used for this research were alumina (Al2O3) and nickel (Ni) coated carbon (C) fibers reinforced AMCs. The major focus is to identify the correlation between atmospheric parameters and the corrosion rates of CF-AMCs in the multitude of microclimates and environments in Hawai'i. The micro-structures of CF-AMCs were obtained to correlate the microstructures with their corrosion performances. Also electrochemical polarization experiments were conducted in the laboratory to explain the corrosion mechanism of CF-AMCs. In addition, CF-AMCs were exposed to seven different test sites for three exposure periods. The various climatic conditions like temperature (T), relative humidity (RH), rainfall (RF), time of wetness (TOW), chloride (Cl- ) and sulfate (SO42-) deposition rate, and pH were monitored for three exposure period. Likewise, mass losses of CF-AMCs at each test site for three exposure periods were determined. The microstructure of the CF-AMCS showed that Al/C/50f MMCs contained a Ni-rich phase in the matrix, indicating that the Ni coating on the C fiber dissolved in the matrix. The intermetallic phases obtained in Al-2wt% Cu/Al 2O3/50f-T6 MMC and Al-2wt%-T6 monolith were rich in Cu and Fe. The intermetallic phases obtained in Al 7075/Al2O3/50f-T6 MMC and Al 7075-T6 monolith also contained traces of Mg, Zn, Ni, and Si. Electrochemical polarization experiment indicated that the Al/Al 2O3/50f Al-2wt% Cu/Al2O3/50f-T6 and Al 7075/Al2O3/50f-T6 MMC showed similar corrosion trends as their respective monoliths pure Al, Al-2wt%-T6 and Al 7075-T6 in both aerated and deaerated condition. Al2O3 fiber, being an insulator, did not have a great effect on the polarization behavior of the composites. Al/C/50f MMCs corroded at a much faster rate as compared to pure Al monolith due to the galvanic effect between C and Al

  10. Characterization of graphite-matrix pulsed reactor fuels

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  11. Metal Matrix Microencapsulated Fuel Technology for LWR Applications

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  12. 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

  13. The corrosion of spent UO2 fuel in synthetic groundwater

    International Nuclear Information System (INIS)

    Forsyth, R.S.; Werme, L.D.; Bruno, J.

    1985-10-01

    Leaching of high burnup BWR fuel for up to 3 years showed that both U and Pu attain saturation rapidly at pH 8.1, giving values of 1-2 mg/l and 1 μg/l respectively. The leaching rate for Sr-90 decreased from about 10 -5 /d to 10 -7 /d but was always higher than the rates for U, Pu, Cm, Ce, Eu and Ru. Congruent dissolution was only attained at pH values of about 4. When reducing conditions were imposed on the pH 8.1 groundwater by means of H 2 /Ar in the presence of a Pd catalyst, significanly lower leach rates were attained. The hypothesis that alpha radiolytic decomposition of water is a driving force for UO 2 corrosion even under reducing conditions has been examined in leaching tests on low burnup (low alpha dose-rate) fuel. No significant effect of alpha radiolysis under the experimental conditions was detected. Thermodynamically the calculated uranium solubilities in the pH range 4-8.2 generally agreed, well with the measured ones, although assumptions made for certain parameters in the calculations limit the validity of the results. (Author)

  14. 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

  15. 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)

  16. 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)

  17. 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

  18. 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

  19. 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)

  20. 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)

  1. Metal Matrix Composite Coatings of Cupronickel Embedded with Nanoplatelets for Improved Corrosion Resistant Properties

    Directory of Open Access Journals (Sweden)

    Casey R. Thurber

    2018-01-01

    Full Text Available The deterioration of metals under the influence of corrosion is a costly problem faced by many industries. Therefore, particle-reinforced composite coatings are being developed in different technological fields with high demands for corrosion resistance. This work studies the effects of nanoplatelet reinforcement on the durability, corrosion resistance, and mechanical properties of copper-nickel coatings. A 90 : 10 Cu-Ni alloy was coelectrodeposited with nanoplatelets of montmorillonite (Mt embedded into the metallic matrix from electrolytic baths containing 0.05, 0.10, and 0.15% Mt. X-ray diffraction of the coatings indicated no disruption of the crystal structure with addition of the nanoplatelets into the alloy. The mechanical properties of the coatings improved with a 17% increase in hardness and an 85% increase in shear adhesion strength with nanoplatelet incorporation. The measured polarization resistance increased from 11.77 kΩ·cm2 for pure Cu-Ni to 33.28 kΩ·cm2 for the Cu-Ni-0.15% Mt coating after soaking in a simulated seawater environment for 30 days. The incorporation of montmorillonite also stabilized the corrosion potential during the immersion study and increased resistance to corrosion.

  2. 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

  3. Corrosion

    Science.gov (United States)

    Slabaugh, W. H.

    1974-01-01

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

  4. 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.

  5. Optical matrix for nuclear fuel assemblies

    International Nuclear Information System (INIS)

    Romero G, M.; Gonzaga O, A.

    1996-01-01

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

  6. 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

  7. 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

  8. 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

  9. 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.

  10. 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.)

  11. 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

  12. 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

  13. Influence of Fuel-Matrix Interaction on the Deformation of U-Mo Dispersion Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Ryu, Ho Jin [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Kim, Yeon Soo [Argonne National Laboratory, Chicago (United States)

    2014-05-15

    In order to predict the fuel plate failure leading to breakaway swelling in the meat, an understanding of the effects of the fuel-matrix interaction behavior on the deformation of fuel meat is necessary. However, the effects of IL formation on the development of breakaway swelling have not been studied thoroughly. A mechanism that explains large pore growth that leads to breakaway swelling has not been included in the existing fuel performance models. In this study, the effect of the fuel-matrix interaction on large interfacial porosity development at the IL-Al interface is analyzed using both mechanistic correlations and observations from the post-irradiation examination results of U-Mo Dispersion fuels. The effects of fuel-matrix interaction on the fuel performance of U-Mo/Al Dispersion fuel were investigated. Fuel-matrix interaction bears the causes for breakaway swelling that can lead to a fuel failure under a high-power irradiation condition. Fission gas atoms are released from U-Mo particles to the interaction layer via diffusion and recoil. The fission gases released from the U-Mo and produced in the ILs are further released to the IL-Al interface by diffusion in the IL and recoil. Large pore formation at the IL-Al interface is attributed to the active diffusion of fission gas atoms in the ILs and coalescence between the small bubbles there. A model calculation showed that IL growth increases the probability of forming a breakaway swelling condition. ILs are connected to each other and the Al matrix decreases as ILs grow. When more ILs are interconnected, breakaway swelling can occur when the effective stress from the fission gas pressure in the IL-Al interfacial pore becomes larger than the yield strength of the Al matrix.

  14. 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...

  15. 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...

  16. 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)

  17. 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

  18. 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

  19. 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)

  20. 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)

  1. 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.

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

    International Nuclear Information System (INIS)

    Ewing, Rodney C.

    2003-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Ewing, Rodney C.

    2003-09-14

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

  4. 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

  5. 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.

  6. 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

  7. 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.

  8. Standardized Gasoline Compression Ignition Fuels Matrix

    KAUST Repository

    Badra, Jihad

    2018-04-03

    Direct injection compression ignition engines running on gasoline-like fuels have been considered an attractive alternative to traditional spark ignition and diesel engines. The compression and lean combustion mode eliminates throttle losses yielding higher thermodynamic efficiencies and the better mixing of fuel/air due to the longer ignition delay times of the gasoline-like fuels allows better emission performance such as nitric oxides (NOx) and particulate matter (PM). These gasoline-like fuels which usually have lower octane compared to market gasoline have been identified as a viable option for the gasoline compression ignition (GCI) engine applications due to its lower reactivity and lighter evaporation compared to diesel. The properties, specifications and sources of these GCI fuels are not fully understood yet because this technology is relatively new. In this work, a GCI fuel matrix is being developed based on the significance of certain physical and chemical properties in GCI engine operation. Those properties were chosen to be density, temperature at 90 volume % evaporation (T90) or final boiling point (FBP) and research octane number (RON) and the ranges of these properties were determined from the data reported in literature. These proposed fuels were theoretically formulated, while applying realistic constraints, using species present in real refinery streams. Finally, three-dimensional (3D) engine computational fluid dynamics (CFD) simulations were performed using the proposed GCI fuels and the similarities and differences were highlighted.

  9. Standardized Gasoline Compression Ignition Fuels Matrix

    KAUST Repository

    Badra, Jihad; Bakor, Radwan; AlRamadan, Abdullah; Almansour, Mohammed; Sim, Jaeheon; Ahmed, Ahfaz; Viollet, Yoann; Chang, Junseok

    2018-01-01

    Direct injection compression ignition engines running on gasoline-like fuels have been considered an attractive alternative to traditional spark ignition and diesel engines. The compression and lean combustion mode eliminates throttle losses yielding higher thermodynamic efficiencies and the better mixing of fuel/air due to the longer ignition delay times of the gasoline-like fuels allows better emission performance such as nitric oxides (NOx) and particulate matter (PM). These gasoline-like fuels which usually have lower octane compared to market gasoline have been identified as a viable option for the gasoline compression ignition (GCI) engine applications due to its lower reactivity and lighter evaporation compared to diesel. The properties, specifications and sources of these GCI fuels are not fully understood yet because this technology is relatively new. In this work, a GCI fuel matrix is being developed based on the significance of certain physical and chemical properties in GCI engine operation. Those properties were chosen to be density, temperature at 90 volume % evaporation (T90) or final boiling point (FBP) and research octane number (RON) and the ranges of these properties were determined from the data reported in literature. These proposed fuels were theoretically formulated, while applying realistic constraints, using species present in real refinery streams. Finally, three-dimensional (3D) engine computational fluid dynamics (CFD) simulations were performed using the proposed GCI fuels and the similarities and differences were highlighted.

  10. 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)

  11. 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.

  12. 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

  13. 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)

  14. 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

  15. 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.

  16. Status of the inert matrix fuel program at PSI

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  17. 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

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

    International Nuclear Information System (INIS)

    Christensen, H.; Sunder, S.

    1998-09-01

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

  19. 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)

  20. INFLUENCE OF FUEL-MATRIX INTERACTION ON THE BREAKAWAY SWELLING OF U-MO DISPERSION FUEL IN AL

    OpenAIRE

    HO JIN RYU; YEON SOO KIM

    2014-01-01

    In order to advance understanding of the breakaway swelling behavior of U-Mo/Al dispersion fuel under a high-power irradiation condition, the effects of fuel-matrix interaction on the fuel performance of U-Mo/Al dispersion fuel were investigated. Fission gas release into large interfacial pores between interaction layers and the Al matrix was analyzed using both mechanistic models and observations of the post-irradiation examination results of U-Mo dispersion fuels. Using the model prediction...

  1. 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.

  2. 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)

  3. 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)

  4. 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...

  5. Modelling the waterside corrosion of PWR fuel rods

    International Nuclear Information System (INIS)

    Abram, T.J.

    1997-01-01

    The mechanism of zirconium alloy cladding corrosion in PWRs is briefly reviewed, and an engineering corrosion model is proposed. The basic model is intended to produce a best-estimate fit to circumferentially-average oxide thickness measurements obtained from inter-span positions, way from the effects of structural or flow mixing grids. The model comprises an initial pre-transition weight gain expression which follows cubic rate kinetics. On reaching a critical oxide thickness, a transition to linear rate kinetics occurs. The post-transition corrosion rate includes a term which is dependent on fast neutron flux, and an Arrhenius thermal corrosion rate which has been fitted to isothermal ex-reactor data. This thermal corrosion rate is enhanced by the presence of lithium in the coolant, and by the concentration of hydrogen in the cladding. Different cladding materials are accounted for in the selection of the model constants, and results for standard Zircaloy-4, low tin (or ''optimized'') Zircaloy-4, and the Westinghouse advanced alloy ZIRLO TM are presented. A method of accounting for the effects of grids is described, and the application of the model within the ENIGMA-B and ZROX codes is discussed. (author). 35 refs, 6 figs, 3 tabs

  6. Modelling the waterside corrosion of PWR fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Abram, T J [Fuel Engineering Dept., British Nuclear Fuels plc, Salwick, Preston (United Kingdom)

    1997-08-01

    The mechanism of zirconium alloy cladding corrosion in PWRs is briefly reviewed, and an engineering corrosion model is proposed. The basic model is intended to produce a best-estimate fit to circumferentially-average oxide thickness measurements obtained from inter-span positions, way from the effects of structural or flow mixing grids. The model comprises an initial pre-transition weight gain expression which follows cubic rate kinetics. On reaching a critical oxide thickness, a transition to linear rate kinetics occurs. The post-transition corrosion rate includes a term which is dependent on fast neutron flux, and an Arrhenius thermal corrosion rate which has been fitted to isothermal ex-reactor data. This thermal corrosion rate is enhanced by the presence of lithium in the coolant, and by the concentration of hydrogen in the cladding. Different cladding materials are accounted for in the selection of the model constants, and results for standard Zircaloy-4, low tin (or ``optimized``) Zircaloy-4, and the Westinghouse advanced alloy ZIRLO{sup TM} are presented. A method of accounting for the effects of grids is described, and the application of the model within the ENIGMA-B and ZROX codes is discussed. (author). 35 refs, 6 figs, 3 tabs.

  7. Plutonium fuel lattice neutron behavior in inert matrix

    International Nuclear Information System (INIS)

    Hernandez L, H.; Lucatero, M. A.

    2010-10-01

    In several countries is had been researching the possibility of using plutonium, as weapon degree as reactor degree, as fuel material in commercial reactors to generate electricity. In special a great development has been in Pressure Water Reactors. However, in Mexico the reactors are Boiling Water Reactors type, reason for which the necessity to considers feasibility to use this fuel type in the reactors of nuclear power plant of Laguna Verde. For this propose a comparison of fuel lattice that compose a fuel assembly is made. The fuel assembly will propose to be used whit in the reactor present different inert matrix, as well as burnable poison. The material that compose the inert matrices used are cerium and zirconia (CeO 2 and ZrO 2 ) and as burnable poisons have gadolinium and erbium (Gd 2 O 4 and ErO 2 ). As far as the hydraulic design used is a cell 10 X 10 with two water channels. The lattice calculations are made with the Helios code a library with 35 energy groups, having determined the pin power factors, the infinite multiplication factor and the neutron flux profiles. (Author)

  8. 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)

  9. Crevice corrosion of titanium under nuclear fuel waste conditions

    International Nuclear Information System (INIS)

    Ikeda, B.M.; Bailey, M.G.; Clarke, C.F.; Shoesmith, D.W.

    1989-11-01

    This report describes our experimental program to investigate the localized corrosion of ASTM Grade-2 titanium. In particular, it describes the study of the crevice corrosion of titanium, the process most likely to lead to the failure of nuclear waste containers constructed from this material. The basic mechanisms of crevice corrosion are discussed in detail. This is followed by a description of our laboratory program and the various immersion tests being performed under irradiated conditions. Experiments and tests were performed in NaCl solutions (generally 1.6 wt.%) and in simulated groundwater at 100 or 150 degrees C. A mechanism for crevice corrosion of titanium is presented and justified experimentally using an electrochemical approach. During the initiation stage, the crevice reaction is controlled by the kinetics of the anodic process. As oxygen is consumed in the propagation step, control switches to the cathodic step. Crevice corrosion eventually stops when the oxygen concentration falls to a low value. Propagation of the crevice can be restarted by the addition of oxygen. Our preliminary results on the effect of varying the iron content of the titanium are presented. An increase in iron content from 0.02 wt.% to 0.13 wt.% leads to passivation, as opposed to propagation, of the crevice. The effects of γ-irradiation, temperature, and oxygen concentration are also briefly discussed. Although our conclusions must be considered tentative, the effects of γ-irradiation appear to be beneficial. some crevice corrosion rates from longer-term immersion tests are also presented. Generally the rates are very low

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Pardo, A. [Departamento de Ciencia de Materiales, Facultad de Quimica, Universidad Complutense, 28040 Madrid (Spain)]. E-mail: anpardo@quim.ucm.es; Merino, M.C. [Departamento de Ciencia de Materiales, Facultad de Quimica, Universidad Complutense, 28040 Madrid (Spain); Merino, S. [Departamento de Tecnologia Industrial, Universidad Alfonso X El Sabio, 28691, Villanueva de la Canada, Madrid (Spain); Viejo, F. [Departamento de Ciencia de Materiales, Facultad de Quimica, Universidad Complutense, 28040 Madrid (Spain); Carboneras, M. [Departamento de Ciencia de Materiales, Facultad de Quimica, Universidad Complutense, 28040 Madrid (Spain); Arrabal, R. [Departamento de Ciencia de Materiales, Facultad de Quimica, Universidad Complutense, 28040 Madrid (Spain)

    2005-07-01

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

  12. 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

  13. 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.

  14. Porous matrix structures for alkaline electrolyte fuel cells

    Science.gov (United States)

    Vine, R. W.; Narsavage, S. T.

    1975-01-01

    A number of advancements have been realized by a continuing research program to develop higher chemically stable porous matrix structures with high bubble pressure (crossover resistance) for use as separators in potassium hydroxide electrolyte fuel cells. More uniform, higher-bubble-pressure asbestos matrices were produced by reconstituting Johns-Manville asbestos paper; Fybex potassium titanate which was found compatible with 42% KOH at 250 F for up to 3000 hr; good agreement was found between bubble pressures predicted by an analytical study and those measured with filtered structures; Teflon-bonded Fybex matrices with bubble pressures greater than 30 psi were obtained by filtering a water slurry of the mixture directly onto fuel cell electrodes; and PBI fibers have satisfactory compatibility with 42% KOH at 250 F.

  15. 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

  16. 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

  17. 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

  18. 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

  19. Zirconia based inert matrix fuel: fabrication concepts and feasibility studies

    International Nuclear Information System (INIS)

    Ingold, F.; Burghartz, M.; Ledergerber, G.

    1999-01-01

    The internal gelation process has traditionally been applied to fabricate standard fuel based on uranium, typically UO2 and MOX. To meet the recent aim to destroy plutonium in the most effective way, a uranium free fuel was evaluated. The fuel development programme at PSI has been redirected toward a fuel based on zirconium oxide or a mixture of zirconia and a conducting material to form ceramic/metal (CERMET) or ceramic/ceramic (CERCER) combinations. A feasibility study was carried out to demonstrate that microspheres based on zirconia and spinel can be fabricated with the required properties. The gelation parameters were investigated to optimise compositions of the starting solutions. Studies to fabricate a composite material (from zirconia and spinel) are ongoing. If the zirconia/spinel ratio is chosen appropriately, the low thermal conductivity of pure zirconia can be compensated by the higher thermal conductivity of spinel. Another solution to offset the low thermal conductivity of zirconia is the development of a CERMET, which consists of fine particles bearing plutonium in a cubic zirconia lattice dispersed in a metallic matrix. The fabrication of such a CERMET is also being studied. (author)

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

    International Nuclear Information System (INIS)

    Lucchini, J.F.

    2001-01-01

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

  1. 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

  2. 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)

  3. 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

  4. 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

  5. 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

  6. 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)

  7. 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

  8. Influence of fuel-matrix interaction on the breakaway swelling of U-Mo dispersion fuel in Al

    Energy Technology Data Exchange (ETDEWEB)

    Ryu, Ho Jin [Dept. of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Kim, Yeon Soo [Nuclear Engineering Division, Argonne National Laboratory, Arogonne (United States)

    2014-04-15

    In order to advance understanding of the breakaway swelling behavior of U-Mo/Al dispersion fuel under a high-power irradiation condition, the effects of fuel-matrix interaction on the fuel performance of U-Mo/Al dispersion fuel were investigated. Fission gas release into large interfacial pores between interaction layers and the Al matrix was analyzed using both mechanistic models and observations of the post-irradiation examination results of U-Mo dispersion fuels. Using the model predictions, advantageous fuel design parameters are recommended to prevent breakaway swelling.

  9. Reprocessability of molybdenum and magnesia based inert matrix fuels

    Directory of Open Access Journals (Sweden)

    Ebert Elena L.

    2015-12-01

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

  10. 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)

  11. 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)

  12. 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

  13. 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)

  14. 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

  15. 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

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  17. 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.

  18. 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

  19. 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.

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

    International Nuclear Information System (INIS)

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

    1994-08-01

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

  1. Influence of reinforcement grade and matrix composition on corrosion resistance of cast aluminium matrix composites (A3xx.x/SiCp) in a humid environment

    Energy Technology Data Exchange (ETDEWEB)

    Pardo, A.; Viejo, F.; Carboneras, M. [Departamento de Ciencia de Materiales, Facultad de Quimica Universidad Complutense, 28040, Madrid (Spain); Merino, M.C. [Departamento de Ciencia de Materiales, Facultad de Quimica Universidad Complutense, 28040, Madrid (Spain); Departamento de Tecnologia Industrial, Universidad Alfonso X El Sabio, 28691, Villanueva de la Canada, Madrid (Spain); Lopez, M.D. [Escuela Superior de Ciencias Experimentales y Tecnologia, Universidad Rey Juan Carlos, 28931, Mostoles, Madrid (Spain); Merino, S. [Departamento de Tecnologia Industrial, Universidad Alfonso X El Sabio, 28691, Villanueva de la Canada, Madrid (Spain)

    2003-05-01

    A study of the influence of the silicon carbide (SiC{sub p}) proportion and the matrix concentration of four aluminium metal matrix composites (A360/SiC/10p, A360/SiC/20p, A380/SiC/10p, A380/SiC/20p) exposed to high relative humid environment was carried out under simulation in a climatic chamber. The matrix of A360/SiC/xxp composites was virtually free of copper while the A380/SiC/xxp matrix contained 3.13-3.45wt% Cu and 1.39-1.44wt% Ni. The kinetics of the corrosion process was studied on the basis of gravimetric tests. The nature of corrosion products was analysed by Scanning Electron Microscopy (SEM) and Low Angle X-Ray Diffraction (XRD) before and after accelerated testing to determine the influence of microstructural changes on corrosion behaviour during exposure to the corrosive environment. The corrosion damage to Al/SiCp composites was low at 80% Relative Humidity (RH) and increased with temperature, SiCp proportion, relative humidity and Cu matrix concentration. The main attack nucleation sites were the interface region between the matrix and the reinforcement particles. The corrosion process was influenced more by the concentration of alloy elements in the matrix than by the proportion of SiCp reinforcement. (Abstract Copyright [2003], Wiley Periodicals, Inc.) [German] Eine Studie zum Einfluss des Siliziumkarbidanteils (SiCp) und der Zusammensetzung des Grundwerkstoffs von vier Aluminiummatrixverbundwerkstoffen (A360/SiC/10p, A360/SiC/20p, A380/SiC/10p, A380/SiC/20p), die in Umgebungen mit relativ hoher Feuchtigkeit ausgelagert waren, wurde unter simulierten Bedingungen in einer Klimakammer durchgefuehrt. Die Matrix des A360/SiC/xxp-Verbundwerkstoffs war praktisch Kupfer-frei waehrend die A380/SiC/xxp Matrix 3,13-3,45 Gew.-% Cu und 1,39-1,44 Gew.-% Ni enthielt. Die Kinetik des Korrosionsprozesses wurde auf der Basis von gravimetrischen Messungen studiert. Die Beschaffenheit der Korrosionsprodukte wurde mittelt REM-Untersuchungen und

  2. 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

  3. 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.

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

    International Nuclear Information System (INIS)

    EATON, G.L.

    2000-01-01

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

  5. 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?

  6. 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

  7. 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.

  8. 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...

  9. 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)

  10. 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.

  11. 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.

  12. 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)

  13. 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.

  14. 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%

  15. Corrosion mechanisms of spent fuel under oxidizing conditions

    International Nuclear Information System (INIS)

    Finn, P.A.; Finch, R.; Buck, E.; Bates, J.

    1997-01-01

    The release of 99 Tc can be used as a reliable marker for the extent of spent oxide fuel reaction under unsaturated high-drip-rate conditions at 90 degrees C. Evidence from leachate data and from scanning and transmission electron microscopy (SEM and TEM) examination of reacted fuel samples is presented for radionuclide release, potential reaction pathways, and the formation of alteration products. In the ATM-103 fuel, 0.03 of the total inventory of 99 Tc is released in 3.7 years under unsaturated and oxidizing conditions. Two reaction pathways that have been identified from SEM are (1) through-grain dissolution with subsequent formation of uranyl alteration products, and (2) grain-boundary dissolution. The major alteration product identified by x-ray diffraction (XRD) and SEM, is Na-boltwoodite, Na[(UO 2 )(SiO 3 OH)]lg-bullet H 2 O, which is formed from sodium and silicon in the water leachant

  16. Comparison of matrix exponential methods for fuel burnup calculations

    International Nuclear Information System (INIS)

    Oh, Hyung Suk; Yang, Won Sik

    1999-01-01

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

  17. 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

  18. 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)

  19. 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

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

    International Nuclear Information System (INIS)

    Odorowski, Melina

    2015-01-01

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

  1. The corrosion behavior of CVI SiC matrix in SiC{sub f}/SiC composites under molten fluoride salt environment

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Hongda [Structural Ceramics and Composites Engineering Research Center, Shanghai Institute of Ceramics, Chinese Academy of Sciences, Shanghai 200050 (China); State Key Laboratory of High Performance Ceramics and Superfine Microstructure, Shanghai Institute of Ceramics, Chinese Academy of Sciences, Shanghai 200050 (China); School of Graduate, University of Chinese Academy of Sciences, Beijing 100049 (China); Feng, Qian [Analysis and Testing Center, Donghua University, Shanghai 201600 (China); Wang, Zhen, E-mail: jeff@mail.sic.ac.cn [Structural Ceramics and Composites Engineering Research Center, Shanghai Institute of Ceramics, Chinese Academy of Sciences, Shanghai 200050 (China); State Key Laboratory of High Performance Ceramics and Superfine Microstructure, Shanghai Institute of Ceramics, Chinese Academy of Sciences, Shanghai 200050 (China); Zhou, Haijun; Kan, Yanmei; Hu, Jianbao [Structural Ceramics and Composites Engineering Research Center, Shanghai Institute of Ceramics, Chinese Academy of Sciences, Shanghai 200050 (China); State Key Laboratory of High Performance Ceramics and Superfine Microstructure, Shanghai Institute of Ceramics, Chinese Academy of Sciences, Shanghai 200050 (China); Dong, Shaoming, E-mail: smdong@mail.sic.ac.cn [Structural Ceramics and Composites Engineering Research Center, Shanghai Institute of Ceramics, Chinese Academy of Sciences, Shanghai 200050 (China); State Key Laboratory of High Performance Ceramics and Superfine Microstructure, Shanghai Institute of Ceramics, Chinese Academy of Sciences, Shanghai 200050 (China)

    2017-04-15

    High temperature corrosion behavior and microstructural evolution of designed chemical-vapor-infiltrated SiC matrix in SiC fiber reinforced SiC ceramic matrix composites in 46.5LiF-11.5NaF-42.0KF (mol. %) eutectic salt at 800 °C for various corrosion time was studied. Worse damage was observed as extending the exposure time, with the mass loss ratio increasing from 0.716 wt. % for 50 h to 5.914 wt. % for 500 h. The mass loss rate showed a trend of first decrease and then increase with the extended corrosion exposure. Compared with the near-stoichiometric SiC matrix layers, the O-contained boundaries between deposited matrix layers and the designed Si-rich SiC matrix layers were much less corrosion resistant and preferentially corroded. Liner relationship between the mass loss ratio and the corrosion time obtained from 50 h to 300 h indicated that the corrosion action was reaction-control process. Further corrosion would lead to matrix layer exfoliation and higher mass loss ratio.

  2. 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

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

    International Nuclear Information System (INIS)

    Ollila, K.

    2011-05-01

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

  4. 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.

  5. 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.)

  6. Burnup simulations and spent fuel characteristics of ZrO{sub 2} based inert matrix fuels

    Energy Technology Data Exchange (ETDEWEB)

    Schneider, E.A. [Department of Mechanical Engineering, University of Texas, Austin, TX (United States); Deinert, M.R. [Department of Theoretical and Applied Mechanics, Cornell University, Ithaca, NY (United States)]. E-mail: mrd6@cornell.edu; Herring, S.T. [Idaho National Laboratory, Idaho Falls, ID (United States); Cady, K.B. [Department of Theoretical and Applied Mechanics, Cornell University, Ithaca, NY (United States)

    2007-03-31

    Reducing the inventory of long lived isotopes that are contained in spent nuclear fuel is essential for maximizing repository capacity and extending the lifetime of related storage. Because of their non-fertile matrices, inert matrix fuels (IMF's) could be an ideal vehicle for using light-water reactors to help decrease the inventory of plutonium and other transuranics (neptunium, americium, curium) that are contained within spent uranium oxide fuel (UOX). Quantifying the characteristics of spent IMF is therefore of fundamental importance to determining its effect on repository design and capacity. We consider six ZrO{sub 2} based IMF formulations with different transuranic loadings in a 1-8 IMF to UOX pin-cell arrangement. Burnup calculations are performed using a collision probability model where transport of neutrons through space is modeled using fuel to moderator transport and escape probabilities. The lethargy dependent neutron flux is treated with a high resolution multigroup thermalization method. The results of the reactor physics model are compared to a benchmark case performed with Montebruns and indicate that the approach yields reliable results applicable to high-level analyses of spent fuel isotopics. The data generated show that a fourfold reduction in the radiological and integrated thermal output is achievable in single recycle using IMF, as compared to direct disposal of an energy equivalent spent UOX.

  7. 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

  8. 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.

  9. 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

  10. 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

  11. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Rodney C. Ewing

    2004-10-07

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

  13. The corrosion of spent UO2-fuel in synthetic groundwater

    International Nuclear Information System (INIS)

    Forsyth, R.S.; Svanberg, K.; Werme, L.

    1983-01-01

    Segments of fuel and clad have been leached in deionized water and in groundwater. The leachants were centrifuged through membrane filters. Both centrifugate and the filters were analysed for U, Sr-90, α- and γ-emitters. The results are discussed in terms of preferential leaching, solubility limitations and adsorption effects. For U an apparent saturation at about 800 ppb was observed. Pu also appeared to attain saturation at a few ppb. For Sr the leach rate was 3x10 -7 /d after ca 400 days. Attempts to impose reducing conditions showed decreased leach rates. (Authors)

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

    International Nuclear Information System (INIS)

    Ewing, Rodney C.

    2004-01-01

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

  15. 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)

  16. 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

  17. 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

  18. 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

  19. Characterization of Corrosion on Outdoor-Exposed Aluminum Metal-Matrix Composites as a Function of Reinforcement Specie and Volume Fraction

    National Research Council Canada - National Science Library

    Adler, Ralph P; Snoha, Daniel J; Hawthorn, George; Hihara, Lloyd H

    2008-01-01

    The Hawaii Corrosion Laboratory and the U.S. Army Research Laboratory collaborated to prepare, environmentally expose for up to 2 years, and evaluate multivariant sets of metal matrix composites (MMCs...

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  1. An Experiment on the Carbonization of Fuel Compact Matrix Graphite for HTGR

    International Nuclear Information System (INIS)

    Lee, Young Woo; Kim, Joo Hyoung; Cho, Moon Sung

    2012-01-01

    The fuel element for HTGR is manufactured by mixing coated fuel particles with matrix graphite powder and forming into either pebble type or cylindrical type compacts depending on their use in different HTGR cores. The coated fuel particle, the so-called TRISO particle, consists of 500-μm spherical UO 2 particles coated with the low density buffer Pyrolytic Carbon (PyC) layer, the inner and outer high density PyC layer and SiC layer sandwiched between the two inner and outer PyC layers. The coated TRISO particles are mixed with a properly prepared matrix graphite powder, pressed into a spherical shape or a cylindrical compact, and finally heat-treated at about 1800 .deg. C. These fuel elements can have different sizes and forms of compact. The basic steps for manufacturing a fuel element include preparation of graphite matrix powder, over coating the fuel particles, mixing the fuel particles with a matrix powder, carbonizing green compact, and the final high-temperature heat treatment of the carbonized fuel compact. The carbonization is a process step where the binder that is incorporated during the matrix graphite powder preparation step is evaporated and the residue of the binder is carbonized during the heat treatment at about 1073 K, In order to develop a fuel compact fabrication technology, and for fuel matrix graphite to meet the required material properties, it is of extreme importance to investigate the relationship among the process parameters of the matrix graphite powder preparation, fabrication parameters of fuel element green compact and the carbonization condition, which has a strong influence on further steps and the material properties of fuel element. In this work, the carbonization behavior of green compact samples prepared from the matrix graphite powder mixtures with different binder materials was investigated in order to elucidate the behavior of binders during the carbonization heat treatment by analyzing the change in weight, density and its

  2. 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.)

  3. 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.)

  4. 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.

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

    Science.gov (United States)

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

    1994-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-12-01

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

  7. 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

  8. 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.

  9. 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

  10. 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

  11. Rupture of Al matrix in U-Mo/Al dispersion fuel by fission induced creep

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Gwan Yoon; Sohn, Dong Seong [UNIST, Daejeon (Korea, Republic of); Kim, Yeon Soo [Argonne National Laboratory, Argonnge (United States); Lee, Kyu Hong [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    This phenomenon was found specifically in the dispersion fuel plate with Si addition in the Al matrix to suppress interaction layer (IL) formation between UMo and Al. It is known that the stresses induced by fission induced swelling in U-Mo fuel particles are relieved by creep deformation of the IL, surrounding the fuel particles, that has a much higher creep rate than the Al matrix. Thus, when IL growth is suppressed, the stress is instead exerted on the Al matrix. The observed rupture in the Al matrix is believed to be caused when the stress exceeded the rupture strength of the Al matrix. In this study, the possibility of creep rupture of the Al matrix between the neighboring U-Mo fuel particles was examined using the ABAQUS finite element analysis (FEA) tool. The predicted rupture time for a plate was much shorter than its irradiation life indicating a rupture during the irradiation. The higher stress leads Al matrix to early creep rupture in this plate for which the Al matrix with lower creep strain rate does not effectively relieve the stress caused by the swelling of the U-Mo fuel particles. For the other plate, no rupture was predicted for the given irradiation condition. The effect of creeping of the continuous phase on the state of stress is significant.

  12. 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)

  13. 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)

  14. 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

  15. Influence of silica nanospheres on corrosion behavior of magnesium matrix syntactic foam

    Science.gov (United States)

    Qureshi, W.; Kannan, S.; Vincent, S.; Eddine, N. N.; Muhammed, A.; Gupta, M.; Karthikeyan, R.; Badari, V.

    2018-04-01

    Over the years, the development of Magnesium alloys as biodegradable implants has seen significant advancements. Magnesium based materials tend to provide numerous advantages in the field of biomedical implants over existing materials such as titanium or stainless steel. The present research focuses on corrosive behavior of Magnesium reinforced with different volume percentages of Hollow Silica Nano Spheres (HSNS). These behaviors were tested in two different simulated body fluids (SBF) namely, Hank’s Buffered Saline Solution (HBSS) and Phosphate Buffered Solution (PBS). This corrosion study was done using the method of electrochemical polarization with a three-electrode configuration. Comparative studies were established by testing pure Mg which provided critical information on the effects of the reinforcing material. The HSNS reinforced Mg displayed desirable characteristics after corrosion experiments; increased corrosion resistance was witnessed with higher volume percentage of HSNS.

  16. Nonequilibrium Alloying of Aluminum for Improving the Corrosion Resistance of Graphite-Reinforced Metal Matrix Composites

    National Research Council Canada - National Science Library

    Shaw, Barbara

    1994-01-01

    .... Unfortunately, MMCs, especially Gr reinforced composites, are extremely susceptible to corrosion with severe attack in chloride-containing environments occurring in as little time as several weeks for Gr/Al composites...

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

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

    Science.gov (United States)

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

    2018-02-01

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

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

    International Nuclear Information System (INIS)

    King, F.; Stroes-Gascoyne, S.

    1996-08-01

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

  20. Corrosion of aluminium-clad spent fuel in LVR-15 research reactor storage facilities. Final report

    International Nuclear Information System (INIS)

    Splichal, K.; Berka, J.; Keilova, E.

    2006-03-01

    The corrosion of the research reactor aluminium clad spent fuel in water was investigated in two storage facilities. The standard racks were delivered by the IAEA and consisted of two aluminium alloys AA 6061 and Szav-1 coupons. Bimetallic couples create aluminium alloy and stainless steel 304 coupons. Rolled and extruded AA 6061 material was also tested. Single coupons, bimetallic and crevice couples were exposed in the at-reactor basin (ARB) and the high-level wastage pool (HLW). The water chemistry parameters were monitored and sedimentation of impurities was measured. The content of impurities of mainly Cl and SO 4 was in the range of 2 to 15 μg/l in the HLW pool; it was about one order higher in ARB. The Fe content was below 2 μg/l for both facilities. After two years of exposure the pitting was evaluated as local corrosion damage. The occurrence of pits was evaluated predominantly on the surfaces of single coupons and on the outer and inner surfaces of bimetallic and crevices coupons. No correlation was found between the pitting initiation and the type of aluminium alloys and rolled and extruded materials. In bimetallic couples the presence of stainless coupons did not have any effect on local corrosion. The depth of pits was lower than 50 μm for considerable areas of coupons and should be compared with the results of other participating institutes. (author)

  1. 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.

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

    International Nuclear Information System (INIS)

    Forsyth, R.

    1995-04-01

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

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

    International Nuclear Information System (INIS)

    Tayal, M.; Hallgrimson, K.; Macquarrie, J.; Alavi, P.; Sato, S.; Kinoshita, Y.; Nishimura, T.

    1997-01-01

    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

  4. 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

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

    International Nuclear Information System (INIS)

    Koh, S.K.

    2005-01-01

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

  6. Measurement of fuel corrosion products using planar laser-induced fluorescence

    International Nuclear Information System (INIS)

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

    1993-01-01

    Characterizing the corrosion behavior of nuclear fuel material in a high-temperature hydrogen environment is critical for ascertaining the operational performance of proposed nuclear thermal propulsion (NTP) concepts. In this paper, we describe an experimental study undertaken to develop and test non-intrusive, laser-based diagnostics for ultimately measuring the distribution of key gas-phase corrosion products expected to evolve during the exposure of NTP fuel to hydrogen. A laser ablation technique is used to produce high temperature, vapor plumes from uranium-free zirconium carbide (ZrC) and niobium carbide (NbC) forms for probing by various optical diagnostics including planar laser-induced fluorescence (PLIF). We discuss the laser ablation technique, results of plume emission measurements, and we describe both the actual and proposed planar LIF schemes for imaging constituents of the ablated ZrC and NbC plumes. Envisioned testing of the laser technique in rf-heated, high temperature gas streams is also discussed

  7. 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)

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

    International Nuclear Information System (INIS)

    Chung, H.M.

    1989-09-01

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

  9. 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

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

    International Nuclear Information System (INIS)

    Travelli, A.

    1988-01-01

    A nuclear fuel-containing plate structure for a nuclear reactor is described; such structure comprising a pair of malleable metallic non-fissionable matrix plates having confronting surfaces which are pressure bonded together and fully united to form a bonded surface, and elongated malleable wire-like fissionable fuel members separately confined and fully enclosed between the matrix plates along the interface to afford a high fuel density as well as structural integrity and effective retention of fission products. The plates have separate recesses formed in the confronting surfaces for closely receiving the wire-like fissionable fuel members. The wire-like fissionable fuel members are made of a maleable uranium alloy capable of being formed into elongated wire-like members and capable of withstanding pressure bonding. The wire-like fissionable fuel members are completely separated and isolated by fully united portions of the interface

  11. 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

  12. Improving the corrosion wear resistance of AISI 316L stainless steel by particulate reinforced Ni matrix composite alloying layer

    Science.gov (United States)

    Xu, Jiang; Zhuo, Chengzhi; Tao, Jie; Jiang, Shuyun; Liu, Linlin

    2009-01-01

    In order to overcome the problem of corrosion wear of AISI 316L stainless steel (SS), two kinds of composite alloying layers were prepared by a duplex treatment, consisting of Ni/nano-SiC and Ni/nano-SiO2 predeposited by brush plating, respectively, and subsequent surface alloying with Ni-Cr-Mo-Cu by a double glow process. The microstructure of the two kinds of nanoparticle reinforced Ni-based composite alloying layers was investigated by means of SEM and TEM. The electrochemical corrosion behaviour of composite alloying layers compared with the Ni-based alloying layer and 316L SS under different conditions was characterized by potentiodynamic polarization test and electrochemical impedance spectroscopy. Results showed that under alloying temperature (1000 °C) conditions, amorphous nano-SiO2 particles still retained the amorphous structure, whereas nano-SiC particles were decomposed and Ni, Cr reacted with SiC to form Cr6.5Ni2.5Si and Cr23C6. In static acidic solution, the corrosion resistance of the composite alloying layer with the brush plating Ni/nano-SiO2 particles interlayer is lower than that of the Ni-based alloying layer. However, the corrosion resistance of the composite alloying layer with the brush plating Ni/nano-SiO2 particles interlayer is prominently superior to that of the Ni-based alloying layer under acidic flow medium condition and acidic slurry flow condition. The corrosion resistance of the composite alloying layer with the brush plating Ni/nano-SiC particles interlayer is evidently lower than that of the Ni-based alloying layer, but higher than that of 316L SS under all test conditions. The results show that the highly dispersive nano-SiO2 particles are helpful in improving the corrosion wear resistance of the Ni-based alloying layer, whereas carbides and silicide phase are deleterious to that of the Ni-based alloying layer due to the fact that the preferential removal of the matrix around the precipitated phase takes place by the chemical

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-05-15

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

  14. Study of the uniform corrosion of an aluminium alloy used for the fuel cladding of the Jules Horowitz experimental reactor

    International Nuclear Information System (INIS)

    Wintergerst, M.

    2008-01-01

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

  15. 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

  16. Innovative inert matrix-thoria fuels for in-reactor plutonium disposition

    International Nuclear Information System (INIS)

    Vettraino, F.; Padovani, E.; Luzzi, L.; Lombardi, C.; Thoresen, H.; Oberlander, B.; Iversen, G.; Espeland, M.

    1999-01-01

    The present leading option for plutonium disposition, either civilian or weapons Pu, is to burn it in LWRs after having converted it to MOX fuel. However, among the possible types of fuel which can be envisaged to burn plutonium in LWRs, innovative U-free fuels such as inert matrix and thoria fuel are novel concept in view of a more effective and ultimate solution from both security and safety standpoint. Inert matrix fuel is an non-fertile oxide fuel consisting of PuO 2 , either weapon-grade or reactor-grade, diluted in inert oxides such as for ex. stabilized ZrO 2 or MgAl 2 O 4 , its primary advantage consisting in no-production of new plutonium during irradiation, because it does not contain uranium (U-free fuel) whose U-238 isotope is the departure nuclide for breeding Pu-239. Some thoria addition in the matrix (thoria-doped fuel) may be required for coping with reactivity feedback needs. The full thoria-plutonia fuel though still a U-free variant cannot be defined non-fertile any more because the U-233 generation. The advantage of such a fuel option consisting basically on a remarkable already existing technological background and a potential acceleration in getting rid of the Pu stocks. All U-free fuels are envisaged to be operated under a once-through cycle scheme being the spent fuel outlooked to be sent directly to the final disposal in deep geological formations without requiring any further reprocessing treatment, thanks to the quality-poor residual Pu and a very high chemical stability under the current fuel reprocessing techniques. Besides, inert matrix-thoria fuel technology is suitable for in-reactor MAs transmutation. An additional interest in Th containing fuel refers to applicability in ADS, the innovative accelerated driven subcritical systems, specifically aimed at plutonium, minor actnides and long lived fission products transmutation in a Th-fuel cycle scheme which enables to avoid generations of new TRUs. A first common irradiation experiment

  17. The behaviour under irradiation of molybdenum matrix for inert matrix fuel containing americium oxide (CerMet concept)

    Energy Technology Data Exchange (ETDEWEB)

    D' Agata, E., E-mail: elio.dagata@ec.europa.eu [European Commission, Joint Research Centre, Institute for Energy and Transport, P.O. Box 2, 1755 ZG Petten (Netherlands); Knol, S.; Fedorov, A.V. [Nuclear Research and Consultancy Group, P.O. Box 25, 1755 ZG Petten (Netherlands); Fernandez, A.; Somers, J. [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, 76125 Karlsruhe (Germany); Klaassen, F. [Nuclear Research and Consultancy Group, P.O. Box 25, 1755 ZG Petten (Netherlands)

    2015-10-15

    Americium is a strong contributor to the long term radiotoxicity of high activity nuclear waste. Transmutation by irradiation in nuclear reactors or Accelerator Driven System (ADS, subcritical reactors dedicated to transmutation) of long-lived nuclides like {sup 241}Am is therefore an option for the reduction of radiotoxicity of waste packages to be stored in a repository. In order to safely burn americium in a fast reactor or ADS, it must be incorporated in a matrix that could be metallic (CerMet target) or ceramic (CerCer target). One of the most promising matrix to incorporate Am is molybdenum. In order to address the issues (swelling, stability under irradiation, gas retention and release) of using Mo as matrix to transmute Am, two irradiation experiments have been conducted recently at the High Flux Reactor (HFR) in Petten (The Netherland) namely HELIOS and BODEX. The BODEX experiment is a separate effect test, where the molybdenum behaviour is studied without the presence of fission products using {sup 10}B to “produce” helium, the HELIOS experiment included a more representative fuel target with the presence of Am and fission product. This paper covers the results of Post Irradiation Examination (PIE) of the two irradiation experiments mentioned above where molybdenum behaviour has been deeply investigated as possible matrix to transmute americium (CerMet fuel target). The behaviour of molybdenum looks satisfying at operating temperature but at high temperature (above 1000 °C) more investigation should be performed.

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

    International Nuclear Information System (INIS)

    2006-08-01

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

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

    DEFF Research Database (Denmark)

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

    2015-01-01

    Abstract Motivated by similar investigations recently published (Pu et al., 2015), we report a comparative corrosion study of three sets of samples relevant as bipolar plates for polymer electrolyte fuel cells: stainless steel, stainless steel with a nickel seed layer (Ni/SS) and stainless steel...

  20. Pitting corrosion behaviour study of aluminium matrix composites (A3xx.x/SiCp); Estudio del comportamiento a la corrosion por picadura de materiales compuestos de matriz de aluminio (A3xx.x/SiCp)

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-07-01

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

  1. Corrosion behavior and pitting susceptibility of in-situ Ti-based metallic glass matrix composites in 3.5 wt.% NaCl solutions

    Science.gov (United States)

    Xu, K. K.; Lan, A. D.; Yang, H. J.; Han, P. D.; Qiao, J. W.

    2017-11-01

    The Ti62Zr12V13Cu4Be9, Ti58Zr16V10Cu4Be12, Ti46Zr20V12Cu5Be17, and Ti40Zr24V12Cu5Be19 metallic glass matrix composites (MGMCs) were prepared by copper mould casting. The corrosion resistance and the pitting susceptibility of Ti-based MGMCs were tested on their cross-sectional areas in 3.5 wt.% NaCl solutions by potentiodynamic polarization measurements. The composites with lower Ti contents (Ti40Zr24V12Cu5Be19 and Ti46Zr20V12Cu5Be17) exhibit a low resistance to the chloride induced pitting and local corrosion. The preferential dissolution of amorphous matrix is explained by the high chemical reactivity of beryllium element compared to that of stable dendrites and by the detected lower Ti and V contents. However, fairly good passivity was found in the composite with higher Ti contents (Ti62Zr12V13Cu4Be9). XPS measurements revealed that protective Ti-enriched oxide film was formed on the composite surface, additionally, lower content of beryllium element in amorphous matrix hinder the selective corrosion of amorphous matrix. The assessment of experimental observation leads to a proposed corrosion mechanism involving selective dissolution of amorphous matrix and chloride induced pitting process.

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

    International Nuclear Information System (INIS)

    Durazzo, M.; Ramanathan, L.V.

    1984-01-01

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

  3. Coated powder for electrolyte matrix for carbonate fuel cell

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  4. 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)

  5. A computational technique to identify the optimal stiffness matrix for a discrete nuclear fuel assembly model

    International Nuclear Information System (INIS)

    Park, Nam-Gyu; Kim, Kyoung-Joo; Kim, Kyoung-Hong; Suh, Jung-Min

    2013-01-01

    Highlights: ► An identification method of the optimal stiffness matrix for a fuel assembly structure is discussed. ► The least squares optimization method is introduced, and a closed form solution of the problem is derived. ► The method can be expanded to the system with the limited number of modes. ► Identification error due to the perturbed mode shape matrix is analyzed. ► Verification examples show that the proposed procedure leads to a reliable solution. -- Abstract: A reactor core structural model which is used to evaluate the structural integrity of the core contains nuclear fuel assembly models. Since the reactor core consists of many nuclear fuel assemblies, the use of a refined fuel assembly model leads to a considerable amount of computing time for performing nonlinear analyses such as the prediction of seismic induced vibration behaviors. The computational time could be reduced by replacing the detailed fuel assembly model with a simplified model that has fewer degrees of freedom, but the dynamic characteristics of the detailed model must be maintained in the simplified model. Such a model based on an optimal design method is proposed in this paper. That is, when a mass matrix and a mode shape matrix are given, the optimal stiffness matrix of a discrete fuel assembly model can be estimated by applying the least squares minimization method. The verification of the method is completed by comparing test results and simulation results. This paper shows that the simplified model's dynamic behaviors are quite similar to experimental results and that the suggested method is suitable for identifying reliable mathematical model for fuel assemblies

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

    International Nuclear Information System (INIS)

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

    2000-01-01

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

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  8. Selection of dissolution process for spent fuels and preparation of corrosion test solution simulated to dissolver (contract research)

    International Nuclear Information System (INIS)

    Motooka, Takafumi; Terakado, Shogo; Koya, Toshio; Hamada, Shozo; Kiuchi, Kiyoshi

    2001-03-01

    In order to evaluate the reliability of reprocessing equipment materials used in the Rokkasho Reprocessing Plant, we have proceeded a mock-up test and laboratory tests for getting corrosion parameters. In a dissolver made of zirconium, the simulation of test solutions to the practical solution which includes the high concentration of radioactive elements such as FP and TRU is one of the important issues with respect to the life prediction. On this experiment, the dissolution process of spent fuels and the preparation of test solution for evaluating the corrosion resistance of dissolver materials were selected. These processes were tested in the No.3 cell of WASTEF. The test solution for corrosion tests was prepared by adjusting the uranium and nitric acid concentrations. (author)

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

    International Nuclear Information System (INIS)

    Peacock, H.B. Jr.

    1999-01-01

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

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

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  11. Irradiation of inert matrix and mixed oxide fuel in the Halden test reactor

    International Nuclear Information System (INIS)

    Hellwig, Ch.; Kasemeyer, U.

    2001-01-01

    In a new type of fuel, called Inert Matrix Fuel (IMF), plutonium is embedded in a U-free matrix. This offers advantages for more efficient plutonium consumption, higher proliferation resistance, and for inert behaviour later in a waste repository. In the fuel type investigated at PSI, plutonium is dissolved in yttrium-stabilized zirconium oxide (YSZ), a highly radiation-resistant cubic phase, with addition of erbium as burnable poison for reactivity control. A first irradiation experiment of YSZ-based IMF is ongoing in the OECD Material Test Reactor in Halden (HBWR), together with MOX fuel (Rig IFA-651.1). The experiment is described herein and results are presented of the first 120 days of irradiation with an average assembly burnup of 47 kWd/cm 3 . The results are compared with neutronic calculations performed before the experiment, and are used to model the fuel behaviour with the PSI-modified TRANSURANUS code. The measured fuel temperatures are within the expected range. An unexpectedly strong densification of the IMF during the first irradiation cycle does not alter the fuel temperatures. An explanation for this behaviour is proposed. The irradiation at higher linear heat rates during forthcoming cycles will deliver information about the fission gas release behaviour of the IMF. (author)

  12. Irradiation of inert matrix and mixed oxide fuel in the Halden test reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hellwig, Ch.; Kasemeyer, U

    2001-03-01

    In a new type of fuel, called Inert Matrix Fuel (IMF), plutonium is embedded in a U-free matrix. This offers advantages for more efficient plutonium consumption, higher proliferation resistance, and for inert behaviour later in a waste repository. In the fuel type investigated at PSI, plutonium is dissolved in yttrium-stabilized zirconium oxide (YSZ), a highly radiation-resistant cubic phase, with addition of erbium as burnable poison for reactivity control. A first irradiation experiment of YSZ-based IMF is ongoing in the OECD Material Test Reactor in Halden (HBWR), together with MOX fuel (Rig IFA-651.1). The experiment is described herein and results are presented of the first 120 days of irradiation with an average assembly burnup of 47 kWd/cm{sup 3}. The results are compared with neutronic calculations performed before the experiment, and are used to model the fuel behaviour with the PSI-modified TRANSURANUS code. The measured fuel temperatures are within the expected range. An unexpectedly strong densification of the IMF during the first irradiation cycle does not alter the fuel temperatures. An explanation for this behaviour is proposed. The irradiation at higher linear heat rates during forthcoming cycles will deliver information about the fission gas release behaviour of the IMF. (author)

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

    Science.gov (United States)

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

    2018-02-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Wong Moreno, Adriana del Carmen

    1998-03-01

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

  15. A kinetic model for the stability of spent fuel matrix under oxic conditions

    International Nuclear Information System (INIS)

    Bruno, J.; Cera, E.; Duro, L.; Eriksen, T.E.

    1996-01-01

    A kinetic model for the UO 2 -spent fuel dissolution has been developed by integrating all the fundamental and experimental evidence about the redox buffer capacity of the UO 2 matrix itself within the methodological framework of heterogeneous redox reactions and dissolution kinetics. The purpose of the model is to define the geochemical stability of the spent fuel matrix and its resistance to internal and external disturbances. The model has been built in basis the reductive capacity (RDC) of the spent fuel/water system. A sensitivity analysis has been performed in order to identify the main parameters that affect the RDC of the system, the oxidant consumption and the radionuclide release. The number of surface co-ordination sites, the surface area to volume ratio, the kinetics of oxidants generation by radiolysis and the kinetics of oxidative dissolution of UO 2 , have been found to be the main parameters that can affect the reductive capacity of the spent fuel matrix. The model has been checked against some selected UO 2 and spent fuel dissolution data, performed under oxidizing conditions. The results are quite encouraging. (orig.)

  16. 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

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

    Directory of Open Access Journals (Sweden)

    Milene Adriane Luciano

    2014-10-01

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

  18. On the corrosion behavior of zircaloy-4 in spent fuel pools under accidental conditions

    International Nuclear Information System (INIS)

    Lavigne, O.; Shoji, T.; Sakaguchi, K.

    2012-01-01

    Highlights: ► Corrosion behavior of oxidized Zr-4 in alkaline media in presence of chloride and radical forms. ► Generation of radical forms by sonolysis of water. ► Limited increase of the passive current densities under polarization with the increase of pH and the presence of radicals. ► Decrease of the passive range of oxidized Zr-4 with presence of Cl − (E pit ∼ 0.6 V SCE ). ► Decrease of the pitting potential when oxide layer is scratched or damaged (E pit ∼ 0.16 V SCE ). - Abstract: After zircaloy cladding tubes have been subjected to irradiation in the reactor core, they are stored temporarily in spent fuel pools. In case of an accident, the integrity of the pool may be affected and the composition of the coolant may change drastically. This was the case in Fukushima Daiichi in March 2011. Successive incidents have led to an increase in the pH of the coolant and to chloride contamination. Moreover, water radiolysis may occur owing to the remnant radioactivity of the spent fuel. In this study, we propose to evaluate the corrosion behavior of oxidized Zr-4 (in autoclave at 288 °C for 32 days) in function of the pH and the presence of chloride and radical forms. The generation of radicals is achieved by the sonolysis of the solution. It appears that the increase in pH and the presence of radicals lead to an increase in current densities. However, the current densities remain quite low (depending on the conditions, between 1 and 10 μA cm −2 ). The critical parameter is the presence of chloride ions. The chloride ions widely decrease the passive range of the oxidized samples (the pitting potential is measured around +0.6 V (vs. SCE)). Moreover, if the oxide layer is scratched or damaged (which is likely under accidental conditions), the pitting potential of the oxidized sample reaches the pitting potential of the non-oxidized sample (around +0.16 V (vs. SCE)), leaving a shorter stable passive range for the Zr-4 cladding tubes.

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

    International Nuclear Information System (INIS)

    Gebhardt, O.

    1993-01-01

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

  20. 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)

  1. Current status of studies on nodular corrosion

    International Nuclear Information System (INIS)

    Yasuda, Takayoshi; Kawasaki, Satoru; Echigoya, Hironori; Kinoshita, Yutaka; Kubota, Hiroyuki; Konishi, Takao; Yamanaka, Tuneyasu.

    1993-01-01

    The studies on nodular corrosion formed on the outer surface of BWR fuel cladding tubes were reviewed. Main factors affecting the corrosion behavior were material and environmental conditions and combined effect. The effects of such material conditions as fabrication process, alloy elements, texture and surface treatment and environmental factors as neutron irradiation, thermo-hydrodynamic, water chemistry, purity of the coolant and contact with foreign metals on the corrosion phenomena were surveyed. Out-of-reactor corrosion test methods and models for the corrosion mechanism were also reviewed. Suppression of the accumulated annealing temperature during tube reduction process improved the nodular corrosion resistance of Zircaloys. Improved resistance for the nodular corrosion was reported for the unirradiated Zircaloys with some additives. Detailed irradiation test under the BWR conditions is needed to confirm the trend. Concerning the environmental factors, boiling on the cladding surface due to heat flux reduces the nodular corrosion susceptibility, while oxidizing radical generated from dissolved oxygen accelerates the corrosion. Concerning corrosion mechanisms, importance of such phenomena as the depleted zone of alloying elements in zirconium matrix, reduction of H + to H 2 in oxide layer, electrochemical property of precipitates, crystallographic anisotropy of oxidation rates were revealed. (author) 59 refs

  2. Corrosion/95 conference papers

    International Nuclear Information System (INIS)

    Anon.

    1995-01-01

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

  3. Comparative evaluation of coating techniques for the corrosion protection of disposal container for spent nuclear fuel

    International Nuclear Information System (INIS)

    Chun, Kwan Sik; Kim, Sung Soo; Park, Chong Mook; Choi, Jong Won

    2005-02-01

    To propose a suitable coating technique to prevent corrosion on metal or metal alloys of a waste container to be used for the disposal of spent nuclear fuel, several methods related to spray coating and vapor deposition techniques have been comparatively evaluated, based on some major factors recommended. From these comparative results, it can be suggested that the best coating methods among the existing techniques in Korea would be HVOF and low pressure plasma spray. Even though the surface of the container coated by these methods would be coated, pores could be remained in the coated film. And therefore post-treatment methods for eliminating the pores have been briefly introduced to keep the life time of the container. The other techniques, the cold spray and hollow cathode discharge, may become excellent coating methods in the future if they are extensively researched to apply for coating on the container. An optimal process among the recommended methods should be selected by considering the state of container, such as an empty or a loaded container, and also related coating materials. For the support to this, the characteristics of the coating materials and the coated films and the durability of this film under a repository condition should be analyzed in detail

  4. Dimensional Behavior of Matrix Graphite Compacts during Heat Treatments for HTGR Fuel Element Fabrication

    International Nuclear Information System (INIS)

    Lee, Young-Woo; Yeo, Seunghwan; Cho, Moon Sung

    2015-01-01

    The carbonization is a process step where the binder that is incorporated during the matrix graphite powder preparation step is evaporated and the residue of the binder is carbonized during the heat treatment at about 1073 K. This carbonization step is followed by the final high temperature heat treatment where the carbonized compacts are heat treated at 2073-2173 K in vacuum for a relatively short time (about 2 hrs). In order to develop a fuel compact fabrication technology, and for fuel matrix graphite to meet the required material properties, it is essential to investigate the relationship among the process parameters of the matrix graphite powder preparation, the fabrication parameters of fuel element green compact and the heat treatments conditions, which has a strong influence on the further steps and the material properties of fuel element. In this work, the dimensional changes of green compacts during the carbonization and final heat treatment are evaluated when compacts have different densities from different pressing conditions and different final heat treatment temperatures are employed, keeping other process parameters constant, such as the binder content, carbonization time, temperature and atmosphere (two hours ant 1073K and N2 atmosphere). In this work, the dimensional variations of green compacts during the carbonization and final heat treatment are evaluated when compacts have different densities from different pressing conditions and different final heat treatment temperatures are employed

  5. Reactor fuel rod

    International Nuclear Information System (INIS)

    Inui, Mitsuhiro; Mori, Kazuma.

    1990-01-01

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

  6. Immobilization of preconditioned spent fuel from nuclear research reactors in a ceramic matrix

    International Nuclear Information System (INIS)

    Russo, Diego O.; Rodriguez, Diego S.; Heredia, Arturo D.; Sanfilippo, Miguel; Sterba, Mario E.; Mateos, Patricia

    2002-01-01

    The fuel elements from nuclear research reactors consist in a laminated sandwich of aluminum with a core of some uranium compound. To process this material its necessary to previously eliminate the aluminum covering the fuel, before the conditioning of the rest of the fuel in a stable matrix, in order to obtain an acceptable waste form for a subsequent disposition in a geological repository. Normally, mechanical and chemical methods are proposed for that purpose. One of the most developed techniques for immobilization of the radioactive elements above mentioned, is the vitrification. In this work we propose a method named CERUS (in Spanish Ceramizacion de Elementos Radiactivos con Uranio Sinterizado - Ceramization of radioactive elements with sintered uranium). This is a sinterization of the pre-treated fuel elements mixed with natural uranium oxide. The properties of the blocks obtained are adequate for final disposal in a deep geological reservoir. (author)

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

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

    Science.gov (United States)

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

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

  9. 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.)

  10. Anisotropic Material Behavior of Uni-axially Compacted Graphite Matrix for HTGR Fuel Compact Fabrication

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Young-Woo; Yeo, Seunghwan; Yoon, Ji-Hae; Cho, Moon Sung [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    In developing the fuel compact fabrication technology, and fuel graphite material to meet the required material properties, it is essential to investigate the relationship among the process parameters of the matrix graphite powder preparation, the fabrication parameters of fuel element green compact and the heat treatments conditions and the material properties of fuel element. It was observed, during this development, that the pressing technique employed for the compaction fabrication prior to the two successive heat treatments (carbonization and final high temperature heat treatment) was of extreme importance in determining the material properties of the final compact product. In this work, the material behavior of the uni-axially pressed graphite matrix during the carbonization and final heat treatment are evaluated and summarized along the different directions, viz., perpendicular and parallel directions to pressing direction. In this work, the dimensional variations and variations in thermal expansion, thermal conductivity and Vickers hardness of the graphite matrix compact samples in the axial and radial directions prepared by uni-axial pressing are evaluated, and compared with those of samples prepared by cold isostatic pressing with the available data. From this work, the followings are observed. 1) Dimensional changes of matrix graphite green compacts during carbonization show that the difference in radial and axial variations shows a large anisotropic behavior in shrinkage. The radial variation is very small while the axial variation is large. During carbonization, the stresses caused by the force would be released in to the axial direction together with the phenolic resin vapor. 2) Dimensional variation of compact samples in perpendicular and parallel directions during carbonization shows a large difference in behavior when compact sample is prepared by uni-axial pressing. However, when compact sample is prepared by cold isostatic pressing, there is

  11. Anisotropic Material Behavior of Uni-axially Compacted Graphite Matrix for HTGR Fuel Compact Fabrication

    International Nuclear Information System (INIS)

    Lee, Young-Woo; Yeo, Seunghwan; Yoon, Ji-Hae; Cho, Moon Sung

    2016-01-01

    In developing the fuel compact fabrication technology, and fuel graphite material to meet the required material properties, it is essential to investigate the relationship among the process parameters of the matrix graphite powder preparation, the fabrication parameters of fuel element green compact and the heat treatments conditions and the material properties of fuel element. It was observed, during this development, that the pressing technique employed for the compaction fabrication prior to the two successive heat treatments (carbonization and final high temperature heat treatment) was of extreme importance in determining the material properties of the final compact product. In this work, the material behavior of the uni-axially pressed graphite matrix during the carbonization and final heat treatment are evaluated and summarized along the different directions, viz., perpendicular and parallel directions to pressing direction. In this work, the dimensional variations and variations in thermal expansion, thermal conductivity and Vickers hardness of the graphite matrix compact samples in the axial and radial directions prepared by uni-axial pressing are evaluated, and compared with those of samples prepared by cold isostatic pressing with the available data. From this work, the followings are observed. 1) Dimensional changes of matrix graphite green compacts during carbonization show that the difference in radial and axial variations shows a large anisotropic behavior in shrinkage. The radial variation is very small while the axial variation is large. During carbonization, the stresses caused by the force would be released in to the axial direction together with the phenolic resin vapor. 2) Dimensional variation of compact samples in perpendicular and parallel directions during carbonization shows a large difference in behavior when compact sample is prepared by uni-axial pressing. However, when compact sample is prepared by cold isostatic pressing, there is

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-10-01

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

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

    International Nuclear Information System (INIS)

    Uctug, Fehmi Goerkem; Holmes, Stuart M.

    2011-01-01

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

  14. Mixed PWR core loadings with inert matrix Pu-fuel assemblies

    International Nuclear Information System (INIS)

    Stanculescu, A.; Kasemeyer, U.; Paratte, J.-M.; Chawla, R.

    1999-01-01

    The most efficient way to enhance plutonium consumption in light water reactors is to eliminate the production of plutonium all together. This requirement leads to fuel concepts in which the uranium is replaced by an inert matrix. At PSI, studies have focused on employing ZrO 2 as inert matrix. Adding a burnable poison to such a fuel proves to be necessary. As a result of scoping studies, Er 2 O 3 was identified as the most suitable burnable poison material. The results of whole-core three-dimensional neutronics analyses indicated, for a present-day 1000 MW e pressurised water reactor, the feasibility of an asymptotic equilibrium four-batch cycle fuelled solely with the proposed PuO 2 -Er 2 O 3 -ZrO 2 inert matrix fuel (IMF). The present paper presents the results of more recent investigations related to 'real-life' situations, which call for transition configurations in which mixed IMF and UO 2 assembly loadings must be considered. To determine the influence of the introduction of IMF assemblies on the characteristics of a UO 2 -fuelled core, three-dimensional full-core calculations have been performed for a present-day 1000 MW e PWR containing up to 12 optimised IMF assemblies. (author)

  15. Actinide transmutation using inert matrix fuels versus recycle in a low conversion fast burner reactor

    Energy Technology Data Exchange (ETDEWEB)

    Deinert, M.R.; Schneider, E.A.; Recktenwald, G.; Cady, K.B. [The Department of Mechanical Engineering, The University of Texas at Austin, 1 University Station, C2200, Austin, 78712 (United States)

    2009-06-15

    Reducing the disposal burden of the long lived radioisotopes that are contained within spent uranium oxide fuel is essential for ensuring the sustainability of nuclear power. Because of their non-fertile matrices, inert matrix fuels (IMFs) could allow light-water reactors to achieve a significant burn down of plutonium and minor actinides that are that are currently produced as a byproduct of operating light-water reactors. However, the extent to which this is possible is not yet fully understood. We consider a ZrO{sub 2} based IMF with a high transuranic loading and show that the neutron fluence (and the subsequent fuel residence time required to achieve it) present a practical limit for the achievable actinide burnup. The accumulation of transuranics in spent uranium oxide fuel is a major obstacle for the sustainability of nuclear power. While commercial light-water reactors (LWR's) produce these isotopes, they can be used to transmute them. At present, the only viable option for doing this is to partly fuel reactors with mixed oxide fuel (MOX) made using recycled plutonium. However, because of parasitic neutron capture in the uranium matrix of MOX, considerable plutonium and minor actinides are also bred as the fuel is burned. A better option is to entrain the recycled isotopes in a non-fertile matrix such as ZrO{sub 2}. Inert matrices such as these were originally envisioned for burning plutonium from dismantled nuclear weapons [1]. However, because they achieve a conversion ratio of zero, they have also been considered as a better alternative to MOX [2-6]. Plutonium and minor actinides dominate the long term heat and radiological outputs from spent nuclear fuel. Recent work has shown that that IMFs can be used to reduce these outputs by at least a factor of four, on a per unit of energy generated basis [6]. The degree of reduction is strongly dependent on IMF burnup. In principle, complete transmutation of the transuranics could be achieved though this

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

    Science.gov (United States)

    Dur, Ender

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

  17. 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)

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

    Science.gov (United States)

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

    2014-05-01

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

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

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

    Directory of Open Access Journals (Sweden)

    Bastidas, D. M.

    2006-12-01

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

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

  1. 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

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

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  3. A combinatorial matrix of rare earth chloride mixtures as corrosion inhibitors of AA2024-T3: Optimisation using potentiodynamic polarisation and EIS

    International Nuclear Information System (INIS)

    Muster, T.H.; Sullivan, H.; Lau, D.; Alexander, D.L.J.; Sherman, N.; Garcia, S.J.; Harvey, T.G.; Markley, T.A.; Hughes, A.E.; Corrigan, P.A.; Glenn, A.M.; White, P.A.; Hardin, S.G.; Mardel, J.; Mol, J.M.C.

    2012-01-01

    A combinatorial matrix of four rare earth chlorides has been evaluated for the corrosion inhibition of aluminium alloy AA2024-T3 in aqueous solution. Two electrochemical techniques, potentiodynamic polarisation (PP) and electrochemical impedance spectroscopy (EIS), were used to evaluate AA2024-T3 corrosion in 0.1 M NaCl with the addition of 10 −3 M of rare earth chloride mixtures at time periods up to 18 h. PP experiments showed rare earth inhibition of up to 98% within the first hour and thereafter corrosion rates were steadily decreased. The open-circuit potential (OCP) of AA2024-T3 decreased as a function of time for all solutions indicating predominantly cathodic inhibition. However, differing trends in the OCP were observed during PP and EIS experiments and are discussed in terms of likely time-dependent mechanisms. A comparative study of optimisation models indicated the best mixture at 10 −3 M total inhibitor concentration was predicted to be 72% cerium (Ce) and 28% (praseodymium (Pr)/lanthanum (La)) ions. As the amount of Ce is decreased from this level the corrosion inhibition is predicted to decrease also, regardless of what other rare earths (La, Pr and Nd) are added alone or in combination. Individually, La, Pr and Nd show varying levels of corrosion inhibition activity, all of which are inferior to that of Ce. If Ce is absent entirely, then a mixture of approximately 50% Pr and 50% Nd is predicted to be preferred. This is one of the first applications of combinatorial design for the optimisation of corrosion inhibitor mixtures.

  4. Effect of the Heat Treatment on the Graphite Matrix of Fuel Element for HTGR

    International Nuclear Information System (INIS)

    Lee, Chungyong; Lee, Seungjae; Suh, Jungmin; Jo, Youngho; Lee, Youngwoo; Cho, Moonsung

    2013-01-01

    In this paper, the cylinder-formed fuel element for the block type reactor is focused on, which consists of the large part of graphite matrix. One of the most important properties of the graphite matrix is the mechanical strength for the high reliability because the graphite matrix should be enabled to protect the TRISO particles from the irradiation environment and the impact from the outside. In this study, the three kinds of candidate graphites and Phenol as a binder were chosen and mixed with each other, formed and heated for the compressive strength test. The objective of this research is to optimize the kinds and composition of the mixed graphite and the forming process by evaluating the compressive strength before/after heat treatment (carbonization of binder). In this study, the effect of heat treatment on graphite matrix was studied in terms of the density and the compressive strength. The size (diameter and length) of pellet is increased by heat treatment. Due to additional weight reduction and swelling (length and diameter) of samples the density of graphite pellet is decreased from about 2.0 to about 1.7g/cm 3 . From the mechanical test results, the compressive strength of graphite pellets was related to the various conditions such as the contents of binder, the kinds of graphite and the heat treatment. Both the green pellet and the heat treated pellet, the compressive strength of G+S+P pellets is relatively higher than that of R+S+P pellets. To optimize fuel element matrix, the effect of Phenol and other binders, graphite composition and the heat treatment on the mechanical properties will be deeply investigated for further study

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

    DEFF Research Database (Denmark)

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

    2016-01-01

    friendly conditions. In order to reach these goals, and to enable a secure and nearly carbon neutral heat and power generation, recently, the Biofficiency proposal, was granted under Horizon2020, aiming to: Develop next generation, biomass-fired CHP plant, increasing the steamtemperatures up to 600°C...... and by intelligent plant design. Broaden the feedstocks for pulverized fuel (PF) and fluidized bed (FB) powerplants, using pre-treatment methods with focus on the reduction of harmful,inorganic elements: Cl, S and the alkali metals. Prevent power plant damage due to high-temperature Cl-induced corrosion. Reduce...

  6. Corrosion studies on materials of construction for spent nuclear fuel reprocessing plant equipment

    International Nuclear Information System (INIS)

    Kamachi Mudali, U.; Dayal, R.K.; Gnanamoorthy, J.B.

    1993-01-01

    Corrosion studies on specimens of commercial Type 304L stainless steel (SS), nuclear grade type 304L SS, extra low-carbon nitric acid grade (NAG) Uranus-16 SS, NAG Uranus-65 SS, Ti, Ti-5% Ta, Ti-0.25% Pd, Zircaloy-2, weldments of Ti and of Ti-5% Ta, and surface-modified (thermally oxidised and anodised) Ti were carried out to assess their corrosion resistance in nitric acid medium. The results indicated that Zircaloy-2, Ti-5% Ta, Uranus-16 SS and Uranus-65 SS have excellent corrosion resistance in boiling nitric acid solution. Specimens of Zircaloy-2, Ti-5% Ta and thermally-oxidised Ti showed excellent corrosion resistance also in a simulated uranium-containing reprocessing medium in a concentrated nitric acid solution. SEM and XRD analyses were carried out on the tested specimens to examine the scale morphology and phases present on the surface. (orig.)

  7. Influence of alkali metal oxides and alkaline earth metal oxides on the mitigation of stress corrosion cracking in CANDU fuel sheathing

    Energy Technology Data Exchange (ETDEWEB)

    Metzler, J.; Ferrier, G.A.; Farahani, M.; Chan, P.K.; Corcoran, E.C., E-mail: Joseph.Metzler@rmc.ca [Royal Military College of Canada, Kingston, ON (Canada)

    2015-07-01

    Stress corrosion cracking (SCC)can cause failures of CANDU Zircaloy-4 fuel sheathing. The process occurs when a corrosive element (i.e.,iodine) interacts with a susceptible material that is under sufficient strain at a high temperature. Currently, there is an ongoing effort to improve SCC mitigation strategies for future iterations of CANDU reactors. A potential mechanism for SCC mitigation involves utilizing alkali metal oxides and alkaline earth metal oxides that will sequester corrosive iodine while actively repairing a protective oxide layer on the sheath. SCC tests performed with sodium oxide (Na{sub 2}O) and calcium oxide (CaO) have shown to decrease significantly the sheath degradation. (author)

  8. Fabrication of BN Nanosheet Reinforced ZrO{sub 2} Composite Pellets for Inert Matrix Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Shukeir, Malik; Umer, Malik; Lee, Bin; Ryu, Ho Jin [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2014-10-15

    Plutonium also can be resulted from the dismantlement of nuclear weapons. This will result in the increase of the stockpile of plutonium. For that purpose many organizations are focusing their R-D work on the concept of Inert Matrix Fuel IMF, where a U-free matrix is used to eliminate the U-Pu conversion. R-D work was standardized around Zirconiabased IMF as a result of many screening and ranking studies performed on various candidates. Regardless of its outstanding radiation resistance, chemical stability and its high melting point, it has a very low thermal conductivity, which could be detrimental for the fuel matrix especially in case of accidents. A reinforcement phase could be used for the enhancement of the thermomechanical properties. Among many possible reinforcements, 2D structured nanosheets have emerged as an excellent candidate to enhance the thermal properties and mechanical properties simultaneously. In this approach Boron Nitride Nanosheets BNNS are used for that purpose. BNNS have a very low density, very high thermal conductivity, very high mechanical properties and high neutron absorption cross-section for Boron which is used frequently as a burnable poison. They have properties similar to graphene but they exhibit superior thermal stability in the oxide structure. Despite all the studies on other reinforcements, BNNS reinforced ZrO{sub 2} has not yet been reported. In this study, pure ZrO{sub 2} and partially stabilized Zirconia PSZ (using Yttria) ceramics are mixed with different volume fractions of BNNS.

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

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  10. Mathematical model of water transport in Bacon and alkaline matrix-type hydrogen-oxygen fuel cells

    Science.gov (United States)

    Prokopius, P. R.; Easter, R. W.

    1972-01-01

    Based on general mass continuity and diffusive transport equations, a mathematical model was developed that simulates the transport of water in Bacon and alkaline-matrix fuel cells. The derived model was validated by using it to analytically reproduce various Bacon and matrix-cell experimental water transport transients.

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

    Science.gov (United States)

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

    2013-06-01

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

  12. The possible effects of alfa and beta radiolysis on the matrix dissolution of spent nuclear fuel

    International Nuclear Information System (INIS)

    Grenthe, I.; Puigdomenech, I.; Bruno, J.

    1983-01-01

    The effects of oxidants on the retainment of actinides in a nuclear repository have been modelled by using an equilirium procedure. The oxidants are formed as a result of α- and #betta#-radiolysis when spent nuclear fuel is exposed to ground water. From an equilibrium point of view, the strongest reductants in the system (Zr, Pb and Cu) are expected to be oxidized first, leaving the actinoids in the oxidation states they have in the fuel matrix. This is expected to result in a negligible mobilization of the actinoids due to the very low solubility of the MO 2 oxides. However, the formation of protective layers of oxides will most likely decrease the effectiveness of the metallic reducing agents. This will lead to an increased oxidation of the spent fuel which results in an increased actinoid mobilization. The results of the equilibrium calculations show that the oxidation of the fuel matrix results in the formation of UO 2 (OH) 2 (s) and to the formation of the soluble complex UO 2 (CO 3 ) 3 4 . The transport of uranium is limited by the total concentration of carbonate in the aqueous phase. Neptunium may be quantitatvely solubilized as various Np(V) species and transported by ground water from the repository. Plutonium is retained at the repository site as insoluble PuO 2 . Only very small amounts are transported by ground water. The mobile actinoids may be reprecipitated when they encounter reducing conditions along the flow path. The conditions for repricipitation for typical ground water compositions have been modelled by using solubility - pe diagrams. (Authors)

  13. Exploiting the plutonium stockpiles in PWRs by using inert matrix fuel

    International Nuclear Information System (INIS)

    Lombardi, C.; Mazzola, A.

    1996-01-01

    The plutonium coming from dismantled warheads and that already stockpiled coming from spent fuel reprocessing have raised many concerns related to proliferation resistance, environmental safety and economy. The option of disposing of plutonium by fission is one of the most widely discussed and many proposals for plutonium burning in a safe and economical manner have been put forward. Due to their diffusion, PWRs appear to be the main candidates for the reduction of the plutonium stockpiles. In order to achieve a high plutonium consumption rate, a uranium-free fuel may be conceived, based on the dilution of PuO 2 within a carrier matrix made of inert oxide. In this paper, a partial loading of inert matrix fuel in a current technology PWR was investigated with 3-D calculations. The results indicated that this solution has good plutonium elimination capabilities: commercial PWRs operating in a once-through cycle scheme can transmute more than 98% of the loaded Pu-239 and 73 or 81% of the overall initially loaded reactor grade or weapons grade plutonium, respectively. The plutonium still let in the spent fuel was of poor quality and then offered a better proliferation resistance. Power peaking problems could be faced with the adoption of burnable absorbers: IFBA seemed to be particularly suitable. In spite of a reduction of the overall plutonium loaded mass by a factor 3.7 or 5.4 depending on its quality, there was no evidence of an increase of the minor actinides radiotoxicity after a time period of about 25 years. (author)

  14. Fabrication of inert matrix fuel for the incineration of plutonium - a feasibility study

    International Nuclear Information System (INIS)

    Burghartz, M.; Ledergerber, G.; Ingold, F.; Xie, T.; Botta, F.; Idemitsu, K.

    1998-01-01

    The internal gelation process has been applied to fabricate classical fuel based on uranium like UO 2 and MOX. For recent aims to destroy plutonium in the most effective way, a uranium free fuel was evaluated. The fuel development at PSI has been redirected to a fuel based on zirconium oxide or a mixture of zirconia and a conducting material leading to ceramic/metal (CERMET) or ceramic/ceramic (CERCER) combinations. A feasibility study was carried out to demonstrate that microspheres based on zirconia and spinel can be fabricated. The gelation parameters were investigated leading to optimised compositions for the starting solutions. Studies to fabricate a composite material (from zirconia and spinel) are ongoing. If the zirconia/spinel ratio is chosen appropriately, the low thermal conductivity of pure zirconia could be compensated by the higher thermal conductivity of spinel. Another solution to improve the low thermal conductivity of zirconia is the development of a CERMET, which consists of fine particles bearing plutonium in a cubic zirconia dispersed in a metallic matrix. The fabrication of such a CERMET is also being studied. (author)

  15. 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

  16. Studies of the Influence of Water Radiolysis to the Spent Fuel Matrix Dissolution Process

    International Nuclear Information System (INIS)

    Quinones, J.; Serrano, J.

    2001-01-01

    The disposal of high level radioactive waste in geological deep repositories relies on the long term stability of spent fuel matrix, which must be assured for thousands of years. One of these factors considered within the studies of performance assessment on spent fuel under final repository conditions is the effect of the radiation on its leaching behaviour. Due to the radiation from spent fuel can modify some properties of both solid phase and leachant and therefore it would alter the chemical behaviour of the near field. Particularizing in the effect of the radiation on the leachant, it will cause generation of radiolytic species that could change the redox potential of the environment and therefore may bring on variations in the leaching process. In this work, we compiled the leaching experiments performed in an irradiation facility (Nayade), in order to emulate γ radiation field of a spent fuel at different cooling times. Initial dose rate used was 0.014 (Gy/s) using source of ''60 Co. The spent fuel chemical analogue utilised was SIMFUEL (natural UO 2 doped with non-radioactive elements simulating fission products) and the leachant selected were saline and granite bentonite waters both under initial anoxic conditions. Preliminary results indicate that radiation produces an increase of the uranium dissolution rate, being the concentrations measured close to those obtained in oxic atmosphere without radiation field. In addition the solubility solid phases from experimental conditions were calculated, for both granite bentonite water and 5 m NaCl media. On the other hand, a tentative approach to model the role of γ radiolysis in these SIMFUEL tests has been carried out as well. (Author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-02-01

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

  18. 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.

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

    International Nuclear Information System (INIS)

    Smith, H.D.

    1986-03-01

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

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

    Science.gov (United States)

    Travelli, Armando

    1988-01-01

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

  1. Dependence of the specific surface area of the nuclear fuel with the matrix oxidation

    International Nuclear Information System (INIS)

    Gomez, F.; Quinones, J.; Iglesias, E.; Rodriguez, N.

    2008-01-01

    This paper is focused on the study of the changes in the specific surface area measured using BET techniques. The objective is to obtain a relation between this parameter and the change in the matrix stoichiometry (i.e., oxidation increase). None of the actual models used for extrapolating the behaviour of the spent fuel matrix under repository conditions have included this dependence yet. In this work the specific surface area of different uranium oxide were measured using N 2 (g) and Kr(g). The starting material was UO 2+x (s) with a size powder distribution lower than 20 μm. The results included in this paper shown a strong dependence on specific surface area with the matrix stoichiometry, i.e., and increase of more than one order of magnitude (SUO 2 = 6 m 2 *g -1 and SU 3 O 8 = 16.07 m 2 *g -1 ). Furthermore, the particle size distribution measured as a function of the thermal treatment done shows changes on the powder size related to the changes observed in the uranium oxide stoichiometry. (authors)

  2. 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.)

  3. 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.)

  4. Scanning electron microscopy analysis of fuel/matrix interaction layers in highly-irradiated U-Mo dispersion fuel plates with Al and Al-Si alloy matrices

    Energy Technology Data Exchange (ETDEWEB)

    Keiser, Dennis D. Jr; Jue, Jan Fong; Miller, Brandon D.; Gan, Jian; Robinson, Adom B.; Medvedev, Pavel; Madden, James; Wachs, Dan; Meyer, Mitch [Nuclear Fuels and Materials Division, Idaho National Laboratory (United States)

    2014-04-15

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

  5. 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

  6. Corrosion/Deterioration of Fuel Tank Materials Wetted in Methanolic Electrolytes

    Science.gov (United States)

    1987-12-15

    R32 DD FORM 1473, 84 MAR 83 APR edition may be used until exhausted. SECURITY CLASSIFICATION OF THIS PAGE All other editions are obsolete *U.S... aiSA A12A A1AS5!S1tit "IiaA Id~A1131 *AWFl 1t1103114W) No us" C C A C A C C A~tt A A CS Sell lA~~~l, 5158-lti A ICC :1L1 1113iC AlMeAtt2~ TIttrljACn...Corrosion Resistance of Metals and Alloys, Second Edition , Reinhold Publishing Corporation, NY, 1963, p. 217. 4. Uhlig, H. H., Coroion.and Corrosion

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

    International Nuclear Information System (INIS)

    King, F.

    1996-09-01

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

  8. Some results on development, irradiation and post-irradiation examinations of fuels for fast reactor-actinide burner (MOX and inert matrix fuel)

    International Nuclear Information System (INIS)

    Poplavsky, V.; Zabudko, L.; Moseev, L.; Rogozkin, B.; Kurina, I.

    1996-01-01

    Studies performed have shown principal feasibility of the BN-600 and BN-800 cores to achieve high efficiency of Pu burning when MOX fuel with Pu content up to 45% is used. Valuable experience on irradiation behaviour of oxide fuel with high Pu content (100%) was gained as a result of operation of two BR-10 core loadings where the maximum burnup 14 at.% was reached. Post-irradiation examination (PIE) allowed to reveal some specific features of the fuel with high plutonium content. Principal irradiation and PIE results are presented in the paper. Use of new fuel without U-238 provides the maximum burning capability as in this case the conversion ratio is reduced to zero. Technological investigations of inert matrix fuels have been continued now. Zirconium carbide, zirconium nitride, magnesium oxide and other matrix materials are under consideration. Inert matrices selection criteria are discussed in the paper. Results of technological study, of irradiation in the BOR-60 reactor and PIE results of some inert matrix fuels are summarized in this report. (author). 2 refs, 1 fig., 3 tabs

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

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  10. Aluminium oxide as an encapsulation material for unreprocessed nuclear fuel waste - evaluation from the viewpoint of corrosion

    International Nuclear Information System (INIS)

    1980-03-01

    The Nuclear Fuel Safety Project (KBS) has proposed that spent unreprocessed nuclear fuel shall be disposed of by encapsulation in canisters of high-purity alumina sintered under isostatic pressure. The canisters will have a wall thickness of 100 mm and are to be placed in vertical boreholes extending from horizontal tunnels 500 m below ground in igneous rock. In each borehole one canister is deposited embedded in a quartz sand/bentonite buffer. An expert group of 10 Swedish specialists has arrived at the following conclusions. The alumina is not thermodynamically stable in water. In pure water hydration will occur, below 100degC leading to the formation of either Al(OH) 3 in the amorphous state or crystalline gibbsite (Al 2 O 3 x 3H 2 O). Corrosion may take place by slow dissolution or flaking off of a surface layer. Various immersion tests showed that the corrosion rate will be less than 0.1 μm/year, probably one or two powers of ten lower. If the alumina canister in the storage has sufficiently large surface defects and is under sufficiently high mechanical tension the defects may grow slowly into propagating cracks, ultimately leading to fracture, so-called delayed fracture. On the basis of results from fracture mechanical studies and after introduction of safety factors with respect to possible unknown features of the delayed fracture it was judged possible to eliminate the risk of delayed fracture if the canisters pass the following production control: - Proof testing at 150 MN/m 2 , using acoustic emission technique to ensure that crack growth does not occur during the unstressing cycle. - Surface acoustic wave examination with respect to surface inclusions, canisters with inclusions larger than 100 μm within a 100 μm deep surface zone being rejected. Canisters which pass the production control mentioned are estimated to have a life of hundreds of thousands of years. (author)

  11. Characterization of the Microstructure of Irradiated U-Mo Dispersion Fuel with a Matrix that Contains Si

    International Nuclear Information System (INIS)

    Keiser, Jr. D.D.; Robinson, A.B.; Jue, J.F.; Medvedev, P.; Finlay, M.R.

    2009-01-01

    RERTR U-Mo dispersion fuel plates are being developed for application in research reactors throughout the world. Of particular interest is the irradiation performance of U-Mo dispersion fuels with Si added to the Al matrix. Si is added to improve the performance of U-Mo dispersion fuels. Microstructural examinations have been performed on fuel plates with Al-2Si matrix after irradiation to around 50% LEU burnup. Si-rich layers were observed in many areas around the various U-7Mo fuel particles. In one local area of one of the samples, where the Si-rich layer had developed into a layer devoid of Si, relatively large fission gas bubbles were observed in the interaction phase. There may be a connection between the growth of these bubbles and the amount of Si present in the interaction layer. Overall, it was found that having Si-rich layers around the fuel particles after fuel plate fabrication positively impacted the overall performance of the fuel plate

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

    International Nuclear Information System (INIS)

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

    1969-01-01

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

  13. Irradiation behavior of the interaction product of U-Mo fuel particle dispersion in an Al matrix

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yeon Soo, E-mail: yskim@anl.gov [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Hofman, G.L. [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States)

    2012-06-15

    Highlights: Black-Right-Pointing-Pointer We in-pile tested U-Mo dispersion in Al matrix. Black-Right-Pointing-Pointer We observed interaction layer growth between U-Mo and Al and pore formation there. Black-Right-Pointing-Pointer Pores degrades thermal conductivity and structural integrity of the fueled zone. Black-Right-Pointing-Pointer The amorphous behavior of interaction layers is thought to be the main reason for unstable large pore growth. Black-Right-Pointing-Pointer A mechanism for pore formation and possible remedy to prevent it are proposed. - Abstract: Irradiation performance of U-Mo fuel particles dispersed in Al matrix is stable in terms of fuel swelling and is suitable for the conversion of research and test reactors from highly enriched uranium (HEU) to low enriched uranium (LEU). However, tests of the fuel at high temperatures and high burnups revealed obstacles caused by the interaction layers forming between the fuel particle and matrix. In some cases, fission gas filled pores grow and interconnect in the interdiffusion layer resulting in fuel plate failure. Postirradiation observations are made to examine the behavior of the interdiffusion layers. The interdiffusion layers show a fluid-like behavior characteristic of amorphous materials. In the amorphous interdiffusion layers, fission gas diffusivity is high and the material viscosity is low so that the fission gas pores readily form and grow. Based on the observations, a pore formation mechanism is proposed and potential remedies to suppress the pore growth are also introduced.

  14. Heat transfer coefficient for lead matrixing in disposal containers for used reactor fuel

    International Nuclear Information System (INIS)

    Mathew, P.M.; Taylor, M.; Krueger, P.A.

    1985-02-01

    In the Canadian Nuclear Fuel Waste Management Program, metal matrices with low melting points are being evaluated for their potential to provide support for the shell of disposal containers for used fuel, and to act as an additional barrier to the release of radionuclides. The metal matrix would be incorporated into the container by casting. To study the heat transfer processes during solidification, a steady-state technique was used, involving lead as the cast metal, to determine the overall heat transfer coefficient between the lead and some of the candidate container materials. The existence of an air gap between the cast lead and the container material appeared to control the overall heat transfer coefficient. The experimental observations indicated that the surface topography of the container material influences the heat transfer and that a smoother surface results in a greater heat transfer than a rough surface. The experimental results also showed an increasing heat transfer coefficient with increasing temperature difference across the container base plates; a model developed to base-plate bending can explain the observed results

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

    Directory of Open Access Journals (Sweden)

    LIU Yongbao

    2017-03-01

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

  16. Canisters for spent-fuel disposal: Design measures against localized corrosion

    International Nuclear Information System (INIS)

    Werme, L.O.; Oversby, V.M.

    2000-01-01

    Common to all high-level-waste disposal concepts is the encapsulation of the waste into metal canisters. The purpose of this waste canister is to isolate the radioactive waste from contact with its surroundings for a desired time period. The design service life ranges from hundreds to thousands of years depending on the disposal concept. After the isolation has been breached, other barriers in the disposal system will delay and attenuate the radioactive releases to acceptable levels. In a deep geologic repository, the waste package will be exposed to chemical attack and, depending on the type of repository, to mechanical stresses. Each of these factors will by itself or in combination inevitably lead to loss of confinement some time in the future. In the design of the Swedish waste canister, the corrosion resistance is provided by an outer shell of pure copper while an insert supplies the mechanical strength cast nodular iron. The close fit between the insert and the copper results in very small tensile stresses in the copper over very limited areas once the repository has been saturated. Measurements of stress corrosion crack growth show that annealed copper cannot maintain sufficiently high stress intensity factors for cracks to grow. For annealed copper, the stress intensity factor was limited to 25 MPa·m 1/2 because of extensive plastic deformation. For cold-worked copper, no crack growth could be observed for stress intensity factors 1/2 . Through the choices of canister material, canister, and repository design, and considering the expected chemical conditions, the risks for localized corrosion can be lowered to an acceptable level, if not eliminated altogether, and the releases from prematurely failed canisters can be kept well within acceptable dose levels

  17. Corrosion/96 conference papers

    International Nuclear Information System (INIS)

    Anon.

    1996-01-01

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

  18. 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

  19. Investigation of intergranular stress corrosion cracking in the fuel pool at Three Mile Island Unit 1

    International Nuclear Information System (INIS)

    Czajkowski, C.J.

    1985-01-01

    An intergranular stress corrosion cracking failure of 304 stainless steel pipe in 2000 ppM B as H 3 BO 3 + H 2 O at 100 0 C has been investigated. Constant extension rate testing has produced an intergranular type failure in material in air. Chemical analysis was performed on both the base metal and weld material, in addition to fractography, EPR testing and optical microscopy in discerning the mode of failure. Various effects of Cl - , O 2 , and MnS are discussed. The results have indicated that the cause of failure was the severe sensitization coupled with probable contamination by S and possibly by Cl ions

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

    International Nuclear Information System (INIS)

    Mattsson, E.

    1979-01-01

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

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

    International Nuclear Information System (INIS)

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

    2012-01-01

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

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

    Science.gov (United States)

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

    2012-08-01

    CERMET fuel with either PuO2 or enriched UO2 dispersed in uranium metal matrix has a strong potential of becoming a fuel for the liquid metal cooled fast breeder reactors (LMR's). In fact it may act as a bridge between the advantages and disadvantages associated with the two extremes of fuel systems (i.e. ceramic fuel and metallic fuel) for fast reactors. At Bhabha Atomic Research Centre (BARC), R & D efforts are on to develop this CERMET fuel by powder metallurgy route. This paper describes the development of flow sheet for preparation of UO2 dispersed in uranium metal matrix pellets for three different compositions i.e. U-20 wt%UO2, U-25 wt%UO2 and U-30 wt%UO2. It was found that the sintered pellets were having excellent integrity and their linear mass was higher than that of carbide fuel pellets used in Fast Breeder Test Reactor programme (FBTR) in India. The pellets were characterized by X-ray diffraction (XRD) technique for phase analysis and lattice parameter determination. The optical microstructures were developed and reported for all the three different U-UO2 compositions.

  3. 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)

  4. Irradiation performance of U-Mo-Ti and U-Mo-Zr dispersion fuels in Al-Si matrixes

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yeon Soo, E-mail: yskim@anl.gov [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Hofman, G.L. [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Robinson, A.B.; Wachs, D.M. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Ryu, H.J.; Park, J.M.; Yang, J.H. [Korea Atomic Energy Research Institute, 150 Deokjin-dong, Yuseong-gu, Daejeon 305-353 (Korea, Republic of)

    2012-08-15

    Performance of U-7 wt.%Mo with 1 wt.%Ti, 1 wt.%Zr or 2 wt.%Zr, dispersed in an Al-5 wt.%Si alloy matrix, was investigated through irradiation tests in the ATR at INL and HANARO at KAERI. Post-irradiation metallographic features show that the addition of Ti or Zr suppresses interaction layer growth between the U-Mo and the Al-5 wt.%Si matrix. However, higher fission gas swelling was observed in the fuel with Zr addition, while no discernable effect was found in the fuel with Ti addition as compared to U-Mo without the addition. Known to have a destabilizing effect on the {gamma}-phase U-Mo, Zr, either as alloy addition or fission product, is ascribed for the disadvantageous result. Considering its benign effect on fuel swelling, with slight disadvantage from neutron economy point of view, Ti may be a better choice for this purpose.

  5. Effects of irradiation on the microstructure of U-7Mo dispersion fuel with Al-2Si matrix

    Science.gov (United States)

    Keiser, Dennis D.; Jue, Jan-Fong; Robinson, Adam B.; Medvedev, Pavel; Gan, Jian; Miller, Brandon D.; Wachs, Daniel M.; Moore, Glenn A.; Clark, Curtis R.; Meyer, Mitchell K.; Ross Finlay, M.

    2012-06-01

    The Reduced Enrichment for Research and Test Reactor (RERTR) program is developing low-enriched uranium U-Mo dispersion fuels for application in research and test reactors around the world. As part of this development, fuel plates have been irradiated in the Advanced Test Reactor and then characterized using optical metallography (OM) and scanning electron microscopy (SEM) to determine the as-irradiated microstructure. To demonstrate the irradiation performance of U-7Mo dispersion fuel plates with 2 wt.% Si added to the matrix, fuel plates were tested to moderate burnups at intermediate fission rates as part of the RERTR-6 experiment. Further testing was performed to higher fission rates as part of the RERTR-7A experiment, and very aggressive testing (high temperature, high fission density, and high fission rate) was performed in the RERTR-9A, RERTR-9B, and AFIP-1 experiments. As-irradiated microstructures were compared to those observed after fabrication to determine the effects of irradiation on the microstructure. Based on comparison of the microstructural characterization results for each irradiated sample, some general conclusions can be drawn about how the microstructure evolves during irradiation: there is growth during irradiation of the fuel/matrix interaction (FMI) layer created during fabrication; Si diffuses from the FMI layer to deeper depths in the U-7Mo particles as the irradiation conditions are made more aggressive; lowering of the Si content in the FMI layer results in an increase in the size of the fission gas bubbles; as the FMI layer grows during irradiation, more Si diffuses from the matrix to the FMI layer/matrix interface; and interlinking of fission gas bubbles in the fuel plate microstructure that may indicate breakaway swelling is not observed.

  6. Effects of irradiation on the microstructure of U-7Mo dispersion fuel with Al-2Si matrix

    Energy Technology Data Exchange (ETDEWEB)

    Keiser, Dennis D., E-mail: Dennis.Keiser@inl.gov [Nuclear Fuels and Materials Division, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Jue, Jan-Fong; Robinson, Adam B.; Medvedev, Pavel; Gan, Jian; Miller, Brandon D.; Wachs, Daniel M.; Moore, Glenn A.; Clark, Curtis R.; Meyer, Mitchell K. [Nuclear Fuels and Materials Division, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Ross Finlay, M. [Australian Nuclear Science and Technology Organization, PMB 1, Menai, NSW 2234 (Australia)

    2012-06-15

    The Reduced Enrichment for Research and Test Reactor (RERTR) program is developing low-enriched uranium U-Mo dispersion fuels for application in research and test reactors around the world. As part of this development, fuel plates have been irradiated in the Advanced Test Reactor and then characterized using optical metallography (OM) and scanning electron microscopy (SEM) to determine the as-irradiated microstructure. To demonstrate the irradiation performance of U-7Mo dispersion fuel plates with 2 wt.% Si added to the matrix, fuel plates were tested to moderate burnups at intermediate fission rates as part of the RERTR-6 experiment. Further testing was performed to higher fission rates as part of the RERTR-7A experiment, and very aggressive testing (high temperature, high fission density, and high fission rate) was performed in the RERTR-9A, RERTR-9B, and AFIP-1 experiments. As-irradiated microstructures were compared to those observed after fabrication to determine the effects of irradiation on the microstructure. Based on comparison of the microstructural characterization results for each irradiated sample, some general conclusions can be drawn about how the microstructure evolves during irradiation: there is growth during irradiation of the fuel/matrix interaction (FMI) layer created during fabrication; Si diffuses from the FMI layer to deeper depths in the U-7Mo particles as the irradiation conditions are made more aggressive; lowering of the Si content in the FMI layer results in an increase in the size of the fission gas bubbles; as the FMI layer grows during irradiation, more Si diffuses from the matrix to the FMI layer/matrix interface; and interlinking of fission gas bubbles in the fuel plate microstructure that may indicate breakaway swelling is not observed.

  7. Milling Behavior of Matrix Graphite Powders with Different Binder Materials in HTGR Fuel Element Fabrication: I. Variation in Particle Size Distribution

    International Nuclear Information System (INIS)

    Lee, Young Woo; Cho, Moon Sung

    2011-01-01

    The fuel element for HTGR is manufactured by mixing coated fuel particles with matrix graphite powder and forming into either pebble type or cylindrical type compacts depending on their use in different HTGR cores. The coated fuel particle, the so-called TRISO particle, consists of 500-μm spherical UO 2 particles coated with the low density buffer Pyrolytic Carbon (PyC) layer, the inner and outer high density PyC layer and SiC layer sandwiched between the two inner and outer PyC layers. The coated TRISO particles are mixed with a matrix graphite powder properly prepared and pressed into a spherical shape or a cylindrical compact finally heat-treated at about 1900 .deg. C. These fuel elements can have different sizes and forms of compact. The basic steps for manufacturing a fuel element include preparation of graphite matrix powder, overcoating the fuel particles, mixing the fuel particles with a matrix powder, carbonizing green compact, and the final high-temperature heat treatment of the carbonized fuel compact. In order to develop a fuel compact fabrication technology, it is important to develop a technology to prepare the matrix graphite powder (MGP) with proper characteristics, which has a strong influence on further steps and the material properties of fuel element. In this work, the milling behavior of matrix graphite powder mixture with different binder materials and their contents was investigated by analyzing the change in particle size distribution with different milling time

  8. Advanced LMFBR fuel cladding susceptability to stress corrosion due to reprocessing impurities

    International Nuclear Information System (INIS)

    Henslee, S.P.

    1987-03-01

    The potential degradation of LMFBR fuel cladding alloys by chlorides, when used in metallic fuel systems, was evaluated. The alloys tested were D-9 and HT-9 stainless steels, austenitic and ferritic alloys respectively. These two alloys were tested in parallel with and their performance compared to the austenitic stainless steel Type 316. All alloys were tested for 7400 hours in a stress rupture environment with chloride exposure at either 550/degree/C 650/degree/C. None of the alloys tested were found to exhibit any degradation in time-to-rupture by the presence of chlorides under the conditions imposed during testing. 8 refs., 4 figs., 2 tabs

  9. Synthesize and characterization of a novel anticorrosive cobalt ferrite nanoparticles dispersed in silica matrix (CoFe2O4-SiO2) to improve the corrosion protection performance of epoxy coating

    International Nuclear Information System (INIS)

    Gharagozlou, M.; Ramezanzadeh, B.; Baradaran, Z.

    2016-01-01

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

  10. A study on the generation of radioactive corrosion product at PWR for extended fuel cycle

    International Nuclear Information System (INIS)

    Min Chul Song; Kun Jai Lee

    2001-01-01

    Current nuclear power plant operating practice is to extend the time between refueling from a 12 month operating cycle to an 18-24 month period. This current to longer fuel cycles has complicated the dilemma of finding optimum pH range for the primary coolant chemistry. The International Commission on Radiological Protection (ICRP) in ICRP publication No. 60 recommends optimization of operator radiation exposure (ORE) in nuclear power plants. CRUD formed in the plants is the major source of ORE and its transport mechanism is not understood. To analyze the generation of CRUD at the extended fuel cycle, the COTRAN code, which was developed at the Korea Advanced Institute of Science and Technology (KAIST), was used. It predicts that the activity of CRUD decreases as the pH of the coolant increases. For the same period of different fuel cycles, as the operating fuel cycle duration is increased, the generation of the CRUD increases. In this paper, enriched boric acid (40% enriched 10 B concentration) for reactivity control is adopted as the required chemical shim rather than natural boric acid. The effect of the enriched boric acid (EBA) is that the neutron absorption capability of the chemical shim is maintained while decreasing the required boron and lithium concentration in the reactor coolant system. By employing enriched boric acid, the amounts of CRUD generated are reduced, because the high pH-operating period is extended. From the waste generation point of view, more filters or ion exchangers to remove CRUD are required and the amounts of waste are increased at the extended fuel cycle. (author)

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

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

    DEFF Research Database (Denmark)

    Montgomery, Melanie; Larsen, OH

    2002-01-01

    undertaken where coal has been co-fired with 10% straw and 20% straw (% energy basis) for up to approx. 3000 hours. Two types of exposure were undertaken to investigate corrosion: a) the exposure of metal rings on water/air cooled probes, and b) the exposure of a range of materials built into the existing...... and potassium sulphate. These components give rise to varying degrees of accelerated corrosion. This paper concerns co-firing of straw with coal to reduce the corrosion rate from straw to an acceptable level. A field investigation at Midtkraft Studstrup suspension-fired power plant in Denmark has been...... for 100% straw-firing. The corrosion products and course of corrosion for the various steel types were investigated using light optical and scanning electron microscopy. Catastrophic corrosion due to potassium chloride was not observed. Instead a more modest corrosion rate due to potassium sulphate rich...

  13. Study on thermal conductivity of HTR spherical fuel element matrix graphite

    International Nuclear Information System (INIS)

    Zhang Kaihong; Liu Xiaoxue; Zhao Hongsheng; Li Ziqiang; Tang Chunhe

    2014-01-01

    Taking the spherical fuel element matrix graphite ball samples as an example, this paper introduced the principle and method of laser thermal conductivity meter, as well as the specific heat capacity, and analyzed the effects of different test methods and sampling methods on the thermal conductivities at 1000 ℃ of graphite material. The experimental results show that the thermal conductivities of graphite materials tested by synchronous thermal analyzer combining with laser thermal conductivity meter were different from that directly by laser thermal conductivity meter, the former was more reliable and accurate than the later; When sampling from different positions, central samples had higher thermal conductivities than edging samples, which was related to the material density and porosity at the different locations; the thermal conductivities had obvious distinction between samples from different directions, which was because the layer structure of polycrystalline graphite preferred orientation under pressure, generally speaking, the thermal conductivities perpendicular to the molding direction were higher than that parallel to the molding direction. Besides this, the test results show that the thermal conductivities of all the graphite material samples were greater than 30 W/(m (K), achieving the thermal performance index of high temperature gas cooled reactor. (authors)

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  15. Post-irradiation examinations of inert matrix nitride fuel irradiated in JMTR (01F-51A capsule)

    International Nuclear Information System (INIS)

    Iwai, Takashi; Nakajima, Kunihisa; Kikuchi, Hironobu; Honda, Junichi; Hatakeyama, Yuichi; Ono, Katsuto; Matsui, Hiroki; Arai, Yasuo

    2007-03-01

    A plutonium nitride fuel pin containing inert matrix such as ZrN and TiN was encapsulated in 01F-51A and irradiated in JMTR. Minor actinides are surrogated by plutonium. Average linear powers and burnups were 408W/cm, 30000MWd/t(Zr+Pu) [132000MWd/t-Pu] for (Zr,Pu)N and 355W/cm, 38000MWd/t(Ti+Pu) [153000MWd/t-Pu] for (TiN,PuN). The irradiated capsule was transported to Reactor Fuel Examination Facility and subjected to non-destructive and destructive post irradiation examinations. Any failure was not observed in the irradiated fuel pin. Very low fission gas release rate of about 1.6% was measured. The inner surface of cladding tube did not show any signs of chemical interaction with fuel pellet. (author)

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  17. 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.

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

    International Nuclear Information System (INIS)

    Yang, Meijun; Zhang, Dongming

    2014-01-01

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

  19. Disintegration of graphite matrix from the simulative high temperature gas-cooled reactor fuel element by electrochemical method

    International Nuclear Information System (INIS)

    Tian Lifang; Wen Mingfen; Li Linyan; Chen Jing

    2009-01-01

    Electrochemical method with salt as electrolyte has been studied to disintegrate the graphite matrix from the simulative high temperature gas-cooled reactor fuel elements. Ammonium nitrate was experimentally chosen as the appropriate electrolyte. The volume average diameter of disintegrated graphite fragments is about 100 μm and the maximal value is less than 900 μm. After disintegration, the weight of graphite is found to increase by about 20% without the release of a large amount of CO 2 probably owing to the partial oxidation to graphite in electrochemical process. The present work indicates that the improved electrochemical method has the potential to reduce the secondary nuclear waste and is a promising option to disintegrate graphite matrix from high temperature gas-cooled reactor spent fuel elements in the head-end of reprocessing.

  20. Performance of HVOF carbide coatings under erosion/corrosion

    International Nuclear Information System (INIS)

    Simard, S.; Arsenault, B.; Legoux, J.G.; Hawthorne, H.M.

    1999-01-01

    Cermet based materials are known to have an excellent performance under several wear conditions. High velocity oxy-fuel (HVOF) technology allows the deposition of such hard materials in the form of protective coatings onto different surfaces. Under slurry erosion, the performance of the coatings is influenced by the occurrence of corrosion reactions on the metallic matrix. Indeed, wet conditions promote the dissolution of metallic binder resulting in a potential synergic effect between the corrosion and wear mechanisms. The composition of the metallic matrix plays a key role on the stability of the coatings and their degradation rate. In this work, four coatings based on tungsten carbide embedded in different metallic binders were evaluated with regard to corrosion and wear. (author)

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

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  2. 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.

  3. Corrosion and corrosion control

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  4. Dictionary corrosion and corrosion control

    International Nuclear Information System (INIS)

    1985-01-01

    This dictionary has 13000 entries in both languages. Keywords and extensive accompanying information simplify the choice of word for the user. The following topics are covered: Theoretical principles of corrosion; Corrosion of the metals and alloys most frequently used in engineering. Types of corrosion - (chemical-, electro-chemical, biological corrosion); forms of corrosion (superficial, pitting, selective, intercrystalline and stress corrosion; vibrational corrosion cracking); erosion and cavitation. Methods of corrosion control (material selection, temporary corrosion protection media, paint and plastics coatings, electro-chemical coatings, corrosion prevention by treatment of the corrosive media); Corrosion testing methods. (orig./HP) [de

  5. Electrochemical and micro-gravimetric corrosion studies on spent fuel provide relevant source term data for a repository performance assessment

    International Nuclear Information System (INIS)

    Wegen, Detlef H.; Bottomley, Paul D. W.; Glatz, Jean-Paul

    2004-01-01

    Various electrochemical methods (corrosion potential monitoring, AC impedance analysis and electrochemical noise monitoring) were used in the investigation of UO 2 samples: natural and doped with two different levels of 238 Pu (0.1 and 10 wt%) simulating the increasing α-intensities seen with time in the repository. The results were compared and were able to show the intense, but also the very local nature of the radiolysis and to demonstrate that corrosion rates were proportional to α-radiolysis and hence the 238 Pu content; the corrosion rates were in accordance with earlier work at ITU. By contrast it was seen that the redox potentials only gave information as to the bulk solution that did not reflect the true conditions at the electrode interface that were driving the corrosion processes of UO 2 dissolution in groundwaters. The study shows how electrochemical techniques can provide vital information on the corrosion mechanism at the UO 2 /solution interface

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

    International Nuclear Information System (INIS)

    Herrera, V.; Zamora R, L.

    1997-01-01

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

  7. 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; Etude des mecanismes et des cinetiques de corrosion aqueuse de l'alliage d'aluminium AlFeNi utilise comme gainage du combustible nucleaire de reacteurs experimentaux

    Energy Technology Data Exchange (ETDEWEB)

    Wintergerst, M.

    2009-05-15

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-07-01

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

  9. High-temperature deformation and processing maps of Zr-4 metal matrix with dispersed coated surrogate nuclear fuel particles

    Science.gov (United States)

    Chen, Jing; Liu, Huiqun; Zhang, Ruiqian; Li, Gang; Yi, Danqing; Lin, Gaoyong; Guo, Zhen; Liu, Shaoqiang

    2018-06-01

    High-temperature compression deformation of a Zr-4 metal matrix with dispersed coated surrogate nuclear fuel particles was investigated at 750 °C-950 °C with a strain rate of 0.01-1.0 s-1 and height reduction of 20%. Scanning electron microscopy was utilized to investigate the influence of the deformation conditions on the microstructure of the composite and damage to the coated surrogate fuel particles. The results indicated that the flow stress of the composite increased with increasing strain rate and decreasing temperature. The true stress-strain curves showed obvious serrated oscillation characteristics. There were stable deformation ranges at the initial deformation stage with low true strain at strain rate 0.01 s-1 for all measured temperatures. Additionally, the coating on the surface of the surrogate nuclear fuel particles was damaged when the Zr-4 matrix was deformed at conditions of high strain rate and low temperature. The deformation stability was obtained from the processing maps and microstructural characterization. The high-temperature deformation activation energy was 354.22, 407.68, and 433.81 kJ/mol at true strains of 0.02, 0.08, and 0.15, respectively. The optimum deformation parameters for the composite were 900-950 °C and 0.01 s-1. These results are expected to provide guidance for subsequent determination of possible hot working processes for this composite.

  10. Corrosion properties of modified PNC1520 austenitic stainless steel in supercritical water as a fuel cladding candidate material for supercritical water reactor

    International Nuclear Information System (INIS)

    Nakazono, Yoshihisa; Iwai, Takeo; Abe, Hiroaki

    2009-01-01

    The supercritical water-cooled reactor (SCWR) has been designed and investigated because of its high thermal efficiency and plant simplification. There are some advantages including the use of a single phase coolant with high enthalpy. Supercritical Water (SCW) has never been used in nuclear power applications. There are numerous potential problems, particularly with materials. As the operating temperature of SCWR will be between 553 K and 893 K with a pressure of 25 MPa, the selection of materials is difficult and important. The PNC1520 austenitic stainless steel has been developed by Japan Atomic Energy Agency (JAEA) as a nuclear fuel cladding material for a Na-cooled fast breeder reactor. Austenitic Fe-base steels were selected for possible use in supercritical water systems because of their corrosion resistance and radiation resistance. The PNC1520 austenitic stainless steel was selected for possible use in supercritical water systems. The corrosion data of PNC1520 in SCW is required but does not exist. The purpose of the present study is to research the corrosion properties for PNC1520 austenitic stainless steel in SCW. The SCW corrosion test was performed for the standard PNC1520 (1520S) and the Ti-additional type of PNC1520 (1520T) by using a SCW autoclave. The 1520S and 1520T are the first trial production materials of SCWR cladding candidate material in our group. Corrosion and compatibility tests on the austenitic 1520S and 1520T steels in supercritical water were performed at 673, 773 and 600degC with exposures up to 1000 h. We have evaluated the amount of weight gain, weight loss and weight of scale after the corrosion test in SCW for 1520S and 1520T austenitic steels. After 1000 h corrosion test performed, the weight gains of both austenitic stainless steels were less than 2 g/m 2 at 400degC and 500degC. But 1520T weight increases more and weight loss than 1520S at 600degC. The SEM observation result of the surface after 1000 h corrosion of an test

  11. Instant release fraction and matrix release of high burn-up UO{sub 2} spent nuclear fuel: Effect of high burn-up structure and leaching solution composition

    Energy Technology Data Exchange (ETDEWEB)

    Serrano-Purroy, D., E-mail: Daniel.serrano-purroy@ec.europa.eu [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, D-76125 Karlsruhe (Germany); Clarens, F.; Gonzalez-Robles, E. [CTM Centre Tecnologic, Avda. Bases de Manresa 1, 08240 Barcelona (Spain); Glatz, J.P.; Wegen, D.H. [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, D-76125 Karlsruhe (Germany); Pablo, J. de [CTM Centre Tecnologic, Avda. Bases de Manresa 1, 08240 Barcelona (Spain); Department of Chemical Engineering, Universitat Politecnica de Catalunya, Avda. Diagonal 647, 08028 Barcelona (Spain); Casas, I.; Gimenez, J. [Department of Chemical Engineering, Universitat Politecnica de Catalunya, Avda. Diagonal 647, 08028 Barcelona (Spain); Martinez-Esparza, A. [ENRESA, C/Emilio Vargas 7, 28043 Madrid (Spain)

    2012-08-15

    Two weak points in Performance Assessment (PA) exercises regarding the alteration of Spent Nuclear Fuel (SNF) are the contribution of the so-called Instant Release Fraction (IRF) and the effect of High Burn-Up Structure (HBS). This manuscript focuses on the effect of HBS in matrix (long term) and instant release of a Pressurised Water Reactor (PWR) SNF irradiated in a commercial reactor with a mean Burn-Up (BU) of 60 GWd/tU. In order to study the HBS contribution, two samples from different radial positions have been prepared. One from the centre of the SNF, labelled CORE, and one from the periphery, enriched with HBS and labelled OUT. Static leaching experiments have been carried out with two synthetic leaching solutions: bicarbonate (BIC) and Bentonitic Granitic Groundwater (BGW), and in all cases under oxidising conditions. IRF values have been calculated from the determined Fraction of Inventory in Aqueous Phase (FIAP). In all studied cases, some radionuclides (RN): Rb, Sr and Cs, have shown higher release rates than uranium, especially at the beginning of the experiment, and have been considered as IRF. Redox sensitive RN like Mo and Tc have been found to dissolve slightly faster than uranium and further studies might be needed to confirm if they can also be considered part of the IRF. Most of the remaining studied RN, mainly actinides and lanthanides, have been found to dissolve congruently with the uranium matrix. Finally, Zr, Ru and Rh presented lower release rates than the matrix. Higher matrix release has been determined for CORE than for OUT samples showing that the formation of HBS might have a protective effect against the oxidative corrosion of the SNF. On the contrary, no significant differences have been observed between the two studied leaching solutions (BIC and BGW). Two different IRF contributions have been determined. One corresponding to the fraction of inventory segregated in the external open grain boundaries, directly available to water and

  12. Kinetic modelling of bentonite - canister interaction. Implications for Cu, Fe and Pb corrosion in a repository for spent nuclear fuel

    International Nuclear Information System (INIS)

    Wersin, P.; Bruno, J.; Spahiu, K.

    1993-06-01

    The chemical corrosion of three potential canister materials, Fe, Cu, and Pb is reviewed in terms of their thermodynamic and kinetic behavior in a repository. Thermodynamic predictions which are compatible with sedimentological observations indicate that for all three metals, chemical corrosion is expected at any time in a repository. From the kinetic information obtained by experimental and archeological data, long-term corrosion rates are assessed. In the case of Fe, the selected data allow extrapolation to repository conditions with a tolerable degree of uncertainty except for the possible effect of local corrosion in the initial oxic phase, For the other two metals, the scarcity of consistent experimental and archeological data limits the feasibility of this approach. In view of this shortcoming, a kinetic, single-box model, based on the STEADYQL code, is presented for quantitative prediction of long-term canister-bentonite interaction. The model is applied to the corrosion of Cu under anoxic conditions and upper and lower limits of corrosion rates are derived. The possibilities of extending this single-box model to a multi-box, diffusion-extended version are discussed. Finally, further potentials of STEADYQL for future applications of near field modelling are highlighted. 32 refs

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  14. Burnup simulations of an inert matrix fuel using a two region, multigroup reactor physics model

    Energy Technology Data Exchange (ETDEWEB)

    Schneider, E. [Dept. of Mechanical Engineering, Univ. of Texas at Austin, 1 Univ. Place C2200, Austin, TX 78712 (United States); Deinert, M.; Bingham Cady, K. [Dept. of Theoretical and Applied Mechanics, Cornell Univ., Ithaca, NY 14853 (United States)

    2006-07-01

    Determining the time dependent concentration of isotopes in a nuclear reactor core is of fundamental importance to analysis of nuclear fuel cycles and the impact of spent fuels on long term storage facilities. We present a fast, conceptually simple tool for performing burnup calculations applicable to obtaining isotopic balances as a function of fuel burnup. The code (VBUDS: visualization, burnup, depletion and spectra) uses a two region, multigroup collision probability model to determine the energy dependent neutron flux and tracks the buildup and burnout of 24 actinides, as well as fission products. The model has been tested against benchmarked results for LWRs burning UOX and MOX, as well as MONTEBURNS simulations of zirconium oxide based IMF, all with strong fidelity. As an illustrative example, VBUDS burnup calculation results for an IMF fuel are presented in this paper. (authors)

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

    International Nuclear Information System (INIS)

    Baldi, St.; Porta, J.

    2000-08-01

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

  16. Nuclear fuels for material test reactors

    International Nuclear Information System (INIS)

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

    1982-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-03-15

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

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

    International Nuclear Information System (INIS)

    Zwicky, Hans-Urs; Low, Jeanett; Ekeroth, Ella

    2011-03-01

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

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  20. The stress-corrosion behaviour in water media containing chlorine of the brazing joint of grids for PWR fuel element

    International Nuclear Information System (INIS)

    Zhang Weijie; Li Wenqing.

    1985-01-01

    This paper details the testing results of the stress-corrosion behaviour in the 150 deg C water media containing chlorine for the brazing joints made from three alloy systems, which are Ni-Cr-Si, Ni-Cr-P and Ni-P, including 16 compositions. The test results indicate that, in the Ni-Cr-Si system, Ni-Cr-Si-Ge brazing joint is the best, to resist stress-corrosion, while Ni-Cr-Si-P-Ge-Pd and BNi5 brazing joints are better. In the Ni-Cr-P system, only the Ni-Cr-P-Mo-Zr brazing joint has an excellent resistance to stress-corrosion

  1. Transport dynamics of a high-power-density matrix-type hydrogen-oxygen fuel cell

    Science.gov (United States)

    Prokopius, P. R.; Hagedorn, N. H.

    1974-01-01

    Experimental transport dynamics tests were made on a space power fuel cell of current design. Various operating transients were introduced and transport-related response data were recorded with fluidic humidity sensing instruments. Also, sampled data techniques were developed for measuring the cathode-side electrolyte concentration during transient operation.

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

    International Nuclear Information System (INIS)

    Petel'guzov, I.A.

    2002-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-07-01

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

  4. Spectrophotometric determination of silicon in silumin matrix

    International Nuclear Information System (INIS)

    Samanta, Papu; Pandey, K.L.; Kumar, Pradeep; Bagchi, A.C.; Abdulla, K.K.

    2015-01-01

    In dispersion fuel, fissile material is dispersed in inert matrix. Aluminum-silicon-nickel (silumin) alloy is employed as inert matrix owing to its high thermal conductivity, high castability, high corrosion resistance. All these properties depend on the chemical composition and the structure of silumin. Silicon is stringent specification in silumin. A spectrophotometric method has been developed for the determination of silicon content in silumin matrix. Silumin matrix was fused with LiOH and subsequent dissolution in water along with few drops of conc. sulphuric acid. The molybodo-silicic formed by the addition of ammonium molybdate is reduced to molybdenum blue by ascorbic acid in the presence of antimony. The absorbance was measured at 810 nm. Aluminum and nickel were found to be non-interfering with the silicon determination. (author)

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

    Science.gov (United States)

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

    2013-09-01

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

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  7. Characterisation of the corrosion products of non-irradiated material test reactors fuel elements (MTR-FE)

    Energy Technology Data Exchange (ETDEWEB)

    Mazeina, L.; Curtius, H.; Fachinger, J. [Inst. for Safety Research and Reactor Technology, Research Centre Juelich (Germany)

    2003-07-01

    In a high concentrated Mg-rich brine a non-irradiated MTR-FE corroded. The formed corrosion products consists of an amorphous part and of hydrotalcites, which were identified as Mg-Al-hydrotalcites with chloride anions in the interlayer. (orig.)

  8. 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.

  9. Investigation of neutronic behavior in a CANDU reactor with different (Am, Th, {sup 235}U)O{sub 2} fuel matrixes

    Energy Technology Data Exchange (ETDEWEB)

    Gholamzadeh, Z. [Talca Univ. (Chile). Dept. of Physics; Feghhi, S.A.H. [Shahid Beheshti Univ., Tehran (Iran, Islamic Republic of). Dept. of Radiation Application

    2014-11-15

    Recently thorium-based fuel matrixes are taken into consideration for nuclear waste incineration because of thorium proliferation resistance feature moreover its breeding or convertor ability in both thermal and fast reactors. In this work, neutronic influences of adding Am to (Th-{sup 235}U)O{sub 2} on effective delayed neutron fraction, reactivity coefficients and burn up of a fed CANDU core has been studied using MCNPX 2.6.0 computational code. Different atom fractions of Am have been introduced in the fuel matrix to evaluate its effects on neutronic parameters of the modeled core. The computational data show that adding 2% atom fraction of Am to thorium-based fuel matrix won't noticeably change reactivity coefficients in comparison with the fuel matrix containing 1% atom fraction of Am. The use of 2% atom fraction of Am resulted in a higher delayed neutron fraction. According to the obtained data, 32.85 GWd burn up of the higher Americium-containing fuel matrix resulted in 55.2%, 26.5%, 41.9% and 2.14% depletion of {sup 241}Am, {sup 243}Am, {sup 235}U and {sup 232}Th respectively. 132.8 kg of {sup 233}U fissile element is produced after the burn up time and the nuclear core multiplication factor increases in rate of 2390 pcm. The less americium-containing fuel matrix resulted in higher depletion of {sup 241/243}Am, {sup 235}U and {sup 232}Th while the nuclear core effective multiplication factor increases in rate of 5630 pcm after the burn up time with 9.8 kg additional {sup 233}U production.

  10. Inert matrix fuels for incineration of plutonium and transmutation of americium

    International Nuclear Information System (INIS)

    Matzke, Hj.

    2000-01-01

    In conventional U-based nuclear fuels, both Pu and higher actinides (mainly Am, but also Np and Cm) are formed by neutron capture reactions and α- or β-decay. If a strategy of reprocessing is adopted as in some European nations and in Japan, the separated Pu can be recycled as (U, Pu)O 2 (or mixed-oxide-MOX) fuel. The high-level liquid waste of reprocessing is presently vitrified. However, the alternative of separating the minor actinides from the fission products (partitioning) and subsequent transmutation in existing reactors or in new dedicated actinide burners is widely studied as a possible means to reduce the radiotoxicity of the waste

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

    International Nuclear Information System (INIS)

    Gaulard, C.

    2012-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-07-01

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

  13. Oxidation and reduction behaviors of a prototypic MgO-PuO{sub 2-x} inert matrix fuel

    Energy Technology Data Exchange (ETDEWEB)

    Miwa, Shuhei, E-mail: miwa.shuhei@jaea.go.jp; Osaka, Masahiko

    2017-04-15

    Oxidation and reduction behaviors of prototypic MgO-based inert matrix fuels (IMFs) containing PuO{sub 2-x} were experimentally investigated by means of thermogravimetry. The oxidation and reduction kinetics of the MgO-PuO{sub 2-x} specimen were determined. The oxidation and reduction rates of the MgO-PuO{sub 2-x} were found to be low compared with those of PuO{sub 2-x}. It is note that the changes in O/Pu ratios of MgO-PuO{sub 2-x} from stoichiometry were smaller than those of PuO{sub 2-x} at high oxygen partial pressure.

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

    International Nuclear Information System (INIS)

    Gibert, C.

    1999-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-07-01

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

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

    Science.gov (United States)

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

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

  17. Corrosion inhibitors. Manufacture and technology

    International Nuclear Information System (INIS)

    Ranney, M.W.

    1976-01-01

    Detailed information is presented relating to corrosion inhibitors. Areas covered include: cooling water, boilers and water supply plants; oil well and refinery operations; fuel and lubricant additives for automotive use; hydraulic fluids and machine tool lubes; grease compositions; metal surface treatments and coatings; and general processes for corrosion inhibitors

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-08-01

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

  19. Modeling of high-density U-MO dispersion fuel plate performance

    International Nuclear Information System (INIS)

    Hayes, S.L.; Meyer, M.K.; Hofman, G.L.; Rest, J.; Snelgrove, J.L.

    2002-01-01

    Results from postirradiation examinations (PIE) of highly loaded U-Mo/Al dispersion fuel plates over the past several years have shown that the interaction between the metallic fuel particles and the matrix aluminum can be extensive, reducing the volume of the high-conductivity matrix phase and producing a significant volume of low-conductivity reaction-product phase. This phenomenon results in a significant decrease in fuel meat thermal conductivity during irradiation. PIE has further shown that the fuel-matrix interaction rate is a sensitive function of irradiation temperature. The interplay between fuel temperature and fuel-matrix interaction makes the development of a simple empirical correlation between the two difficult. For this reason a comprehensive thermal model has been developed to calculate temperatures throughout the fuel plate over its lifetime, taking into account the changing volume fractions of fuel, matrix and reaction-product phases within the fuel meat owing to fuel-matrix interaction; this thermal model has been incorporated into the dispersion fuel performance code designated PLATE. Other phenomena important to fuel thermal performance that are also treated in PLATE include: gas generation and swelling in the fuel and reaction-product phases, incorporation of matrix aluminum into solid solution with the unreacted metallic fuel particles, matrix extrusion resulting from fuel swelling, and cladding corrosion. The phenomena modeled also make possible a prediction of fuel plate swelling. This paper presents a description of the models and empirical correlations employed within PLATE as well as validation of code predictions against fuel performance data for U-Mo experimental fuel plates from the RERTR-3 irradiation test. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-06-01

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

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

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  2. 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

  3. 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

  4. 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)

  5. Corrosion engineering

    Energy Technology Data Exchange (ETDEWEB)

    Fontana, M.G.

    1986-01-01

    This book emphasizes the engineering approach to handling corrosion. It presents corrosion data by corrosives or environments rather than by materials. It discusses the corrosion engineering of noble metals, ''exotic'' metals, non-metallics, coatings, mechanical properties, and corrosion testing, as well as modern concepts. New sections have been added on fracture mechanics, laser alloying, nuclear waste isolation, solar energy, geothermal energy, and the Statue of Liberty. Special isocorrosion charts, developed by the author, are introduced as a quick way to look at candidates for a particular corrosive.

  6. Stress-corrosion cracking properties of candidate fuel cladding alloys for the Canadian SCWR: a summary of literature data and recent test results

    Energy Technology Data Exchange (ETDEWEB)

    Zheng, W.; Zeng, Y., E-mail: Wenyue@NRcan.gc.ca [CanmetMATERIALS, Hamilton, ON (Canada); Luo, J. [Univ. of Alberta, Edmonton, AB (Canada); Novotny, R. [JRC-European Commission, Patten (Netherlands); Li, J.; Amirkhiz, B.S., E-mail: Jian.li@nrcan.gc.ca [CanmetMATERIALS, Hamilton, ON (Canada); Guzonas, D. [Atomic Energy of Canada Limited, Chalk River, ON (Canada); Matchim, M.; Collier, J.; Yang, L., E-mail: lin.yang@nrcan.gc.ca [CanmetMATERIALS, Hamilton, ON (Canada)

    2014-07-01

    Cracking of fuel claddings is a serious concern when selecting candidate alloys for the development of a next-generation reactor. Whether the cracking is due to an environment-metal interaction such as stress-corrosion, or a pure metallurgical process such as localized plastic deformation along grain boundaries, the final impact is the same: cracking of the cladding can lead to fuel failure. In the course of a review of potential candidate alloys in preparation for further assessment under conditions relevant to the Canadian SCWR concept, relevant cracking studies reported for five short-listed alloys (namely 310S, 347H, 800H, 625 and 214) in the open literature were examined, and the key findings are provided in this paper. Discussions are also made of the recent SCC data from capsule tests and slow-strain rate tests (SSRT) in supercritical water. The data suggest that there is a threshold strain level below which SCC is not developed during SSRT tests. The practical implication of this finding is also discussed. (author)

  7. To the corrosion of austenitic steels in sodium loops

    International Nuclear Information System (INIS)

    Schad, M.

    1978-03-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-03-15

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

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

    International Nuclear Information System (INIS)

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

    2017-01-01

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

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

    Directory of Open Access Journals (Sweden)

    Geoff Recktenwald

    2013-04-01

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

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

    Directory of Open Access Journals (Sweden)

    Gholamzadeh Zohreh

    2014-12-01

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

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

    International Nuclear Information System (INIS)

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

    1999-01-01

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

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  14. Microstructural Characterization of a Mg Matrix U-Mo Dispersion Fuel Plate Irradiated in the Advanced Test Reactor to High Fission Density: SEM Results

    Science.gov (United States)

    Keiser, Dennis D.; Jue, Jan-Fong; Miller, Brandon D.; Gan, Jian; Robinson, Adam B.; Medvedev, Pavel G.; Madden, James W.; Moore, Glenn A.

    2016-06-01

    Low-enriched (U-235 RERTR-8 experiment at high temperature, high fission rate, and high power, up to high fission density. This paper describes the results of the scanning electron microscopy (SEM) analysis of an irradiated fuel plate using polished samples and those produced with a focused ion beam. A follow-up paper will discuss the results of transmission electron microscopy (TEM) analysis. Using SEM, it was observed that even at very aggressive irradiation conditions, negligible chemical interaction occurred between the irradiated U-7Mo fuel particles and Mg matrix; no interconnection of fission gas bubbles from fuel particle to fuel particle was observed; the interconnected fission gas bubbles that were observed in the irradiated U-7Mo particles resulted in some transport of solid fission products to the U-7Mo/Mg interface; the presence of microstructural pathways in some U-9.1 Mo particles that could allow for transport of fission gases did not result in the apparent presence of large porosity at the U-7Mo/Mg interface; and, the Mg-Al interaction layers that were present at the Mg matrix/Al 6061 cladding interface exhibited good radiation stability, i.e. no large pores.

  15. Operation Strategies Based on Carbon Corrosion and Lifetime Investigations for High Temperature Polymer Electrolyte Membrane Fuel Cell Stacks

    DEFF Research Database (Denmark)

    Kannan, A.; Kaczerowski, J.; Kabza, A.

    2018-01-01

    This paper is aimed to develop operation strategies or high temperature polymer electrolyte fuel cells (HT-PEMFCs) stacks in order to enhance the endurance by mitigating carbon oxidation reaction. The testing protocols are carefully designed to suit the operating cycle for the realistic application...

  16. Corrosion of non-irradiated UAl{sub x}-Al fuel in the presence of clay pore solution. A quantitative XRD secondary phase analysis applying the DDM method

    Energy Technology Data Exchange (ETDEWEB)

    Neumann, Andreas [Halle-Wittenberg Univ. (Germany). Dept. of Mineralogy and Geochemistry; RWTH Aachen Univ. (Germany). Inst. of Crystallography; Klinkenberg, Martina; Curtius, Hildegard [Forschungszentrum Juelich GmbH (Germany). Inst. of Energy and Climate Research, IEK-6 Nuclear Waste Management

    2017-04-01

    Corrosion experiments with non-irradiated metallic UAl{sub x}-Al research reactor fuel elements were carried out in autoclaves to identify and quantify the corrosion products. Such compounds, considering the long-term safety assessment of final repositories, can interact with the released inventory and this constitutes a sink for radionuclide migration in formation waters. Therefore, the metallic fuel sample was subjected to clay pore solution to investigate its process of disintegration by analyzing the resulting products and the remnants, i.e. the secondary phases. Due to the fast corrosion rate a full sample disintegration was observed within the experimental period of 1 year at 90 C. The obtained solids were subdivided into different grain size fractions and prepared for analysis. The elemental analysis of the suspension showed that, uranium and aluminum are concentrated in the solids, whereas iron was mainly dissolved. Non-ambient X-ray diffraction (XRD) combined with the derivative difference minimization (DDM) method was applied for the qualitative and quantitative phase analysis (QPA) of the secondary phases. Gypsum and hemihydrate (bassanite), residues of non-corroded nuclear fuel, hematite, and goethite were identified. The quantitative phase analysis showed that goethite is the major crystalline phase. The amorphous content exceeded 80 wt% and hosted the uranium. All other compounds were present to a minor content. The obtained results by XRD were well supported by complementary scanning electron microscopy (SEM) and energy dispersive X-ray spectroscopy (EDS) analysis.

  17. Plastics for corrosion inhibition

    CERN Document Server

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

    2005-01-01

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

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

    Directory of Open Access Journals (Sweden)

    Olufunmilayo O. Joseph

    2016-06-01

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

  19. 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)

  20. 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)

  1. Investigation on microstructural, anti-corrosion and mechanical properties of doped Zn–Al–SnO2 metal matrix composite coating on mild steel

    International Nuclear Information System (INIS)

    Fayomi, O.S.I.; Popoola, A.P.I.; Aigbodion, V.S.

    2015-01-01

    Highlights: • Properties of nanocomposite Zn–Al coating containing SnO 2 nanoparticles. • The morphology and structure of the coating were analysed. • The anticorrosion activities of the coating prepared. • The mechanical properties were found to improve with the amount of the SnO 2 embedded. - Abstract: In this study, the microstructural, mechanical and anti-corrosion properties of nanocomposite Zn–Al coating containing SnO 2 nanoparticles prepared from sulphates electrolyte by electrodeposition on mild steel substrate was investigated. The morphologies of the coating were analysed using SEM/EDS, AFM Raman and X-ray diffraction. The anticorrosion behaviour of the coating prepared with different concentrations of SnO 2 (7 and 13 g/L) and potential of (0.3 and 0.5 V) was examined in 3.65% NaCl solution by using linear polarization techniques. The wear and hardness properties of the coatings were performed under accelerated reciprocating dry sliding wear tests and diamond micro-hardness tester respectively. The results obtained showed that the incorporation of SnO 2 in the plating bath brings an increase in corrosion resistance and mechanical properties of Zn–Al–SnO 2 composite coatings. The SEM images showed a homogeneous grain structure and finer morphology of the coatings. The hardness values was found to improve with the amount of the SnO 2 embedded into the Zn–Al metal deposit and effective deposition parameters

  2. Aspects of high temperature corrosion of boiler tubes

    Energy Technology Data Exchange (ETDEWEB)

    Spiegel, M.; Bendick, W. [Salzgitter-Mannesmann-Forschung GmbH, Duisburg (Germany)

    2008-07-01

    The development of new boiler steels for power generation has to consider significant creep strength as well as oxidation and corrosion resistance. High temperature corrosion of boiler materials concerns steam oxidation as well as fireside corrosion of parts, in contact with the flue gas. It will be shown that depending on the quality of the fuel, especially chlorine and sulphur are responsible for most of the fireside corrosion problems. Corrosion mechanisms will be presented for flue gas induced corrosion (HCl) and deposit induced corrosion (chlorides and sulfates). Especially for the 700 C technology, deposit induced corrosion issues have to be considered and the mechanisms of corrosion by molten sulfates 'Hot Corrosion' will be explained. Finally, an overview will be given on the selection of suitable materials in order to minimise corrosion relates failures. (orig.)

  3. Limits to the use of highly compacted bentonite as a deterrent for microbiologically influenced corrosion in a nuclear fuel waste repository

    Science.gov (United States)

    Stroes-Gascoyne, Simcha; Hamon, Connie J.; Maak, Peter

    Recent studies have suggested that microbial activity in highly compacted bentonite (⩾1600 kg/m 3) is severely suppressed. Therefore, it appears that the dry density of emplaced bentonite barriers in a geological repository for nuclear waste may be tailored such that a microbiologically unfavorable environment can be created adjacent to used fuel containers. This would ensure that microbiologically influenced corrosion is a negligible contributor to the overall corrosion process. However, this premise is valid only as long as the emplaced bentonite maintains a uniform high dry density (⩾1600 kg/m 3) because it has been shown that high dry density only suppresses microbial activity but not necessarily eliminates the viable microbial population in bentonite. In a repository, a reduction in the dry density of highly compacted bentonite may occur at a number of interface locations, such as placement gaps, contact regions with materials of different densities and contact points with water-carrying fractures in the rock. Experiments were carried out in our laboratory to examine the effects of a reduction in dry density (from 1600 kg/m 3 to about 1000 kg/m 3) on the recovery of microbial culturability in compacted bentonite. Results showed that upon expansion of compacted bentonite into a void, the resulting reduction in dry density stimulated or restored culturability of indigenous microbes. In a repository this would increase the possibility of in situ activity, which might be detrimental for the longevity of waste containers. Reductions in dry density, therefore, should be minimized or eliminated by adequate design and placement methods of compacted bentonite. Materials compliance models can be used to determine the required as-placed dry densities of bentonite buffer and gap fillings to achieve specific targets for long-term equilibrium dry densities for various container placement room designs. Locations where flowing fractures could be in contact with highly

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

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

    Science.gov (United States)

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

    2018-01-01

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

  6. Investigation on microstructural, anti-corrosion and mechanical properties of doped Zn–Al–SnO{sub 2} metal matrix composite coating on mild steel

    Energy Technology Data Exchange (ETDEWEB)

    Fayomi, O.S.I., E-mail: ojosundayfayomi3@gmail.com [Department of Chemical, Metallurgical and Materials Engineering, Tshwane University of Technology, P.M.B. X680, Pretoria (South Africa); Department of Mechanical Engineering, Covenant University, P.M.B 1023, Ota, Ogun State (Nigeria); Popoola, A.P.I. [Department of Chemical, Metallurgical and Materials Engineering, Tshwane University of Technology, P.M.B. X680, Pretoria (South Africa); Aigbodion, V.S. [Department of Metallurgical and Materials Engineering, University of Nigeria, Nsukka (Nigeria)

    2015-02-25

    Highlights: • Properties of nanocomposite Zn–Al coating containing SnO{sub 2} nanoparticles. • The morphology and structure of the coating were analysed. • The anticorrosion activities of the coating prepared. • The mechanical properties were found to improve with the amount of the SnO{sub 2} embedded. - Abstract: In this study, the microstructural, mechanical and anti-corrosion properties of nanocomposite Zn–Al coating containing SnO{sub 2} nanoparticles prepared from sulphates electrolyte by electrodeposition on mild steel substrate was investigated. The morphologies of the coating were analysed using SEM/EDS, AFM Raman and X-ray diffraction. The anticorrosion behaviour of the coating prepared with different concentrations of SnO{sub 2} (7 and 13 g/L) and potential of (0.3 and 0.5 V) was examined in 3.65% NaCl solution by using linear polarization techniques. The wear and hardness properties of the coatings were performed under accelerated reciprocating dry sliding wear tests and diamond micro-hardness tester respectively. The results obtained showed that the incorporation of SnO{sub 2} in the plating bath brings an increase in corrosion resistance and mechanical properties of Zn–Al–SnO{sub 2} composite coatings. The SEM images showed a homogeneous grain structure and finer morphology of the coatings. The hardness values was found to improve with the amount of the SnO{sub 2} embedded into the Zn–Al metal deposit and effective deposition parameters.

  7. Long-term corrosion studies

    International Nuclear Information System (INIS)

    Gdowski, G.

    1998-01-01

    The scope of this activity is to assess the long-term corrosion properties of metallic materials under consideration for fabricating waste package containers. Three classes of metals are to be assessed: corrosion resistant, intermediate corrosion resistant, and corrosion allowance. Corrosion properties to be evaluated are general, pitting and crevice corrosion, stress-corrosion cracking, and galvanic corrosion. The performance of these materials will be investigated under conditions that are considered relevant to the potential emplacement site. Testing in four aqueous solutions, and vapor phases above them, and at two temperatures are planned for this activity. (The environmental conditions, test metals, and matrix are described in detail in Section 3.0.) The purpose and objective of this activity is to obtain the kinetic and mechanistic information on degradation of metallic alloys currently being considered for waste package containers. This information will be used to provide assistance to (1) waste package design (metal barrier selection) (E-20-90 to E-20-92), (2) waste package performance assessment activities (SIP-PA-2), (3) model development (E-20-75 to E-20-89). and (4) repository license application

  8. Corrosion in power engineering

    International Nuclear Information System (INIS)

    1988-03-01

    The proceedings contain the full texts of 25 papers of which 10 fall under the INIS Subject Scope. They concern the problems of corrosion in WWER type nuclear power plants. The topics include structural materials and equipment of the primary and the secondary circuits of nuclear power plants, components used in disposal of spent nuclear fuel, sodium valves for fast reactors and basic study of the properties of materials used in nuclear power. (Z.M.). 12 figs., 6 tabs., 46 refs

  9. Ultrasonic Guided Wave Technology for Non-Invasive Assessment of Corrosion-Induced Damage in Piping for Pollution Prevention in DOD Fuel Storage Facilities

    Science.gov (United States)

    2011-08-01

    using a mixture of rock salt and tap water. A combination of lacquer and duct tape was used to mask the desired corrosion areas and to prevent excessive...Corrosion Measurements," NDT efT E International, Vol. 37, No.3, 2004, pp. 167-180. Demma, A, P. Cawley, M. Lowe and B. Pavlakovic, "The Effect of

  10. Corrosion Engineering.

    Science.gov (United States)

    White, Charles V.

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-07-01

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

  12. 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.

  13. Corrosion engineering in nuclear power industry

    International Nuclear Information System (INIS)

    Prazak, M.; Tlamsa, J.; Jirousova, D.; Silber, K.

    1990-01-01

    Corrosion problems in nuclear power industry are discussed from the point of view of anticorrosion measures, whose aim is not only increasing the lifetime of the equipment but, first of all, securing ecological safety. A brief description is given of causes of corrosion damage that occurred at Czechoslovak nuclear power plants and which could have been prevented. These involve the corrosion of large-volume radioactive waste tanks made of the CSN 17247 steel and of waste piping of an ion exchange station made of the same material, a crack in a steam generator collector, contamination of primary circuit water with iron, and corrosion of CrNi corrosion-resistant steel in a spent fuel store. It is concluded that if a sufficient insight into the corrosion relationships exists and a reasonable volume of data is available concerning the corrosion state during the nuclear facility performance, the required safety can be achieved without adopting extremely costly anticorrosion measures. (Z.M.)

  14. Electrolyte loss mechanism of molten carbonate fuel cells. 2.; Application to the cell with matrix electrolyte layer; Yoyu tansan`engata nenryo denchi ni okeru denkaishitsu loss kiko ni tsuite. 2.; Matrix gata denkaishitsuso wo yusuru denchi eno oyo

    Energy Technology Data Exchange (ETDEWEB)

    Sonai, A; Murata, K [Toshiba Research and Development Center, Kawasaki (Japan)

    1993-11-01

    A single cell of molten carbonate fuel cell using a matrix electrolyte layer fabricated by using the doctor blade process has been operated for several thousand hours, measured of electrolyte loss amount, and analyzed by using a new electrolyte loss mechanism. The result may be summarized as follows: according to a result of measuring the matrix layer pore distribution, the average pore size has increased little by little; pores with diameters greater than 2 {mu}m at which no electrolyte retention becomes possible remain at nearly constant ratio up to 1800 hours, but increased after 2500 hours; the pore capacity in ports with the largest electrolyte retaining diameter of 2 {mu}m or less showed slight decrease with time in the anode, and an initial decrease followed by flatness, and then a sharp decrease after 1800 hours in the matrix layer; the electrolyte loss measurement values have remained nearly constant for 25 hours to 1800 hours, but increased sharply thereafter; and the electrolyte loss in this single cell due to pore capacity decrease in pores as power generating parts with diameters smaller than 2 {mu}m was explained quantitatively by a new electrolyte loss mechanism. 11 refs., 6 figs.

  15. Corrosion and alteration of materials from the nuclear industry

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  16. Corrosion-Resistant Ti- xNb- xZr Alloys for Nitric Acid Applications in Spent Nuclear Fuel Reprocessing Plants

    Science.gov (United States)

    Manivasagam, Geetha; Anbarasan, V.; Kamachi Mudali, U.; Raj, Baldev

    2011-09-01

    This article reports the development, microstructure, and corrosion behavior of two new alloys such as Ti-4Nb-4Zr and Ti-2Nb-2Zr in boiling nitric acid environment. The corrosion test was carried out in the liquid, vapor, and condensate phases of 11.5 M nitric acid, and the potentiodynamic anodic polarization studies were performed at room temperature for both alloys. The samples subjected to three-phase corrosion testing were characterized using scanning electron microscopy (SEM) and energy-dispersive X-ray microanalysis (EDAX). As Ti-2Nb-2Zr alloy exhibited inferior corrosion behavior in comparison to Ti-4Nb-4Zr in all three phases, weldability and heat treatment studies were carried out only on Ti-4Nb-4Zr alloy. The weldability of the new alloy was evaluated using tungsten inert gas (TIG) welding processes, and the welded specimen was thereafter tested for its corrosion behavior in all three phases. The results of the present investigation revealed that the newly developed near alpha Ti-4Nb-4Zr alloy possessed superior corrosion resistance in all three phases and excellent weldability compared to conventional alloys used for nitric acid application in spent nuclear reprocessing plants. Further, the corrosion resistance of the beta heat-treated Ti-4Nb-4Zr alloy was superior when compared to the sample heat treated in the alpha + beta phase.

  17. 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

  18. Analysis of crack initiation in the vicinity of an interface in brittle materials. Applications to ceramic matrix composites and nuclear fuels

    International Nuclear Information System (INIS)

    Poitou, B.

    2007-11-01

    In this study, criterions are proposed to describe crack initiation in the vicinity of an interface in brittle bi-materials. The purpose is to provide a guide for the elaboration of ceramic multi-layer structures being able to develop damage tolerance by promoting crack deflection along interfaces. Several cracking mechanisms are analyzed, like the competition between the deflection of a primary crack along the interface or its penetration in the second layer. This work is first completed in a general case and is then used to describe the crack deviation at the interface in ceramic matrix composites and nuclear fuels. In this last part, experimental tests are carried out to determine the material fracture properties needed to the deflection criteria. An optimization of the fuel coating can be proposed in order to increase its toughness. (author)

  19. Fluorescence excitation-emission matrix (EEM) spectroscopy and cavity ring-down (CRD) absorption spectroscopy of oil-contaminated jet fuel using fiber-optic probes.

    Science.gov (United States)

    Omrani, Hengameh; Barnes, Jack A; Dudelzak, Alexander E; Loock, Hans-Peter; Waechter, Helen

    2012-06-21

    Excitation emission matrix (EEM) and cavity ring-down (CRD) spectral signatures have been used to detect and quantitatively assess contamination of jet fuels with aero-turbine lubricating oil. The EEM spectrometer has been fiber-coupled to permit in situ measurements of jet turbine oil contamination of jet fuel. Parallel Factor (PARAFAC) analysis as well as Principal Component Analysis and Regression (PCA/PCR) were used to quantify oil contamination in a range from the limit of detection (10 ppm) to 1000 ppm. Fiber-loop cavity ring-down spectroscopy using a pulsed 355 nm laser was used to quantify the oil contamination in the range of 400 ppm to 100,000 ppm. Both methods in combination therefore permit the detection of oil contamination with a linear dynamic range of about 10,000.

  20. 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

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

  1. On the possibility of reprocessing of fuel elements of dispersion type with copper matrix by pyrochemical methods

    International Nuclear Information System (INIS)

    Vasin, B.D.; Ivanov, V.A.; Shchetinskij, A.V.; Vavilov, S.K.; Savochkin, Yu.P.; Bychkov, A.V.; Kormilitsyn, M.V.

    2005-01-01

    A consideration is given to pyrochemical processes suitable for separation of uranium dioxide from structural materials when reprocessing cermet type fuel elements. The estimation of the possibility to apply liquid antimony and bismuth, potassium and copper chlorides melts is made. The specimens compacted of copper and uranium dioxide powders in a stainless steel can are used as simulators of fuel element sections. It is concluded that the dissolution of structural materials in molten salts at the stage of uranium dioxide concentration is the process of choice for reprocessing of dispersion type fuel elements [ru

  2. Development and implementation of computational geometric model for simulation of plate type fuel fabrication process with microspheres dispersed in metallic matrix

    International Nuclear Information System (INIS)

    Lage, Aldo M.F.; Reis, Sergio C.; Braga, Daniel M.; Santos, Armindo; Ferraz, Wilmar B.

    2005-01-01

    In this report it is presented the development of a geometric model to simulate the plate type fuel fabrication process with fuels microspheres dispersed in metallic matrix, as well as its software implementation. The developed geometric model encloses the steps of pellets pressing and sintering, as well as the plate rolling passes. The model permits the simulation of structures, where the values of the various variables of the fabrication processes can be studied and modified. The following variables were analyzed: microspheres diameters, density of the powder/microspheres mixing, microspheres density, fuel volume fraction, sintering densification, and rolling passes number. In the model implementation, which was codified in DELPHI programming language, systems of structured analysis techniques were utilized. The structures simulated were visualized utilizing the AutoCAD applicative, what permitted to obtain planes sections in diverse directions. The objective of this model is to enable the analysis of the simulated structures and supply information that can help in the improvement of the dispersion microspheres fuel plates fabrication process, now in development at CDTN (Centro de Desenvolvimento da Tecnologia Nuclear) in cooperation with the CTMSP (Centro Tecnologico da Marinha em Sao Paulo). (author)

  3. Corrosion control for low-cost reliability: Preceedings

    International Nuclear Information System (INIS)

    Anon.

    1993-01-01

    This book is Volume 6 of the preceedings from the 12th International Corrosion Congress. The electric power industry workshop dealt with water chemistry control; monitoring of chemical, electrochemical, and biological corrosion; corrosion product analyses; and nuclear and fossil-fuel power plants. All papers have been processed separately for inclusion on the data base

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

    Science.gov (United States)

    2011-10-30

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

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

    International Nuclear Information System (INIS)

    Olander, D.

    2009-01-01

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

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  7. Nuclear fuel cladding material

    International Nuclear Information System (INIS)

    Nakahigashi, Shigeo.

    1982-01-01

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

  8. Matrix effect studies in the GF-AAS determination of indium and antimony in PHWR and AHWR fuels

    International Nuclear Information System (INIS)

    Goyal, Neelam; Purohit, Paru J.; Kulkarni, Madhuri J.; Godbole, S.V.

    2009-01-01

    Studies on the atomization of indium and antimony in U, Th and U+Th matrices were carried out and the effect of matrix composition on the analyte absorbance was investigated. These studies have shown that the absorbance signal observed for In and Sb in U/Th/mixed matrix was enhanced as compared to that observed in matrix free solutions. Based on these investigations, analytical methods were developed for direct determination of In and Sb in the range 0.5- 25 ppm and 1.0 - 20 ppm respectively on the basis of 20 mg/mL U/ Th/ (3%U+Th) solution using GF-AAS. The performance of these methods was evaluated by analyzing synthetic samples for these matrices. (author)

  9. Investigation of the Thermal Stability of Nd(x)Sc(y)Zr(1-x-y)O(2-δ) Materials Proposed for Inert Matrix Fuel Applications.

    Science.gov (United States)

    Hayes, John R; Grosvenor, Andrew P; Saoudi, Mouna

    2016-02-01

    Inert matrix fuels (IMF) consist of transuranic elements (i.e., Pu, Am, Np, Cm) embedded in a neutron transparent (inert) matrix and can be used to "burn up" (transmute) these elements in current or Generation IV nuclear reactors. Yttria-stabilized zirconia has been extensively studied for IMF applications, but the low thermal conductivity of this material limits its usefulness. Other elements can be used to stabilize the cubic zirconia structure, and the thermal conductivity of the fuel can be increased through the use of a lighter stabilizing element. To this end, a series of Nd(x)Sc(y)Zr(1-x-y)O(2-δ) materials has been synthesized via a co-precipitation reaction and characterized by multiple techniques (Nd was used as a surrogate for Am). The long-range and local structures of these materials were studied using powder X-ray diffraction, scanning electron microscopy, and X-ray absorption spectroscopy. Additionally, the stability of these materials over a range of temperatures has been studied by annealing the materials at 1100 and 1400 °C. It was shown that the Nd(x)Sc(y)Zr(1-x-y)O(2-δ) materials maintained a single cubic phase upon annealing at high temperatures only when both Nd and Sc were present with y ≥ 0.10 and x + y > 0.15.

  10. Porous or roughened electrode with an assigned matrix for electrochemical cells with acid electrolytes, especially fuel cell batteries

    Energy Technology Data Exchange (ETDEWEB)

    Boehm, H; Fleischmann, R

    1975-10-16

    According to the invention an electrolyte matrix is to be used, which experiences a definite increase in volume by swelling, and fits on to the electrode in such a way that it is closely meshed with it. Matrix materials treated with sulphuric acid are claimed, such as 1) polystyrol, polyethelyne, polyvinyl alcohol, polyvinyl acetate, methyl cellulose or polyester, which are used alone or compounded with silica gel, aluminium oxide or sea sand, and 2) zeolite, silica, aluminium dioxide, titanium dioxide or zirconium dioxide using binding materials.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1979-12-01

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

  13. The corrosion properties of Zr-Cr-NM alloy metallic waste form for longterm disposal

    Energy Technology Data Exchange (ETDEWEB)

    Han, Seung Youb; Jang, Seon Ah; Eun, Hee Chul; Choi, Jung Hoon; Lee, Ki Rak; Park, Hwan Seo; Ahn, Do Hee [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2017-06-15

    KAERI is conducting research on spent cladding hulls and additive metals to generate a solidifcation host matrix for the noble metal fssion product waste in anode sludge from the electro-refning process to minimize the volume of waste that needs to be disposed of. In this study, alloy compositions Zr-17Cr, Zr-22Cr, and Zr-27Cr were prepared with or without eight noble metals representing fuel waste using induction melting. The microstructures of the resulting alloys were characterized and electrochemical corrosion tests were conducted to evaluate their corrosion characteristics. All the compositions had better corrosion characteristics than other Zr-based alloys that were evaluated for comparison. Analysis of the leach solution after the corrosion test of the Zr-22Cr-8NM specimen indicated that the noble metals were not leached during corrosion under 500 mV imposed voltage, which simulates a highly oxidizing disposal environment. The results of this study confrm that Zr-Cr based compositions will likely serve as chemically stable waste forms.

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

    Directory of Open Access Journals (Sweden)

    Hamilta de Oliveira Santos

    2005-06-01

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

  15. Future and benefits of corrosion research

    International Nuclear Information System (INIS)

    Staehle, Roger W.

    2002-01-01

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

  16. Corrosion in airframes

    OpenAIRE

    PETROVIC ZORAN C.

    2016-01-01

    The introductory chapter provides a brief reference to the issue of corrosion and corrosion damage to aircraft structures. Depending on the nature and dimensions of this non uniformity, three different categories of corrosion are defined: uniform, selective and localized corrosion. The following chapters present the forms of corrosion that can occur in three defined categories of corrosion. Conditions that cause certain types of corrosion in various corrosive environments are discussed. Examp...

  17. Topical problems of corrosion research for nuclear power purposes

    International Nuclear Information System (INIS)

    Eremias, B.

    1978-01-01

    Currently, research is focused on stress corrosion, intergranular corrosion, corrosion in water and steam, hydrogen-induced corrosion and corrosion in liquid sodium. The effort to limit stress corrosion resulted in the application of high nickel content austenitic steels. In these steels, the susceptibility to stress corrosion is mainly affected by previous heat treatment and the presence of chloride ions. Attention is also paid to medium and high-alloy chromium steels and susceptibility is studied to intergranular corrosion and stress corrosion. Of low-alloy steels the 21/4Cr-1Mo type steels stabilized with Nb or nonstabilized are studied with respect to decarburization kinetics and changes in mechanical properties in the presence of hydrogen. Of nonferrous metals zirconium alloys are studied used as cladding materials for fuel elements, mainly Zircaloy 2 and 4, with regard to their resistance to high-temperature oxidation, high-pressure steam action, etc. (J.F.)

  18. Fuel assembly

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  19. Evaluation of corrosion attack of chimney liners

    Directory of Open Access Journals (Sweden)

    Blahetová M.

    2016-06-01

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

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

    International Nuclear Information System (INIS)

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

    2003-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1969-07-01

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

  2. 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)

  3. Corrosion inhibition

    Energy Technology Data Exchange (ETDEWEB)

    Fisher, A O

    1965-12-29

    An acid corrosion-inhibiting composition consists essentially of a sugar, and an alkali metal salt selected from the group consisting of iodides and bromides. The weight ratio of the sugar to the alkali metal salt is between 2:1 and about 20,000:1. Also, a corrosion- inhibited phosphoric acid composition comprising at least about 20 wt% of phosphoric acid and between about 0.1 wt% and about 10 wt% of molasses, and between about 0.0005 wt% and about 1 wt% of potassium iodide. The weight ratio of molasses to iodide is greater than about 2:1. (11 claims)

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

    OpenAIRE

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

    2015-01-01

    The significant use of platinum for catalyzing the cathodic oxygen reduction reactions (ORRs) has hampered the widespread use of polymer electrolyte membrane fuel cells (PEMFCs). The construction of well-defined electrode architecture in nanoscale with enhanced utilization and catalytic performance of Pt might be a promising approach to address such barrier. Inspired by the highly efficient catalytic processes in enzymes with active centers embedded in charge transport pathways, here we demon...

  5. 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

  6. Investigations on radionuclide release and on the corrosion behaviour of spent fuels from research reactors under disposal conditions. Final report; Untersuchungen zur Radionuklidfreisetzung und zum Korrosionsverhalten von bestrahltem Kernbrennstoff aus Forschungsreaktoren unter Endlagerbeedingungen. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Bruecher, H.; Curtius, H.; Fachinger, J.; Kaiser, G.; Mazeina, L.; Nau, K.

    2003-12-01

    From the present report 'Untersuchungen zur Radionuklidfreisetzung und zum Korrosionsverhalten von bestrahltem Kernbrennstoff aus Forschungsreaktoren unter Endlagerbedingungen' with the code number FKZ 9108 carried out in the time periode 01.06.1998 till 30.11.2001 the following results can be withdrawn: U/Al-RR-fuel elements corroded slowly in granite water (Grimsel-West) at 90 C under anaerobic conditions. For a complete dissolution of the fuel element a time period of 10{sup 3} years is assumed according to present conservative results. In salt brines, especially in magnesium chloride rich brines the corrosion rate is high. Addition of GGG40 (basic material of the fuel element container with iron as main element) had an acceleration effect. A complete dissolution of the fuel is achieved within a couple of months. Under aerobic and under anaerobic conditions the bulk of released radionuclides were fixed by the corrosion products formed (secondary phases). The actinides were mobilised by variation of the ionic strength of the leaching solution. This process can be explained by phase conversion reactions within the secondary phases. Secondary phases formed by corrosion of a non-irradiated U/Al-RR-fuel element, were analysed and hydrotalcites were identified as phase components. This result justifies the assumption, that hydrotalcites are components of the corrosion products from irradiated fuels. To clarify the questions which bindings exist between radionuclides and secondary phases, sorption experiments were performed. The sorption experiments were performed in salt brines and in granite water using repository relevant radionuclides and minerals which are considered to represent thermodynamic final components of the secondary phases. Pu sorbed as cationic species quantitatively and the binding is covalent. In granite water the same behaviour was found for Am. (orig.) [German] Aus dem vorliegenden Bericht 'Untersuchungen zur Radionuklidfreisetzung

  7. Corrosion and protection of magnesium alloys

    Energy Technology Data Exchange (ETDEWEB)

    Ghali, E. [Laval Univ., Quebec City, PQ (Canada). Dept. of Mining and Metallurgy

    2000-07-01

    The oxide film on magnesium offers considerable surface protection in rural and some industrial environments and the corrosion rate lies between that of aluminum and low carbon steels. Galvanic coupling of magnesium alloys, high impurity content such as Ni, Fe, Cu and surface contamination are detrimental for corrosion resistance of magnesium alloys. Alloying elements can form secondary particles which are noble to the Mg matrix, thereby facilitating corrosion, or enrich the corrosion product thereby possibly inhibiting the corrosion rate. Bimetallic corrosion resistance can be increased by fluxless melt protection, choice of compatible alloys, insulating materials, and new high-purity alloys. Magnesium is relatively insensible to oxygen concentration. Pitting, corrosion in the crevices, filiform corrosion are observed. Granular corrosion of magnesium alloys is possible due to the cathodic grain-boundary constituent. More homogeneous microstructures tend to improve corrosion resistance. Under fatigue loading conditions, microcrack initiation in Mg alloys is related to slip in preferentially oriented grains. Coating that exclude the corrosive environments can provide the primary defense against corrosion fatigue. Magnesium alloys that contain neither aluminum nor zinc are the most SCC resistant. Compressive surface residual stresses as that created by short peening increase SCC resistance. Cathodic polarization or cladding with a SCC resistant sheet alloy are good alternatives. Effective corrosion prevention for magnesium alloy components and assemblies should start at the design stage. Selective surface preparation, chemical treatment and coatings are recommended. Oil application, wax coating, anodizing, electroplating, and painting are possible alternatives. Recently, it is found that a magnesium hydride layer, created on the magnesium surface by cathodic charging in aqueous solution is a good base for painting. (orig.)

  8. Corrosion and electrochemical behavior of boron/aluminum composites

    International Nuclear Information System (INIS)

    Pohlman, S.L.

    1976-01-01

    The results of an investigation to determine the importance of galvanic corrosion as a mechanism for the interfacial attack in boron/aluminium composites are reported. The results indicated that galvanic corrosion occurred between the aluminium matrix and the aluminium boride intermetallic formed during fabrication at the matrix/filament interface. Electric current measurements revealed that the aluminium matrix was preferentially attacked and the interfacial boride was cathodically protected. 18 references

  9. Limits to the use of highly compacted bentonite as a deterrent for microbially influenced corrosion in a nuclear fuel waste repository

    International Nuclear Information System (INIS)

    Stroes-Gascoyne, Simcha; Hamon, Connie J.; Maak, Peter

    2010-01-01

    Highly compacted bentonite-based sealing materials are being developed for use in future geological repositories for nuclear fuel waste. Such materials would ensure a diffusion-controlled hydrology and additionally form a sorption barrier against radionuclide migration after container breach. Due to some inherent physical characteristics, such as low water activity (a w ), small pores and high swelling pressure, an additional role of highly compacted bentonite may be the elimination of significant microbial activity near used fuel containers, which would reduce the occurrence of microbially influenced corrosion (MIC) to insignificant levels. Several recent studies have examined the indigenous microbial populations in compacted bentonite and the factors that control microbial activity in such environments. Laboratory experiments with Wyoming MX-80 bentonite plugs, compacted to dry densities (DD's) of 0.8 to 2.0 g/cm 3 , and infiltrated with sterile distilled deionised water were carried out. At DD's of 0.8 and 1.3 g/cm 3 , culturability of heterotrophic aerobic bacteria increased by up to four orders of magnitude above back-ground levels. Anaerobic heterotrophic bacteria and SRB did not increase significantly above background levels in any of the tests. At higher DD's all culturability remained at, or fell below, the background levels. However, even at the highest DD tested, some culturability remained and viability was only mildly affected by high DD's. Therefore, the potential for increased microbial activity exist if a substantial reduction in DD of bentonite were to occur in a repository. The microbes that survive in dry as-purchased or highly compacted bentonite appear to be largely spore-forming organisms. Chi Fru and Athar (2008) studied the bacterial colonization of compacted MX-80 bentonite from the surrounding granitic groundwater population, at various temperature ranges. Results suggested that high temperature rather than high DD

  10. Influence of alkali metal oxides and alkaline earth metal oxides on the mitigation of stress corrosion cracking in CANDU fuel sheathing

    Energy Technology Data Exchange (ETDEWEB)

    Metzler, J.; Ferrier, G.A.; Farahani, M.; Chan, P.K.; Corcoran, E.C., E-mail: Joseph.Metzler@rmc.ca [Royal Military College of Canada, Kingston, ON (Canada)

    2015-07-01

    This work investigates strategies to mitigate stress corrosion cracking (SCC) in Zircaloy-4 sheathing materials. The CANLUB coatings currently used in CANDU reactors contain both alkali metal and alkaline earth metal impurities, which can exist as oxides (e.g., Na{sub 2}O and CaO). It is believed that when the corrosive fission product iodine reacts with these oxides, the iodine can be sequestered through the formation of an iodide (e.g.,NaI and CaI{sub 2}). The subsequent O{sub 2} release may repair cracks in the protective ZrO{sub 2} layer on the sheathing, shielding the Zircaloy-4 sheathing from further corrosive fission product attack. For this investigation, O{sub 2} gas, Na{sub 2}O, and CaO were separately introduced into an environment wherein slotted Zircaloy-4 rings endure mechanical stresses in iodine vapour at high temperatures. Controlled additions of O{sub 2} gas created a slight reduction in the corrosive attack on Zircaloy-4 sheathing, while the inclusion of Na{sub 2}O and CaO lead to greater reductions. (author)

  11. Dry sliding wear behavior and corrosion resistance of NiCrBSi coating deposited by activated combustion-high velocity air fuel spray process

    International Nuclear Information System (INIS)

    Liu, Shenglin; Zheng, Xueping; Geng, Gangqiang

    2010-01-01

    NiCrBSi is a Ni-based superalloy widely used to obtain high wear and corrosion resistant coatings. This Ni-based alloy coating has been deposited onto 0Cr13Ni5Mo stainless steel using the AC-HVAF technique. The structure and morphologies of the Ni-based coatings were investigated by X-ray diffraction (XRD), scanning electron microscopy (SEM) and energy-dispersive spectrometer (EDS). The wear resistance and corrosion resistance were studied. The tribological behaviors were evaluated using a HT-600 wear test rig. The wear resistance of the Ni-based coating was shown to be higher than that of the 0Cr13Ni5Mo stainless steel because Fe 3 B, with high hardness, was distributed in the coating so the dispersion strengthening in the Ni-based coating was obvious and this increased the wear resistance of the Ni-based coating in a dry sliding wear test. Under the same conditions, the worn volume of 0Cr13Ni5Mo stainless steel was 4.1 times greater than that of the Ni-based coating. The wear mechanism is mainly fatigue wear. A series of the electrochemical tests was carried out in a 3.5 wt.% NaCl solution in order to examine the corrosion behavior. The mechanisms for corrosion resistance are discussed.

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

    Czech Academy of Sciences Publication Activity Database

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

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

  13. Corrosion inhibitors

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  14. Radionuclide release from research reactor spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Curtius, H., E-mail: h.curtius@fz-juelich.de [Forschungszentrum Juelich, Institut fuer Energieforschung, IEF-6 Sicherheitsforschung und Reaktortechnik, Geb. 05.3, D-52425 Juelich (Germany); Kaiser, G.; Mueller, E.; Bosbach, D. [Forschungszentrum Juelich, Institut fuer Energieforschung, IEF-6 Sicherheitsforschung und Reaktortechnik, Geb. 05.3, D-52425 Juelich (Germany)

    2011-09-01

    Numerous investigations with respect to LWR fuel under non oxidizing repository relevant conditions were performed. The results obtained indicate slow corrosion rates for the UO{sub 2} fuel matrix. Special fuel-types (mostly dispersed fuels, high enriched in {sup 235}U, cladded with aluminium) are used in German research reactors, whereas in German nuclear power plants, UO{sub 2}-fuel (LWR fuel, enrichment in {sup 235}U up to 5%, zircaloy as cladding) is used. Irradiated research reactor fuels contribute less than 1% to the total waste volume. In Germany, the state is responsible for fuel operation and for fuel back-end options. The institute for energy research (IEF-6) at the Research Center Juelich performs investigation with irradiated research reactor spent fuels under repository relevant conditions. In the study, the corrosion of research reactor spent fuel has been investigated in MgCl{sub 2}-rich salt brine and the radionuclide release fractions have been determined. Leaching experiments in brine with two different research reactor fuel-types were performed in a hot cell facility in order to determine the corrosion behaviour and the radionuclide release fractions. The corrosion of two dispersed research reactor fuel-types (UAl{sub x}-Al and U{sub 3}Si{sub 2}-Al) was studied in 400 mL MgCl{sub 2}-rich salt brine in the presence of Fe{sup 2+} under static and initially anoxic conditions. Within these experimental parameters, both fuel types corroded in the experimental time period of 3.5 years completely, and secondary alteration phases were formed. After complete corrosion of the used research reactor fuel samples, the inventories of Cs and Sr were quantitatively detected in solution. Solution concentrations of Am and Eu were lower than the solubility of Am(OH){sub 3}(s) and Eu(OH){sub 3}(s) solid phases respectively, and may be controlled by sorption processes. Pu concentrations may be controlled by Pu(IV) polymer species, but the presence of Pu(V) and Pu

  15. Underground pipeline corrosion

    CERN Document Server

    Orazem, Mark

    2014-01-01

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

  16. Fuel assembly

    International Nuclear Information System (INIS)

    Kawai, Mitsuo.

    1988-01-01

    Purpose: To reduce the corrosion rate and suppress the increase of radioactive corrosion products in reactor water of nuclear fuel assemblies for use in BWR type reactors having spacer springs made of nickel based deposition reinforced type alloys. Constitution: Spacer rings made of nickel based deposition reinforced type alloy are incorporated and used as fuel assemblies after applying treatment of dipping and maintaining at high temperature water followed by heating in steams. Since this can remove the nickel leaching into reactor water at the initial stage, Co-58 as the radioactive corrosion products in the reactor water can be reduced, and the operation at in-service inspection or repairement can be facilitated to improve the working efficiency of the nuclear power plant. The dipping time is desirably more than 10 hours and more desirably more than 30 hours. (Horiuchi, T. )

  17. Fuel element

    International Nuclear Information System (INIS)

    Armijo, J.S.

    1976-01-01

    A fuel element for nuclear reactors is proposed which has a higher corrosion resisting quality in reactor operations. The zirconium alloy coating around the fuel element (uranium or plutonium compound) has on its inside a protection layer of metal which is metallurgically bound to the substance of the coating. As materials are namned: Alluminium, copper, niobium, stainless steel, and iron. This protective metallic layer has another inner layer, also metallurgically bound to its surface, which consists usually of a zirconium alloy. (UWI) [de

  18. Corrosion of carbon steel in oxidizing caustic solutions

    International Nuclear Information System (INIS)

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

    1984-01-01

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

  19. Modelling of zirconium alloys corrosion in LWRs

    International Nuclear Information System (INIS)

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

    1999-01-01

    Chemical parameters, that exerted effect on Zr+1%Nb alloy corrosion and deserved consideration during reactor operation, were defined and a model was developed to describe the influence of physical and chemical parameters on zirconium alloys