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.

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.)

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

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

International Nuclear Information System (INIS)

Dobrevski, I.; Zaharieva, N.

2008-01-01

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

Corrosion of research reactor Al-clad spent fuel in water

International Nuclear Information System (INIS)

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

2002-01-01

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

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

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)

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

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

International Nuclear Information System (INIS)

2009-12-01

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

Corrosion report for the U-Mo fuel concept

Energy Technology Data Exchange (ETDEWEB)

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

2014-08-28

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

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.

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

Corrosion of spent Advanced Test Reactor fuel

International Nuclear Information System (INIS)

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

1994-01-01

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

Control of corrosion in an aqueous nuclear fuel storage basin

International Nuclear Information System (INIS)

Zimmerman, C.A.

1981-01-01

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

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

International Nuclear Information System (INIS)

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

1996-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1996-02-20

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

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)

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

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)

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

PWR fuel rod corrosion in Japan

International Nuclear Information System (INIS)

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

1997-01-01

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

PWR fuel rod corrosion in Japan

Energy Technology Data Exchange (ETDEWEB)

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

1997-02-01

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

Evaluation of corrosion on the fuel performance of stainless steel cladding

Directory of Open Access Journals (Sweden)

de Souza Gomes Daniel

2016-01-01

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

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

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.

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

Energy Technology Data Exchange (ETDEWEB)

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

1997-02-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1997-02-01

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

Problems raised by corrosion in the nuclear fuel cycle

International Nuclear Information System (INIS)

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

1977-01-01

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

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

International Nuclear Information System (INIS)

Howell, J.P.

1999-01-01

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

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

International Nuclear Information System (INIS)

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

1997-01-01

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

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

International Nuclear Information System (INIS)

Kovyrshin, V.; Zaritsky, N.

1999-01-01

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

Corrosion of research reactor aluminium clad spent fuel in water

International Nuclear Information System (INIS)

2003-01-01

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

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.

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.

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

International Nuclear Information System (INIS)

Finch, R. J.

1999-01-01

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

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

International Nuclear Information System (INIS)

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

1991-01-01

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

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.

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.

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)

Effects of alpha-decay on spent fuel corrosion behaviour

International Nuclear Information System (INIS)

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

2004-01-01

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

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

Science.gov (United States)

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

2005-01-01

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

Corrosion effect of fast reactor fuel claddings on their mechanical properties

International Nuclear Information System (INIS)

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

1985-01-01

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

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

Science.gov (United States)

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

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

Flue gas corrosion through halogen compounds in fuel gas

Energy Technology Data Exchange (ETDEWEB)

Eisenmann, R

1987-04-01

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

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

International Nuclear Information System (INIS)

Ramanathan, Lalgudi V.

2009-01-01

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

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

International Nuclear Information System (INIS)

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

1985-03-01

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

Prevention of nuclear fuel cladding materials corrosion

International Nuclear Information System (INIS)

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

1983-01-01

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

Corrosion Studies of Platinum Nano-Particles for Fuel Cells

DEFF Research Database (Denmark)

Shim, Signe Sarah

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

Corrosion testing of uranium silicide fuel specimens

International Nuclear Information System (INIS)

Bourns, W.T.

1968-09-01

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

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

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.

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

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

International Nuclear Information System (INIS)

Park, Moon Ghu; Lee, Sang Hee

1997-01-01

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

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

International Nuclear Information System (INIS)

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

1976-05-01

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

Corrosion testing of uranium silicide fuel specimens

Energy Technology Data Exchange (ETDEWEB)

Bourns, W T

1968-09-15

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

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.)

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

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

International Nuclear Information System (INIS)

Schoenfeld, R.; Wegner, K.

1989-03-01

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

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

International Nuclear Information System (INIS)

Forsyth, R.S.

1991-01-01

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

Drying characteristics of thorium fuel corrosion products

Energy Technology Data Exchange (ETDEWEB)

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

2004-07-01

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

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

Science.gov (United States)

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

2015-08-01

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

Influence of LMFBR fuel pin temperature profiles on corrosion rate

International Nuclear Information System (INIS)

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

1976-01-01

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

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.)

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

International Nuclear Information System (INIS)

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

1992-03-01

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

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

International Nuclear Information System (INIS)

Seljak, Tine; Širok, Brane; Katrašnik, Tomaž

2016-01-01

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

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.

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)

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)

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.

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

International Nuclear Information System (INIS)

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

2002-01-01

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

Corrosion of aluminium-clad spent fuel at RA research reactor

International Nuclear Information System (INIS)

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

2003-01-01

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

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

Science.gov (United States)

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

2010-04-20

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

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.

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

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

International Nuclear Information System (INIS)

Christensen, H.; Sunder, S.

2000-07-01

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

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

International Nuclear Information System (INIS)

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

1996-01-01

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

Fuel element failures caused by iodine stress corrosion

International Nuclear Information System (INIS)

Videm, K.; Lunde, L.

1976-01-01

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

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

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.

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

International Nuclear Information System (INIS)

Brodda, B.G.; Fachinger, J.

1995-01-01

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

Corrosion studies on retrievable spent fuel containers: a progress report

International Nuclear Information System (INIS)

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

1978-12-01

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

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...

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

Science.gov (United States)

Ghosh, Debashis; Mitra, Swapan Kumar

2011-04-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2010-07-01

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

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

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

Directory of Open Access Journals (Sweden)

Majid Asnavandi

2017-01-01

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

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

International Nuclear Information System (INIS)

2013-03-01

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

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)

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)

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)

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

DEFF Research Database (Denmark)

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

2008-01-01

Over the past few years, considerable high temperature corrosion problems have been encountered when firing biomass in power plants due to the high content of potassium chloride in the deposits. Therefore, to combat chloride corrosion problems cofiring of biomass with a fossil fuel has been...... undertaken. This results in potassium chloride being converted to potassium sulphate in the combustion chamber and it is sulphate rich deposits that are deposited on the vulnerable metallic surfaces such as high temperature superheaters. Although this removes the problem of chloride corrosion, other...... corrosion mechanisms appear such as sulphidation and hot corrosion due to sulphate deposits. At Studstrup power plant Unit 4, based on trials with exposure times of 3000 h using 0–20% straw co-firing with coal, the plant now runs with a fuel mix of 10% strawþcoal. Based on results from a 3 years exposure...

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

International Nuclear Information System (INIS)

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

2004-12-01

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

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

International Nuclear Information System (INIS)

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

2014-01-01

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

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

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)

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)

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

Science.gov (United States)

Gannon, Paul; Amendola, Roberta

2012-12-01

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

Corrosion assessment of dry fuel storage containers

Energy Technology Data Exchange (ETDEWEB)

Graves, C.E.

1994-09-01

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

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

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

Energy Technology Data Exchange (ETDEWEB)

Bankiewicz, D.

2012-11-01

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

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

International Nuclear Information System (INIS)

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

1996-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1996-10-01

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

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

Uranyl peroxide enhanced nuclear fuel corrosion in seawater.

Science.gov (United States)

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

2012-02-07

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

Carbonate fuel cell endurance: Hardware corrosion and electrolyte management status

Energy Technology Data Exchange (ETDEWEB)

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

1993-01-01

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

Carbonate fuel cell endurance: Hardware corrosion and electrolyte management status

Energy Technology Data Exchange (ETDEWEB)

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

1993-05-01

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

Corrosion of Spent Nuclear Fuel: The Long-Term Assessment

International Nuclear Information System (INIS)

Ewing, Rodney C.

2003-01-01

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

Copper anode corrosion affects power generation in microbial fuel cells

KAUST Repository

Zhu, Xiuping

2013-07-16

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

Copper anode corrosion affects power generation in microbial fuel cells

KAUST Repository

Zhu, Xiuping; Logan, Bruce E.

2013-01-01

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

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

International Nuclear Information System (INIS)

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

1997-01-01

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

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)

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...

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.

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...

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

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.

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.

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.

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

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

Corrosion of copper under Canadian nuclear fuel waste disposal conditions

International Nuclear Information System (INIS)

King, F.; Litke, C.D.

1990-01-01

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

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

International Nuclear Information System (INIS)

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

2009-01-01

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

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

International Nuclear Information System (INIS)

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

1997-01-01

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

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

International Nuclear Information System (INIS)

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

2007-01-01

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

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

International Nuclear Information System (INIS)

Totemeier, T.C.

1999-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Totemeier, T.C.

1999-04-26

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

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

International Nuclear Information System (INIS)

King, F.; Kolar, M.

1997-12-01

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

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

Science.gov (United States)

2010-08-01

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

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

International Nuclear Information System (INIS)

Pantera, L.

1992-11-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1995-05-01

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

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)

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.

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

International Nuclear Information System (INIS)

DUNCAN, D.R.

1999-01-01

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

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

International Nuclear Information System (INIS)

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

1996-01-01

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

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

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

International Nuclear Information System (INIS)

Lott, Randy G.

2003-01-01

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

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

International Nuclear Information System (INIS)

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

2015-01-01

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

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

International Nuclear Information System (INIS)

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

2012-09-01

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

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

International Nuclear Information System (INIS)

Hirao, Keiichi; Yamane, Toshimi; Minamino, Yoritoshi

1991-01-01

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

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

International Nuclear Information System (INIS)

Durazzo, M.

2005-01-01

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

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

International Nuclear Information System (INIS)

Wollam, C.D.

1980-09-01

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

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

Science.gov (United States)

2012-06-01

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

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.

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

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

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

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.

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)

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

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

International Nuclear Information System (INIS)

King, F.

1996-11-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2010-06-01

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

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)

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

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

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

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

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

International Nuclear Information System (INIS)

Sejnoha, R.; Wood, J.C.

1978-01-01

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

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

International Nuclear Information System (INIS)

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

1986-01-01

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

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.

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.

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

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

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)

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?

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)

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

Energy Technology Data Exchange (ETDEWEB)

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

1996-09-01

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

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

International Nuclear Information System (INIS)

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

1996-09-01

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

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

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

Czech Academy of Sciences Publication Activity Database

Medek, Jiří; Weishauptová, Zuzana

2009-01-01

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

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

International Nuclear Information System (INIS)

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

2008-01-01

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

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

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.

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

DEFF Research Database (Denmark)

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

2014-01-01

This paper describes experimental investigations on corrosion-induced deterioration in reinforced cementitious materials and the subsequent development and implementation of a novel conceptual model. Rejnforced mortar specimens of varying water-to-cement ratios were subjected to current-induced c......This paper describes experimental investigations on corrosion-induced deterioration in reinforced cementitious materials and the subsequent development and implementation of a novel conceptual model. Rejnforced mortar specimens of varying water-to-cement ratios were subjected to current......-dependent concentrations of corrosion products averaged through the specimen thickness. Digital image correlation (DIC) was used to measure corrosion-induced deformations including deformations between steel and cementitious matrix as well as formation and propagation of corrosion-induced cracks. Based on experimental...... observations, a conceptual model was developed to describe the penetration of solid corrosion products into capillary pores of the cementitious matrix. Only capillary pores within a corrosion accommodating region (CAR), i.e. in close proximity of the steel reinforcement, were considered accessible...

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

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

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)

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

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...

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

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

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

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)

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

International Nuclear Information System (INIS)

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

2004-01-01

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

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

International Nuclear Information System (INIS)

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

2002-12-01

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

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

International Nuclear Information System (INIS)

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

1995-01-01

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

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

DEFF Research Database (Denmark)

Montgomery, Melanie; Karlsson, A; Larsen, OH

2002-01-01

plants. The type of corrosion attack can be directly ascribed to the composition of the deposit and the metal surface temperature. A series of field tests have been undertaken in the various straw-fired power plants in Denmark, namely Masnedø, Rudkøbing and Ensted. Three types of exposure were undertaken......In Denmark, straw and other types of biomass are used for generating energy in power plants. Straw has the advantage that it is a "carbon dioxide neutral fuel" and therefore environmentally acceptable. Straw combustion is associated with corrosion problems which are not encountered in coal-fired...... to investigate corrosion: a) the exposure of metal rings on water/air cooled probes, b) the exposure of test tubes in a test superheater, and c) the exposure of test tubes in existing superheaters. Thus both austenitic steels and ferritic steels were exposed in the steam temperature range of 450-600°C...

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

International Nuclear Information System (INIS)

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

2010-01-01

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

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.

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.

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.

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

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.

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-12-15

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

Corrosion characteristics of K-claddings

International Nuclear Information System (INIS)

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

2004-01-01

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

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

International Nuclear Information System (INIS)

Forsyth, R.

1987-03-01

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

Effects of Alloyed Carbon on the General Corrosion and the Pitting Corrosion Behavior of FeCrMnN Stainless Steels

Energy Technology Data Exchange (ETDEWEB)

Ha, Heon-Young; Lee, Tae-Ho; Kim, Sung-Joon [Korea Institute of Materials Science, Changwon (Korea, Republic of)

2011-10-15

The effects of alloyed carbon on the pitting corrosion, the general corrosion, and the passivity behavior of Fe{sub 1}8Cr{sub 1}0Mn{sub 0}.4Nx{sub C} (x=0 ⁓ 0.38 wt%) alloys were investigated by various electrochemical methods and XPS analysis. The alloyed carbon increased the general corrosion resistance of the FeCrMnN matrix. Carbon enhanced the corrosion potential, reduced the metal dissolution rate, and accelerated the hydrogen evolution reaction rate in various acidic solutions. In addition, carbon promoted the pitting corrosion resistance of the matrix in a chloride solution. The alloyed carbon in the matrix increased the chromium content in the passive film, and thus the passive film became more protective.

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.

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.

The development of lower enrichment fuels for Canadian research reactors

Energy Technology Data Exchange (ETDEWEB)

Feraday, M A; Belanger, L; Grolway, C M [AECL, Atomic Energy of Canada Limited, Chalk River, ON (Canada); Foo, M T [CRNL, Combustion Engineering Superheater Ltd., Moncton, NB (Canada)

1983-08-01

As part of the world wide move to proliferation resistant fuels, new fuels which use reduced enrichment uranium are being developed for use in the NRX and NRU reactors. A fuel consisting of particles of a USiAl alloy dispersed in an Al matrix has been selected for development along with Al-37 wt% U alloy and Al-U{sub 3}O{sub 8} cermet as backup fuels. This report outlines the progress made in the development of the Al-USiAl and Al-37 wt% U. Results show that good quality extruded rods containing either fuel can be made with techniques similar to those used to fabricate the current NRX and NRU fuels. However, the new fuels will be more expensive to make. Although the oxidation behaviour of the Al-USiAl is not as good as that of the Al-U alloys, its corrosion behaviour in high temperature water does not seem much worse. The oxidation and aqueous corrosion of A-37 wt% U are not much different from those of the Al-U alloys currently used. (author)

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)

Some proposed mechanisms for internal cladding corrosion

International Nuclear Information System (INIS)

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

1977-01-01

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

Some proposed mechanisms for internal cladding corrosion

Energy Technology Data Exchange (ETDEWEB)

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

1977-04-01

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

High temperature corrosion of metallic interconnects in solid oxide fuel cells

International Nuclear Information System (INIS)

Bastidas, D. M.

2006-01-01

Research and development has made it possible to use metallic interconnects in solid oxide fuel cells (SOFC) instead of ceramic materials. The use of metallic interconnects was formerly hindered by the high operating temperature, which made the interconnect degrade too much and too fast to be an efficient alternative. When the operating temperature was lowered, the use of metallic interconnects proved to be favourable since they are easier and cheaper to produce than ceramic interconnects. However, metallic interconnects continue to be degraded despite the lowered temperature, and their corrosion products contribute to electrical degradation in the fuel cell. coatings of nickel, chromium, aluminium, zinc, manganese, yttrium or lanthanum between the interconnect and the electrodes reduce this degradation during operation. (Author) 66 refs

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

Colloids from the aqueous corrosion of uranium nuclear fuel

Science.gov (United States)

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

2005-12-01

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

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

International Nuclear Information System (INIS)

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

2014-01-01

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

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

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)

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

A study on dissolution and leaching behaviour of spent nuclear fuels

International Nuclear Information System (INIS)

Lee, Chang Heon; Im, Hee Jung; Kim, Jong Gu; Park, Yang Soon; Ha, Yeong Keong

2010-12-01

This state of the art report describes a leaching behaviour of spent nuclear fuels which should be considered for safety assessment of spent nuclear fuel disposal in a deep geological repository. A decisive factor of a dissolution of UO 2 , a matrix of the fuel, is chemical characters (redox potential, pH, concentration of inorganic anions, water radiolysis subsequent by radiation field of the fuels) of ground water expected to be in contact with the fuels after the container has failed due to corrosion as well as atmosphere condition of a deep geological repository, which can change the oxidation state of UO 2 . The release rates of radionuclides from UO 2 matrix depend largely on their location within the fuels, that is, the radionuclides fixed in the fuel/cladding gap and grain boundaries are rapidly released. However, the radionuclides within the grains of the fuel are slowly released, and then their release rate is governed by a dissolution behaviour of UO 2

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)

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.

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)

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

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

International Nuclear Information System (INIS)

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

1987-04-01

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

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

International Nuclear Information System (INIS)

Poblete Maturana, Tomas

2015-01-01

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

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.)

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)

In-reactor fuel cladding external corrosion measurement process and results

International Nuclear Information System (INIS)

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

1999-01-01

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

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

International Nuclear Information System (INIS)

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

2011-01-01

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

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.

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.

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.

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

International Nuclear Information System (INIS)

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

2006-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-05-15

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

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

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

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

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.

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

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

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

International Nuclear Information System (INIS)

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

2001-01-01

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

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

International Nuclear Information System (INIS)

King, F.

1996-11-01

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

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.

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

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.)

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.

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)

A systematic approach for development of a PWR cladding corrosion model

International Nuclear Information System (INIS)

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

2001-01-01

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

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

Directory of Open Access Journals (Sweden)

Gergely András

2016-10-01

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

Semi-empirical corrosion model for Zircaloy-4 cladding

International Nuclear Information System (INIS)

Nadeem Elahi, Waseem; Atif Rana, Muhammad

2015-01-01

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

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

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

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.)

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

Oxidative dissolution of spent fuel and release of nuclides from a copper/iron canister. Model developments and applications

Energy Technology Data Exchange (ETDEWEB)

Longcheng Liu

2001-12-01

Three models have been developed and applied in the performance assessment of a final repository. They are based on accepted theories and experimental results for known and possible mechanisms that may dominate in the oxidative dissolution of spent fuel and the release of nuclides from a canister. Assuming that the canister is breached at an early stage after disposal, the three models describe three sub-systems in the near field of the repository, in which the governing processes and mechanisms are quite different. In the model for the oxidative dissolution of the fuel matrix, a set of kinetic descriptions is provided that describes the oxidative dissolution of the fuel matrix and the release of the embedded nuclides. In particular, the effect of autocatalytic reduction of hexavalent uranium by dissolved H{sub 2}, using UO{sub 2} (s) on the fuel pellets as a catalyst, is taken into account. The simulation results suggest that most of the radiolytic oxidants will be consumed by the oxidation of the fuel matrix, and that much less will be depleted by dissolved ferrous iron. Most of the radiolytically produced hexavalent uranium will be reduced by the autocatalytic reaction with H{sub 2} on the fuel surface. It will reprecipitate as UO{sub 2} (s) on the fuel surface, and thus very little net oxidation of the fuel will take place. In the reactive transport model, the interactions of multiple processes within a defective canister are described, in which numerous redox reactions take place as multiple species diffuse. The effect of corrosion of the cast iron insert of the canister and the reduction of dissolved hexavalent uranium by ferrous iron sorbed onto iron corrosion products and by dissolved H{sub 2} are particularly included. Scoping calculations suggest that corrosion of the iron insert will occur primarily under anaerobic conditions. The escaping oxidants from the fuel rods will migrate toward the iron insert. Much of these oxidants will, however, be consumed

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

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

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

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

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

International Nuclear Information System (INIS)

Bairiot, H.

1983-01-01

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

Fighting corrosion in India

Energy Technology Data Exchange (ETDEWEB)

Rajagopalan, K S; Rangaswamy, N S

1979-03-01

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

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)

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.

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

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.

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)

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)

Influence of Microstructure on Corrosion Property of Mg-Al-Zn Alloy

International Nuclear Information System (INIS)

Lee, Jeong Ja; Na, Seung Chan; Yang, Won Seong; Hwang, WoonSuk; Jang, Si Sung; Yoo, Hwang Ryong

2006-01-01

Influence of microstructure on the corrosion property of Mg-Al-Zn Alloy was investigated using potentiodynamic polarization experiments, galvanic coupling experiments, and scanning electron microscopy in sodium chloride solutions. Pitting was the mot common form of attack in chloride solution, and filiform corrosion was also occurred in AZ91D-T4 alloy. On the contrary, filiform attack in the bulk matrix was predominant corrosion form in AZ91D-T6 alloy, and the number and size of pit were decreased than those of AZ91D-T4 alloy. Galvanic coupling effect between Mg 17 Al 12 and matrix was existed, but the propagation of galvanic corrosion was localized only near the Mg 17 Al 12 phase in AZ91D-6T alloy. The corrosion resistance of Mg-Al matrix increased with decreasing Al content in the matrix. And, it could be regarded that Al content in the matrix is decreased by precipitation of Mg 17 Al 12 curing the aging treatment and it decreases the anodic reaction rate of the matrix and galvanic effect in AZ91D-T6 alloy. It could be considered that the composition and macrostructure of surface protective layer would be varied by precipitation of Mg 17 Al 12 and subsequent decreasing of Al content in the matrix. And it would contribute the corrosion resistance of AZ91D-T6 aging alloy

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)

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.

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.

Corrosion characteristics of copper and leaded bronze in palm biodiesel

Energy Technology Data Exchange (ETDEWEB)

Haseeb, A.S.M.A.; Masjuki, H.H.; Ann, L.J.; Fazal, M.A. [Department of Mechanical Engineering, University of Malaya, 50603 Kuala Lumpur (Malaysia)

2010-03-15

Biodiesel has become more attractive as alternative fuel for automobiles because of its environmental benefits and the fact that it is made from renewable sources. However, corrosion of metals in biodiesel is one of the concerns related to biodiesel compatibility issues. This study aims to characterize the corrosion behavior of commercial pure copper and leaded bronze commonly encountered in the automotive fuel system in diesel engine. Static immersion tests in B0, B50 and B100 fuels were carried out at room temperature for 2640 h. Similar immersion tests in B0, B100 and B100 (oxidized) fuels were also conducted at 60 C for 840 h. At the end of the test, corrosion behavior was investigated by weight loss measurements and changes in surface morphology. Fuels were analyzed by using TAN analyzer, FTIR, MOA (multi-element oil analyzer) to investigate acid concentration, oxidation level with water content and corrosive impurities respectively. Results showed that under the experimental conditions, pure copper was more susceptible to corrosion in biodiesel as compared to leaded bronze. (author)

Modelling of zirconium alloys corrosion in LWRs

International Nuclear Information System (INIS)

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

1999-01-01

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

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

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

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.

Review of oxidation rates of DOE spent nuclear fuel : Part 1 : nuclear fuel

International Nuclear Information System (INIS)

Hilton, B.A.

2000-01-01

The long-term performance of Department of Energy (DOE) spent nuclear fuel (SNF) in a mined geologic disposal system depends highly on fuel oxidation and subsequent radionuclide release. The oxidation rates of nuclear fuels are reviewed in this two-volume report to provide a baseline for comparison with release rate data and technical rationale for predicting general corrosion behavior of DOE SNF. The oxidation rates of nuclear fuels in the DOE SNF inventory were organized according to metallic, Part 1, and non-metallic, Part 2, spent nuclear fuels. This Part 1 of the report reviews the oxidation behavior of three fuel types prototypic of metallic fuel in the DOE SNF inventory: uranium metal, uranium alloys and aluminum-based dispersion fuels. The oxidation rates of these fuels were evaluated in oxygen, water vapor, and water. The water data were limited to pure water corrosion as this represents baseline corrosion kinetics. Since the oxidation processes and kinetics discussed in this report are limited to pure water, they are not directly applicable to corrosion rates of SNF in water chemistry that is significantly different (such as may occur in the repository). Linear kinetics adequately described the oxidation rates of metallic fuels in long-term corrosion. Temperature dependent oxidation rates were determined by linear regression analysis of the literature data. As expected the reaction rates of metallic fuels dramatically increase with temperature. The uranium metal and metal alloys have stronger temperature dependence than the aluminum dispersion fuels. The uranium metal/water reaction exhibited the highest oxidation rate of the metallic fuel types and environments that were reviewed. Consequently, the corrosion properties of all DOE SNF may be conservatively modeled as uranium metal, which is representative of spent N-Reactor fuel. The reaction rate in anoxic, saturated water vapor was essentially the same as the water reaction rate. The long-term intrinsic

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

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

Influences of in-fuel physical-chemical processes on serviceability of energy reactor fuel elements

Energy Technology Data Exchange (ETDEWEB)

Bibilashvili, Yu K; Nekrasova, G A; Sukhanov, G I

1989-01-01

In-fuel physico-chemical processes and their effect on stress corrosion cracking of fuel element zirconium cladding are considered in the review. The mechanism of fission product release from the fuel is studied and the negative role of primarily iodine on the cladding corrosion process is demonstrated. Directions for improving the fuel element claddings and fuel to increase the fuel element serviceability are specified.

Influences of in-fuel physical-chemical processes on serviceability of energy reactor fuel elements

International Nuclear Information System (INIS)

Bibilashvili, Yu.K.; Nekrasova, G.A.; Sukhanov, G.I.

1989-01-01

In-fuel physico-chemical processes and their effect on stress corrosion cracking of fuel element zirconium cladding are considered in the review. The mechanism of fission product release from the fuel is studied and the negative role of primarily iodine on the cladding corrosion process is demonstrated. Directions for improving the fuel element claddings and fuel to increase the fuel element serviceability are specified

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

International Nuclear Information System (INIS)

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

1998-01-01

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

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)

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...

Nuclear fuel elements

International Nuclear Information System (INIS)

Obara, Hiroshi.

1981-01-01

Purpose: To suppress iodine release thereby prevent stress corrosion cracks in fuel cans by dispersing ferrous oxide at the outer periphery of sintered uranium dioxide pellets filled and sealed within zirconium alloy fuel cans of fuel elements. Constitution: Sintered uranium dioxide pellets to be filled and sealed within a zirconium alloy fuel can are prepared either by mixing ferric oxide powder in uranium dioxide powder, sintering and then reducing at low temperature or by mixing iron powder in uranium dioxide powder, sintering and then oxidizing at low temperature. In this way, ferrous oxide is dispersed on the outer periphery of the sintered uranium dioxide pellets to convert corrosive fission products iodine into iron iodide, whereby the iodine release is suppressed and the stress corrosion cracks can be prevented in the fuel can. (Moriyama, K.)

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.)

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)

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.

ELECTROCHEMICAL STUDIES OF URANIUM METAL CORROSION MECHANISM AND KINETICS IN WATER

International Nuclear Information System (INIS)

Boudanova, Natalya; Maslennikov, Alexander; Peretroukhine, Vladimir F.; Delegard, Calvin H.

2006-01-01

During long-term underwater storage of low burn-up uranium metal fuel, a corrosion product sludge forms containing uranium metal grains, uranium dioxide, uranates and, in some cases, uranium peroxide. Literature data on the corrosion of non-irradiated uranium metal and its alloys do not allow unequivocal prediction of the paragenesis of irradiated uranium in water. The goal of the present work conducted under the program 'CORROSION OF IRRADIATED URANIUM ALLOYS FUEL IN WATER' is to study the corrosion of uranium and uranium alloys and the paragenesis of the corrosion products during long-term underwater storage of uranium alloy fuel irradiated at the Hanford Site. The elucidation of the physico-chemical nature of the corrosion of irradiated uranium alloys in comparison with non-irradiated uranium metal and its alloys is one of the most important aspects of this work. Electrochemical methods are being used to study uranium metal corrosion mechanism and kinetics. The present part of work aims to examine and revise, where appropriate, the understanding of uranium metal corrosion mechanism and kinetics in water

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

Development of alternative materials for BWR fuel springs

International Nuclear Information System (INIS)

Uruma, Y.; Osato, T.; Yamazaki, K.

2002-01-01

Major sources of radioactivity introduced into reactor water of BWR were estimated fuel crud and in-core materials (especially, fuel springs). Fuel springs are used for fixation of fuel cladding tubes with spacer grid. Those are small parts (total length is only within 25 mm) and so many numbers are loaded simultaneously and then total surfaces area are calculated up to about 200 m 2 . Fuel springs are located under high radiation field and high oxidative environment. Conventional fuel spring is made of alloy-X750 which is one of nickel-based alloy and is reported to show relatively higher corrosion release rate. 58 Co and 60 Co will be released directly into reactor water from intensely radio-activated fuel springs surface and increase radioactivity concentrations in primary coolant. Corrosion release control from fuel springs is an important technical item and a development of alternative material instead of alloy-X750 for fuel spring is a key subject to achieve ultra low man-rem exposure BWR plant. In present work, alloy-X718 which started usage for PWR fuel springs and stainless steel type 316L which has many mechanical property data are picked up for alternative materials and compared their corrosion behaviors with conventional material. Corrosion experiment was conducted under vapor-water two phases flow which is simulated fuel cladding surface boiling condition. After exposure, corrosion film formed under corrosion test was analyzed in detail and corrosion film amount and corrosion release amount are estimated among three materials. (authors)

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

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

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)

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)

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

International Nuclear Information System (INIS)

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

1983-01-01

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

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

International Nuclear Information System (INIS)

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

1998-04-01

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

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

Polarization Behavior of Squeeze Cast Al2O3 Fiber Reinforced Aluminum Matrix Composites

International Nuclear Information System (INIS)

Ham, S. H.; Kang, Y. C.; Cho, K. M.; Park, I. M.

1992-01-01

Electrochemical polarization behavior of squeeze cast Al 2 O 3 short fiber reinforced Al alloy matrix composites was investigated for the basic understanding of the corrosion properties of the composites. The composites were fabricated with variations of fiber volume fraction and matrix alloys. It was found that the reinforced composites are more susceptible to corrosion attack than the unreinforced matrix alloys in general. Corrosion resistance shows decreasing tendency with increasing Al 2 O 3 fiber volume fraction in AC8A matrix. Effect of the matrix alloys revealed that the AC8A Al matrix composite is less susceptible to corrosion attack than the 2024 and 7075 Al matrix composites. Effect of plastic deformation on electrochemical polarization behavior of the squeeze cast Al/Al 2 O 3 composites was examined after extrusion of AC8A-10v/o Al 2 O 3 . Result shows that corrosion resistance is deteriorated after plastic deformation

High temperature corrosion in a biomass-fired power boiler : Reducing furnace wall corrosion in a waste wood-fired power plant with advanced steam data

OpenAIRE

Alipour, Yousef

2013-01-01

The use of waste (or recycled) wood as a fuel in heat and power stations is becoming more widespread in Sweden (and Europe), because it is CO2 neutral with a lower cost than forest fuel. However, it is a heterogeneous fuel with a high amount of chlorine, alkali and heavy metals which causes more corrosion than fossil fuels or forest fuel. A part of the boiler which is subjected to a high corrosion risk is the furnace wall (or waterwall) which is formed of tubes welded together. Waterwalls are...

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

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

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

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

International Nuclear Information System (INIS)

Verlet, Romain

2015-01-01

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

Influences of the quantity of Mg2Sn phase on the corrosion behavior of Mg-7Sn magnesium alloy

International Nuclear Information System (INIS)

Liu Xianbin; Shan Dayong; Song Yingwei; Chen Rongshi; Han Enhou

2011-01-01

The influence of the quantity of the Mg 2 Sn phase on the corrosion behavior of different solution temperature treated Mg-7Sn magnesium alloy has been investigated by electrochemical measurements, scanning electron microscope (SEM) observation, X-ray diffraction (XRD) and X-ray photoelectron spectroscopy (XPS) analysis. With the increase of solution temperature, the quantity of Mg 2 Sn phase decreased and the tin concentration of matrix increased. The dissolved tin in Mg matrix took part in the film formation and the constituent of film was magnesium oxide and stannic oxide. The corrosion mode and corrosion rate were associated with the quantity of Mg 2 Sn phases and tin concentration of the matrix. If most of tin was present as Mg 2 Sn, the corrosion mode was pitting corrosion and it accelerated the corrosion rate. If most of tin was dissolved in matrix, the corrosion mode was filiform corrosion and it decreased the corrosion rate. The experiment evidences demonstrated that the corrosion resistance can be improved by increasing the tin concentration of matrix and the lowest corrosion rate was observed for sample solution treated at 540 o C.

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)

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.)

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.

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

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

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

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.

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.

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

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.

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

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

Aqueous Corrosion Rates for Waste Package Materials

Energy Technology Data Exchange (ETDEWEB)

S. Arthur

2004-10-08

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

Aqueous Corrosion Rates for Waste Package Materials

International Nuclear Information System (INIS)

Arthur, S.

2004-01-01

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

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.)

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.

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

Microstructural aspects of zircaloy nodular corrosion in steam

International Nuclear Information System (INIS)

Taylor, D.F.

1999-01-01

Zircaloy-2 becomes susceptible to nodular corrosion in high-temperature, high-pressure steam when the total solute concentration of the β-stabilizing alloying elements Fe, Ni and Cr in the α-zirconium matrix falls below a critical value C c that is characteristic of the test conditions. C c for typical commercial Zircaloy-2 in a 24hr/510 C/10.4MPa steam-test is the precipitate-free a-matrix concentration in equilibrium with solute-saturated β phase at about 840 C, the corresponding critical temperature T c .Thus, immunity to nodular corrosion is a metastable condition for α-Zircaloy that requires fast cooling from above T c to achieve adequate solute concentration throughout the matrix. Annealing Zircaloy at any temperature below T c for a sufficiently long time makes it susceptible to nodular corrosion. In the (α+χ) phase field, where χ collectively designates the Fe-, Cr-, and Ni-containing precipitate phases, lowering the solute concentration to less than C c by Ostwald ripening can require many hundreds of hours. Above about 825 C, the temperature of the (α+χ)/(α+β+χ) transus, solute-saturated β phase surrounds each precipitate and a strong inverse activity gradient promotes equilibration with the much lower solute concentration in the α matrix. Sensitization to nodular corrosion occurs most rapidly at about 835 C between the (α+χ)/(α+β+χ) transus and T c . Annealing Zircaloy at temperatures above T c for a sufficiently long time will raise the solute concentration above C c and, with rapid cooling, heal any degree of susceptibility. Annealing within the protective coarsening window between T c and about 850 C, the temperature of the (α+β+χ)/(α+β) transus, achieves rapid precipitate growth in a matrix immune to nodular corrosion

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)

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

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

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

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

Properties of zirconium carbide for nuclear fuel applications

Energy Technology Data Exchange (ETDEWEB)

Katoh, Yutai; Vasudevamurthy, Gokul, E-mail: gvasudev@vcu.edu; Nozawa, Takashi; Snead, Lance L.

2013-10-15

Zirconium carbide (ZrC) is a potential coating, oxygen-gettering, or inert matrix material for advanced high temperature reactor fuels. ZrC has demonstrated attractive properties for these fuel applications including excellent resistance against fission product corrosion and fission product retention capabilities. However, fabrication of ZrC results in a range of stable sub-stoichiometric and carbon-rich compositions with or without substantial microstructural inhomogeneity, textural anisotropy, and a phase separation, leading to variations in physical, chemical, thermal, and mechanical properties. The effects of neutron irradiation at elevated temperatures, currently only poorly understood, are believed to be substantially influenced by those compositional and microstructural features further adding complexity to understanding the key ZrC properties. This article provides a survey of properties data for ZrC, as required by the United States Department of Energy’s advanced fuel programs in support of the current efforts toward fuel performance modeling and providing guidance for future research on ZrC for fuel applications.

Out-of-pile experiments of fuel-cladding chemical interaction, (2)

International Nuclear Information System (INIS)

Konashi, Kenji; Yato, Tadao; Kaneko, Hiromitsu; Honda, Yutaka

1980-01-01

Cesium seems to be one of the most important fission products in the fuel-cladding chemical interaction of fuel pins for LMFBRs. However the FCCI under irradiation cannot always be explained by considering only cesium-oxygen system as the corrosive, since attack does not occur in the cesium-oxygen system unless oxygen potential is sufficiently high. Cesium-tellurium-oxygen system has been proposed to account for heavy cladding attack which was sometimes found in hypostoichiometric mixed oxide fuel pins. In this paper, the experiment on the reaction of liquid tellurium with stainless steel is reported. The type 316 stainless steel claddings for Monju type fuel pins were used as the test specimens. Tellurium was contained into the cladding tubes with end plugs. The temperature dependence of the attack by tellurium was examined in the range from 450 to 900 deg C for 30 min, and the heating time dependence was examined from 5 min to 200 hr at 725 deg C. An infrared lamp furnace was used for the experiment within 7 hr, and a resistance furnace for longer experiment. The character of corrosion was matrix attack, and the reaction products on the stainless steel surfaces consisted of chrome rich inner phase and iron and nickel rich outer phase. The results are reported. (Kako, I.)

Combating corrosion in biomass and waste fired plant

Energy Technology Data Exchange (ETDEWEB)

Henderson, Pamela [Vattenfall AB, Stockholm (Sweden). Research and Development; Hjoernhede, Anders [Vattenfall AB, Gothenburg (Sweden). Power Consultant

2010-07-01

Many biomass- or waste-fired plants have problems with high temperature corrosion especially if the steam temperature is greater than 500 C. An increase in the combustion of waste fuels means that an increasing number of boilers have had problems. Therefore, there is great interest in reducing the costs associated with high temperature corrosion and at the same time there exists a desire to improve the electrical efficiency of a plant by the use of higher steam temperatures. Assuming that the fuel is well-mixed and that there is good combustion control, there are in addition a number of other measures which can be used to reduce superheater corrosion in biomass and waste fired plants, and these are described in this paper. These include the use of fuel additives, specifically sulphur-containing ones; design aspects like placing superheaters in less corrosive positions in a boiler, using tube shielding, a wider pitch between the tubes; operational considerations such as more controlled soot-blowing and the use of better materials. (orig.)

Corrosion of aluminum alloys in simulated dry storage environments

International Nuclear Information System (INIS)

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

1996-01-01

The effect of temperature and relative humidity on the high temperature (up to 150 degrees C) corrosion of aluminum alloys was investigated for dry storage of spent nuclear fuels in a closed or sealed system. A dependency on alloy type, temperature and initial humidity was determined for 1100, 5052 and 6061 aluminum alloys. Results after 4500 hours of environmental testing show that for a closed system, corrosion tends to follow a power law with the rate decreasing with increasing exposure. As corrosion takes place, two phenomena occur: (1) a hydrated layer builds up to resist corrosion, and (2) moisture is depleted from the system and the humidity slowly decreases with time. At a critical level of relative humidity, corrosion reactions stop, and no additional corrosion occurs if the system remains closed. The results form the basis for the development of an acceptance criteria for the dry storage of aluminum clad spent nuclear fuels

Issues for Storing Plant-Based Alternative Fuels in Marine Environments

Science.gov (United States)

2014-05-09

of aerobic metabolites that exacerbated subsequent corrosion processes. 15. SUBJECT TERMS biodiesel , alternative fuel, seawater, corrosion, carbon...2013 Accepted 17 December 2013 Available online 24 December 2013 Keywords: Biodiesel Alternative fuel Seawater Corrosion Carbon steel...high energy density transportation fuels 13). The feasibility of producing large quantities of fatty acid methyl ester (FAME) and lipid-based fuels

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.

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.

High Temperature Corrosion in Biomass-Fired Boilers

DEFF Research Database (Denmark)

Henriksen, Niels; Montgomery, Melanie; Hede Larsen, Ole

2002-01-01

condense on superheater components. This gives rise to specific corrosion problems not previously encountered in coal-fired power plants. The type of corrosion attack can be directly ascribed to the composition of the deposit and the metal surface temperature. To avoid such high corrosion rates, woodchip...... has also been utilised as a fuel. Combustion of woodchip results in a smaller amount of ash, and potassium and chlorine are present in lesser amounts. However, significant corrosion rates were still seen. A case study of a woodchip fired boiler is described. The corrosion mechanisms in both straw-fired...... and woodchip fired boilers are discussed....

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.)

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.

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.

Aqueous corrosion study on U-Zr alloy

International Nuclear Information System (INIS)

Pal, Titas; Venkatesan, V.; Kumar, Pradeep; Khan, K.B.; Kumar, Arun

2009-01-01

In low power or research reactor, U-Zr alloy is a potential candidate for dispersion fuel. Moreover, Zirconium has a low thermal-neutron cross section and uranium alloyed with Zr has excellent corrosion resistance and dimensional stability during thermal cycling. In the present study aqueous corrosion behavior of U-Zr alloy samples was studied in autoclave at 200 deg C temperature. Corrosion rate was determined from weight loss with time. (author)

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.

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.

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.

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.

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

Spent fuel and fuel pool component integrity. Annual report, FY 1979

International Nuclear Information System (INIS)

Johnson, A.B. Jr.; Bailey, W.J.; Schreiber, R.E.; Kustas, F.M.

1980-05-01

International meetings under the BEFAST program and under INFCE Working Group No. 6 during 1978 and 1979 continue to indicate that no cases of fuel cladding degradation have developed on pool-stored fuel from water reactors. A section from a spent fuel rack stand, exposed for 1.5 y in the Yankee Rowe (PWR) pool had 0.001- to 0.003-in.-deep (25- to 75-μm) intergranular corrosion in weld heat-affected zones but no evidence of stress corrosion cracking. A section of a 304 stainless steel spent fuel storage rack exposed 6.67 y in the Point Beach reactor (PWR) spent fuel pool showed no significant corrosion. A section of 304 stainless steel 8-in.-dia pipe from the Three Mile Island No. 1 (PWR) spent fuel pool heat exchanger plumbing developed a through-wall crack. The crack was intergranular, initiating from the inside surface in a weld heat-affected zone. The zone where the crack occurred was severely sensitized during field welding. The Kraftwerk Union (Erlangen, GFR) disassembled a stainless-steel fuel-handling machine that operated for 12 y in a PWR (boric acid) spent fuel pool. There was no evidence of deterioration, and the fuel-handling machine was reassembled for further use. A spent fuel pool at a Swedish PWR was decontaminated. The procedure is outlined in this report

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)

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.

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.

pH-Sensitive Microparticles with Matrix-Dispersed Active Agent

Science.gov (United States)

Li, Wenyan (Inventor); Buhrow, Jerry W. (Inventor); Jolley, Scott T. (Inventor); Calle, Luz M. (Inventor)

2014-01-01

Methods to produce pH-sensitive microparticles that have an active agent dispersed in a polymer matrix have certain advantages over microcapsules with an active agent encapsulated in an interior compartment/core inside of a polymer wall. The current invention relates to pH-sensitive microparticles that have a corrosion-detecting or corrosion-inhibiting active agent or active agents dispersed within a polymer matrix of the microparticles. The pH-sensitive microparticles can be used in various coating compositions on metal objects for corrosion detecting and/or inhibiting.

Double shell slurry low-temperature corrosion tests

International Nuclear Information System (INIS)

Divine, J.R.; Bowen, W.M.; McPartland, S.A.; Elmore, R.P.; Engel, D.W.

1983-09-01

A series of year-long tests have been completed on potential double shell slurry (DSS) compositions at temperatures up to 100 0 C. These tests have sought data on uniform corrosion, pitting, and stress-corrosion cracking. No indication of the latter two types of corrosion were observed within the test matrix. Corrosion rates after four months were generally below the 1 mpy (25 μm/y) design limit. By the end of twelve months all results were below this limit and, except for very concentrated mixtures, all were below 0.5 mpy. Prediction equations were generated from a model fitted to the data. The equations provide a rapid means of estimating the corrosion rate for proposed DSS compositions

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

Water corrosion resistance of ODS ferritic-martensitic steel tubes

International Nuclear Information System (INIS)

Narita, Takeshi; Ukai, Shigeharu; Kaito, Takeji; Ohtsuka, Satoshi; Matsuda, Yasuji

2008-01-01

Oxide dispersion strengthened (ODS) ferritic-martensitic steels have superior radiation resistance; it is possible to achieve a service temperature of up to around 973 K because of their superior creep strength. These advantages of ODS steels facilities their application to long-life cladding tubes in advanced fast reactor fuel elements. In addition to neutron radiation resistance, sufficient general corrosion resistance to maintain the strength of the cladding, and the stress corrosion cracking (SCC) resistance for spent-fuel-pool cooling systems and high-temperature oxidation for the fuel-clad chemical interaction (FCCI) of ODS ferritic steel are required. Although the addition of Cr to ODS is effective in preventing water corrosion and high-temperature oxidation, an excessively high amount of Cr leads to embrittlement due to the formation of a Cr-rich α' precipitate. The Cr content in 9Cr-ODS martensite and 12Cr-ODS ferrite, the ODS steels developed by the Japan Atomic Energy Agency (JAEA), is controlled. In a previous paper, it has been demonstrated that the resistances of 9Cr- and 12Cr-ODS ferritic-martensitic steels for high-temperature oxidation are superior to those of conventional 12Cr ferritic steel. However, the water corrosion data of ODS ferritic-martensitic steels are very limited. In this study, a water corrosion test was conducted on ODS steels in consideration of the spent-fuel-pool cooling condition, and the results were compared with those of conventional austenitic stainless steel and ferritic-martensitic stainless steel. (author)

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.

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.

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

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

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

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

Analysis of corrosion behavior of KOFA cladding

International Nuclear Information System (INIS)

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

1994-01-01

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

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.)

Study on the behavior of waterside corroded PWR fuel rods under reactivity initiated accident conditions

International Nuclear Information System (INIS)

Sasajima, Hideo

1989-06-01

One of the highlighted problems from the fuel reliability point of view is a waterside corrosion of fuel cladding which becomes more significant at extended burnup stages. To date, at highly burned fuel, waterside corrosion was recognized as important because cladding oxidation increased with increasing burn-up. In experiments, as the basic research for the study of high burn-up fuel, the test fuel rods were prepressurized to ranges from 3.47 to 3.55 MPa, oxidized artificially to both 10 and 20 μm in thickness. Regarding fabricated oxide thickness of 10 μm, it is corresponded to be transition point from cubic law to linear law as a function of burn-up. Pulse irradiation experiments by NSRR were carried out to study the behavior of waterside corroded PWR type fuels under RIA conditions. Obtained results are: (1) The failure threshold of tested fuels was 110 cal/g·fuel (0.46 KJ/g·fuel) in enthalpy. This showed that the failure threshold of tested fuels was same as that of the past NSRR experimental data. (2) The failure mechanisms of the tested fuel rods was cladding rupture induced by ballooning. No differences in failure mechanisms existed between the past NSRR prepressurized standard fuel and the tested fuels. (3) Cracks were existed without propagating into cladding matrix, so that it was judged that these were not initiation of failure. (4) Whithin this experimental condition, reduction of cladding thickness being attributed to the increase of oxidation did not failure threshold. (author)

General considerations on the oxide fuel-cladding chemical interaction

International Nuclear Information System (INIS)

Pascard, R.

1977-01-01

Since the very first experimental irradiations in thermal reactors, performed in view of the future Rapsodie fuel general study, corrosion cladding anomalies were observed. After 10 years of Rapsodie and more than two years of Phenix, performance brought definite confirmation of the chemical reactions between the irradiated fuel and cladding. That is the reason for which the fuel designers express an urgent need for determining the corrosion rates. Semi-empirical laws and mechanisms describing corrosion processes are proposed. Erratic conditions for appearance of the oxide-cladding corrosion are stressed upon. Obviously such a problem can be fully appreciated only by a statistical approach based on a large number of observations on the true LMFBR fuel pins

Salt fog corrosion behavior in a powder-processed icosahedral-phase-strengthened aluminum alloy

International Nuclear Information System (INIS)

Watson, T.J.; Gordillo, M.A.; Ernst, A.T.; Bedard, B.A.; Aindow, M.

2017-01-01

Highlights: • Pitting corrosion resistance has been evaluated for an Al-Cr-Mn-Co-Zr alloy. • Pit densities and depths are far lower than for other high-strength Al alloys. • Corrosion proceeds by selective oxidation of the Al matrix around the other phases. - Abstract: The pitting corrosion resistance has been evaluated for a powder-processed Al-Cr-Mn-Co-Zr alloy which contains ≈35% by volume of an icosahedral quasi-crystalline phase and a little Al 9 Co 2 in an Al matrix. ASTM standard salt fog exposure tests show that the alloy exhibits far lower corrosion pit densities and depths than commercial high-strength aerospace Al alloys under the same conditions. Electron microscopy data show that the salt fog exposure leads to the selective oxidation of the face-centered cubic Al matrix around the other phases, and to the development of a porous outer oxide scale.

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.

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.

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.

Thermal behaviour properties and corrosion resistance of organoclay/polyurethane film

Science.gov (United States)

Kurniawan, O.; Soegijono, B.

2018-03-01

Organoclay/polyurethane film composite was prepared by adding organoclay with different content (1, 3, and 5 wt.%) in polyurethane as a matrix. TGA and DSC showed decomposition temperature shifted to a lower point as organoclay content change. FT-IR spectra showed chemical bonding of organoclay and polyurethane as a matrix, which means that the bonding between filler and matrix occured and the composite was stronger but less bonding occur in composite with 5 wt.% organoclay. The corrosion resistance overall increased with the increasing organoclay content. Composite with 5 wt.% organoclay had more thermal stability and corrosion resistance may probably due to exfoliation of organoclay.

Neural network for prediction of superheater fireside corrosion

Energy Technology Data Exchange (ETDEWEB)

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

1999-12-31

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