Thermal behavior analysis of U-Mo/Al dispersion fuel

Energy Technology Data Exchange (ETDEWEB)

Ryu, Ho Jin; Park, Jong Mang; Lee, Yoon Sang; Kim, Chang Kyu [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

2004-07-01

According to the non-proliferation policy under the reduced enrichment for research and test reactors (RERTR) program, low enriched uranium(LEU) fuel such as uranium silicide dispersion fuels are being used in research reactors. Because of a lower enrichment higher uranium density fuels are required for some high performance research reactors. Some uranium alloys with a high uranium density such as U-Mo alloys have been considered as one of the most promising candidates for a dispersion fuel due to the good irradiation performance. An international qualification program to replace the uranium silicide dispersion fuel with U-Mo dispersion fuel is being carried out under the RERTR program. Although U-Mo powders are conventionally supplied by the mechanical comminuting of as-cast U-Mo alloys, KAERI developed a centrifugal atomization method in order to simplify the preparation process and improve the properties. The centrifugally atomized powders have a rapidly solidified gamma uranium structure and a spherical shape. During the in-reactor operation of a dispersion fuel, interdiffusion or chemical reactions between the fuel particles and the matrix occurr. Intermetallic compounds in the form of UAlx are formed as a result of the diffusional reaction. Because the intermetallic compounds are less dense than the combined reactants, the volume of the fuel element increases after the reaction. In addition to the effect on the swelling performance, the reaction layers between the U-Mo and the Al matrix induces a degradation of the thermal properties of the U-Mo/Al dispersion fuels. It is important to investigate the thermal behavior of U-Mo/Al dispersion fuel according to reaction between the fuel particles and the matrix with the burnup and linear power. In this study, a finite element analysis was used for the calculation of the temperature distribution of the U-Mo/Al dispersion fuel with a burnup and linear power. Kinetics data of the reaction layers such as the growth

Thermal behavior analysis of U-Mo/Al dispersion fuel

International Nuclear Information System (INIS)

Ryu, Ho Jin; Park, Jong Mang; Lee, Yoon Sang; Kim, Chang Kyu

2004-01-01

According to the non-proliferation policy under the reduced enrichment for research and test reactors (RERTR) program, low enriched uranium(LEU) fuel such as uranium silicide dispersion fuels are being used in research reactors. Because of a lower enrichment higher uranium density fuels are required for some high performance research reactors. Some uranium alloys with a high uranium density such as U-Mo alloys have been considered as one of the most promising candidates for a dispersion fuel due to the good irradiation performance. An international qualification program to replace the uranium silicide dispersion fuel with U-Mo dispersion fuel is being carried out under the RERTR program. Although U-Mo powders are conventionally supplied by the mechanical comminuting of as-cast U-Mo alloys, KAERI developed a centrifugal atomization method in order to simplify the preparation process and improve the properties. The centrifugally atomized powders have a rapidly solidified gamma uranium structure and a spherical shape. During the in-reactor operation of a dispersion fuel, interdiffusion or chemical reactions between the fuel particles and the matrix occurr. Intermetallic compounds in the form of UAlx are formed as a result of the diffusional reaction. Because the intermetallic compounds are less dense than the combined reactants, the volume of the fuel element increases after the reaction. In addition to the effect on the swelling performance, the reaction layers between the U-Mo and the Al matrix induces a degradation of the thermal properties of the U-Mo/Al dispersion fuels. It is important to investigate the thermal behavior of U-Mo/Al dispersion fuel according to reaction between the fuel particles and the matrix with the burnup and linear power. In this study, a finite element analysis was used for the calculation of the temperature distribution of the U-Mo/Al dispersion fuel with a burnup and linear power. Kinetics data of the reaction layers such as the growth

High density dispersion fuel

International Nuclear Information System (INIS)

Hofman, G.L.

1996-01-01

A fuel development campaign that results in an aluminum plate-type fuel of unlimited LEU burnup capability with an uranium loading of 9 grams per cm 3 of meat should be considered an unqualified success. The current worldwide approved and accepted highest loading is 4.8 g cm -3 with U 3 Si 2 as fuel. High-density uranium compounds offer no real density advantage over U 3 Si 2 and have less desirable fabrication and performance characteristics as well. Of the higher-density compounds, U 3 Si has approximately a 30% higher uranium density but the density of the U 6 X compounds would yield the factor 1.5 needed to achieve 9 g cm -3 uranium loading. Unfortunately, irradiation tests proved these peritectic compounds have poor swelling behavior. It is for this reason that the authors are turning to uranium alloys. The reason pure uranium was not seriously considered as a dispersion fuel is mainly due to its high rate of growth and swelling at low temperatures. This problem was solved at least for relatively low burnup application in non-dispersion fuel elements with small additions of Si, Fe, and Al. This so called adjusted uranium has nearly the same density as pure α-uranium and it seems prudent to reconsider this alloy as a dispersant. Further modifications of uranium metal to achieve higher burnup swelling stability involve stabilization of the cubic γ phase at low temperatures where normally α phase exists. Several low neutron capture cross section elements such as Zr, Nb, Ti and Mo accomplish this in various degrees. The challenge is to produce a suitable form of fuel powder and develop a plate fabrication procedure, as well as obtain high burnup capability through irradiation testing

Solid oxide fuel cells fueled with reducible oxides

Science.gov (United States)

Chuang, Steven S.; Fan, Liang Shih

2018-01-09

A direct-electrochemical-oxidation fuel cell for generating electrical energy includes a cathode provided with an electrochemical-reduction catalyst that promotes formation of oxygen ions from an oxygen-containing source at the cathode, a solid-state reduced metal, a solid-state anode provided with an electrochemical-oxidation catalyst that promotes direct electrochemical oxidation of the solid-state reduced metal in the presence of the oxygen ions to produce electrical energy, and an electrolyte disposed to transmit the oxygen ions from the cathode to the solid-state anode. A method of operating a solid oxide fuel cell includes providing a direct-electrochemical-oxidation fuel cell comprising a solid-state reduced metal, oxidizing the solid-state reduced metal in the presence of oxygen ions through direct-electrochemical-oxidation to obtain a solid-state reducible metal oxide, and reducing the solid-state reducible metal oxide to obtain the solid-state reduced metal.

Oxide dispersion strengthened ferritic steels: a basic research joint program in France

Science.gov (United States)

Boutard, J.-L.; Badjeck, V.; Barguet, L.; Barouh, C.; Bhattacharya, A.; Colignon, Y.; Hatzoglou, C.; Loyer-Prost, M.; Rouffié, A. L.; Sallez, N.; Salmon-Legagneur, H.; Schuler, T.

2014-12-01

AREVA, CEA, CNRS, EDF and Mécachrome are funding a joint program of basic research on Oxide Dispersion Strengthened Steels (ODISSEE), in support to the development of oxide dispersion strengthened 9-14% Cr ferritic-martensitic steels for the fuel element cladding of future Sodium-cooled fast neutron reactors. The selected objectives and the results obtained so far will be presented concerning (i) physical-chemical characterisation of the nano-clusters as a function of ball-milling process, metallurgical conditions and irradiation, (ii) meso-scale understanding of failure mechanisms under dynamic loading and creep, and, (iii) kinetic modelling of nano-clusters nucleation and α/α‧ unmixing.

Oxide dispersion strengthened ferritic steels: a basic research joint program in France

Energy Technology Data Exchange (ETDEWEB)

Boutard, J.-L., E-mail: jean-louis.boutard@cea.fr [Cabinet du Haut-Commissaire, CEA/Saclay, 91191 Gif sur Yvette Cedex (France); Badjeck, V. [LPS, UMR CNRS 8502, Building 510, Université Paris-Sud 11, 91405 Orsay Cedex (France); Barguet, L. [LAUM, UMR CNRS 6613, Building IAM – UFR Sciences, Avenue O. Messiaen, 72085 Le Mans Cedex 9 (France); Barouh, C. [DMN/SRMP, CEA/Saclay, Building 520, 91191 Gif sur Yvette Cedex (France); Bhattacharya, A. [DMN/SRMP, CEA/Saclay, Building 520, 91191 Gif sur Yvette Cedex (France); CSNSM, UMR CNRS 8609, Université Paris-Sud 11, Buildings 104 and 108, 91405 Orsay Campus (France); Colignon, Y. [IM2NP, UMR CNRS 7334, Case 142, Faculté des Sciences, Campus de Saint Jérôme, Aix Marseille Université, 13397 Marseille Cedex 20 (France); Hatzoglou, C. [GPM, UMR CNRS 6634, Technopôle du Madrillet, Avenue de l’Université, BP12, 76801 Saint Etienne du Rouvray Cedex (France); Loyer-Prost, M. [DMN/SRMP, CEA/Saclay, Building 520, 91191 Gif sur Yvette Cedex (France); Rouffié, A.L. [DMN/SRMA, CEA/Saclay, Building 455, 91191 Gif sur Yvette Cedex (France); Sallez, N. [SIMAP, UMR CNRS 5266, INPG, Domaine Universitaire, 1130 rue de la Piscine, BP75, 38402 Saint Martin d’Hères Cedex (France); Salmon-Legagneur, H. [DMN/SRMA, CEA/Saclay, Building 455, 91191 Gif sur Yvette Cedex (France); Schuler, T. [DMN/SRMP, CEA/Saclay, Building 520, 91191 Gif sur Yvette Cedex (France)

2014-12-15

AREVA, CEA, CNRS, EDF and Mécachrome are funding a joint program of basic research on Oxide Dispersion Strengthened Steels (ODISSEE), in support to the development of oxide dispersion strengthened 9–14% Cr ferritic–martensitic steels for the fuel element cladding of future Sodium-cooled fast neutron reactors. The selected objectives and the results obtained so far will be presented concerning (i) physical–chemical characterisation of the nano-clusters as a function of ball-milling process, metallurgical conditions and irradiation, (ii) meso-scale understanding of failure mechanisms under dynamic loading and creep, and, (iii) kinetic modelling of nano-clusters nucleation and α/α′ unmixing.

Oxide dispersion strengthened ferritic steels: a basic research joint program in France

International Nuclear Information System (INIS)

Boutard, J.-L.; Badjeck, V.; Barguet, L.; Barouh, C.; Bhattacharya, A.; Colignon, Y.; Hatzoglou, C.; Loyer-Prost, M.; Rouffié, A.L.; Sallez, N.; Salmon-Legagneur, H.; Schuler, T.

2014-01-01

AREVA, CEA, CNRS, EDF and Mécachrome are funding a joint program of basic research on Oxide Dispersion Strengthened Steels (ODISSEE), in support to the development of oxide dispersion strengthened 9–14% Cr ferritic–martensitic steels for the fuel element cladding of future Sodium-cooled fast neutron reactors. The selected objectives and the results obtained so far will be presented concerning (i) physical–chemical characterisation of the nano-clusters as a function of ball-milling process, metallurgical conditions and irradiation, (ii) meso-scale understanding of failure mechanisms under dynamic loading and creep, and, (iii) kinetic modelling of nano-clusters nucleation and α/α′ unmixing

Performance Evaluation of Metallic Dispersion Fuel for Advanced Research Reactors

Energy Technology Data Exchange (ETDEWEB)

Ryu, Ho Jin; Park, Jong Man; Kim, Chang Kyu; Chae, Hee Taek; Song, Kee Chan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Yeon Soo [Argonne National Laboratory, New York (United States)

2007-07-01

Uranium alloys with a high uranium density has been developed for high power research reactor fuel using low-enriched uranium (LEU). U-Mo alloys have been developed as candidate fuel material because of excellent irradiation behavior. Irradiation behavior of U-Mo/Al dispersion fuel has been investigated to develop high performance research reactor fuel as RERTR international research program. While plate-type and rod-type dispersion fuel elements are used for research reactors, HANARO uses rod-type dispersion fuel elements. PLATE code is developed by Argonne National Laboratory for the performance evaluation of plate-type dispersion fuel, but there is no counterpart for rod-type dispersion fuel. Especially, thermal conductivity of fuel meat decreases during the irradiation mainly because of interaction layer formation at the interface between the U-Mo fuel particle and Al matrix. The thermal conductivity of the interaction layer is not as high as the Al matrix. The growth of interaction layer is interactively affected by the temperature of fuel because it is associated with a diffusion reaction which is a thermally activated process. It is difficult to estimate the temperature profile during irradiation test due to the interdependency of fuel temperature and thermal conductivity changed by interaction layer growth. In this study, fuel performance of rod-type U-Mo/Al dispersion fuels during irradiation tests were estimated by considering the effect of interaction layer growth on the thermal conductivity of fuel meat.

Performance Evaluation of Metallic Dispersion Fuel for Advanced Research Reactors

International Nuclear Information System (INIS)

Ryu, Ho Jin; Park, Jong Man; Kim, Chang Kyu; Chae, Hee Taek; Song, Kee Chan; Kim, Yeon Soo

2007-01-01

Uranium alloys with a high uranium density has been developed for high power research reactor fuel using low-enriched uranium (LEU). U-Mo alloys have been developed as candidate fuel material because of excellent irradiation behavior. Irradiation behavior of U-Mo/Al dispersion fuel has been investigated to develop high performance research reactor fuel as RERTR international research program. While plate-type and rod-type dispersion fuel elements are used for research reactors, HANARO uses rod-type dispersion fuel elements. PLATE code is developed by Argonne National Laboratory for the performance evaluation of plate-type dispersion fuel, but there is no counterpart for rod-type dispersion fuel. Especially, thermal conductivity of fuel meat decreases during the irradiation mainly because of interaction layer formation at the interface between the U-Mo fuel particle and Al matrix. The thermal conductivity of the interaction layer is not as high as the Al matrix. The growth of interaction layer is interactively affected by the temperature of fuel because it is associated with a diffusion reaction which is a thermally activated process. It is difficult to estimate the temperature profile during irradiation test due to the interdependency of fuel temperature and thermal conductivity changed by interaction layer growth. In this study, fuel performance of rod-type U-Mo/Al dispersion fuels during irradiation tests were estimated by considering the effect of interaction layer growth on the thermal conductivity of fuel meat

Behaviour of rock-like oxide fuels under reactivity-initiated accident conditions

International Nuclear Information System (INIS)

Kazuyuki, Kusagaya; Takehiko, Nakamura; Makio, Yoshinaga; Hiroshi, Akie; Toshiyuki, Yamashita; Hiroshi, Uetsuka

2002-01-01

Pulse irradiation tests of three types of un-irradiated rock-like oxide (ROX) fuel - yttria-stabilised zirconia (YSZ) single phase, YSZ and spinel (MgAl 2 O 4 ) homogeneous mixture and particle-dispersed YSZ/spinel - were conducted in the Nuclear Safety Research Reactor to investigate the fuel behaviour under reactivity-initiated accident conditions. The ROX fuels failed at fuel volumetric enthalpies above 10 GJ/m 3 , which was comparable to that of un-irradiated UO 2 fuel. The failure mode of the ROX fuels, however, was quite different from that of the UO 2 fuel. The ROX fuels failed with fuel pellet melting and a part of the molten fuel was released out to the surrounding coolant water. In spite of the release, no significant mechanical energy generation due to fuel/coolant thermal interaction was observed in the tested enthalpy range below∼12 GJ/m 3 . The YSZ type and homogenous YSZ/spinel type ROX fuels failed by cladding burst when their temperatures peaked, while the particle-dispersed YSZ/spinel type ROX fuel seemed to have failed by cladding local melting. (author)

Kinetic parameters of a material test research reactor fueled with various low enriched uranium dispersion fuels

International Nuclear Information System (INIS)

Muhammad, Farhan; Majid, Asad

2009-01-01

The effects of using different low enriched uranium fuels, having same uranium density, on the kinetic parameters of a material test research reactor were studied. For this purpose, the original aluminide fuel (UAl x -Al) containing 4.40 gU/cm 3 of an MTR was replaced with silicide (U 3 Si-Al and U 3 Si 2 -Al) and oxide (U 3 O 8 -Al) dispersion fuels having the same uranium density as of the original fuel. Simulations were carried out to calculate prompt neutron generation time, effective delayed-neutron fraction, core excess reactivity and neutron flux spectrum. Nuclear reactor analysis codes including WIMS-D4 and CITATION were used to carry out these calculations. It was observed that both the silicide fuels had the same prompt neutron generation time 0.02% more than that of the original aluminide fuel, while the oxide fuel had a prompt neutron generation time 0.05% less than that of the original aluminide fuel. The effective delayed-neutron fraction decreased for all the fuels; the decrease was maximum at 0.06% for U 3 Si 2 -Al followed by 0.03% for U 3 Si-Al, and 0.01% for U 3 O 8 -Al fuel. The U 3 O 8 -Al fueled reactor gave the maximum ρ excess at BOL which was 21.67% more than the original fuel followed by U 3 Si-Al which was 2.55% more, while that of U 3 Si 2 -Al was 2.50% more than the original UAl x -Al fuel. The neutron flux of all the fuels was more thermalized, than in the original fuel, in the active fuel region of the core. The thermalization was maximum for U 3 O 8 -Al followed by U 3 Si-Al and then U 3 Si 2 -Al fuel.

Dispersion fuel for nuclear research facilities

International Nuclear Information System (INIS)

Kushtym, A.V.; Belash, M.M.; Zigunov, V.V.; Slabospitska, O.O.; Zuyok, V.A.

2017-01-01

Designs and process flow sheets for production of nuclear fuel rod elements and assemblies TVS-XD with dispersion composition UO_2+Al are presented. The results of fuel rod thermal calculation applied to Kharkiv subcritical assembly and Kyiv research reactor VVR-M, comparative characteristics of these fuel elements, the results of metallographic analyses and corrosion tests of fuel pellets are given in this paper

Fuel oil and dispersant toxicity to the Antarctic sea urchin (Sterechinus neumayeri).

Science.gov (United States)

Alexander, Frances J; King, Catherine K; Reichelt-Brushett, Amanda J; Harrison, Peter L

2017-06-01

The risk of a major marine fuel spill in Antarctic waters is increasing, yet there are currently no standard or suitable response methods under extreme Antarctic conditions. Fuel dispersants may present a possible solution; however, little data exist on the toxicity of dispersants or fuels to Antarctic species, thereby preventing informed management decisions. Larval development toxicity tests using 3 life history stages of the Antarctic sea urchin (Sterechinus neumayeri) were completed to assess the toxicity of physically dispersed, chemically dispersed, and dispersant-only water-accommodated fractions (WAFs) of an intermediate fuel oil (IFO 180, BP) and the chemical dispersant Slickgone NS (Dasic International). Despite much lower total petroleum hydrocarbon concentrations, physically dispersed fuels contained higher proportions of low-to-intermediate weight carbon compounds and were generally at least an order of magnitude more toxic than chemically dispersed fuels. Based on concentrations that caused 50% abnormality (EC50) values, the embryonic unhatched blastula life stage was the least affected by fuels and dispersants, whereas the larval 4-armed pluteus stage was the most sensitive. The present study is the first to investigate the possible implications of the use of fuel dispersants for fuel spill response in Antarctica. The results indicate that the use of a fuel dispersant did not increase the hydrocarbon toxicity of IFO 180 to the early life stages of Antarctic sea urchins, relative to physical dispersal. Environ Toxicol Chem 2017;36:1563-1571. © 2016 SETAC. © 2016 SETAC.

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...

DART model for irradiation-induced swelling of dispersion fuel elements including aluminum-fuel interaction

International Nuclear Information System (INIS)

Rest, J.; Hofman, G.L.

1997-01-01

The Dispersion Analysis Research Tool (DART) contains models for fission-gas-induced fuel swelling, interaction of fuel with the matrix aluminum, for the resultant reaction-product swelling, and for the calculation of the stress gradient within the fuel particle. The effects of an aluminide shell on fuel particle swelling are evaluated. Validation of the model is demonstrated by a comparison of DART calculations of fuel swelling of U 3 SiAl-Al and U 3 Si 2 -Al for various dispersion fuel element designs with the data

Performance evaluation of large U-Mo particle dispersed fuel irradiated in HANARO

International Nuclear Information System (INIS)

Ryu, Ho Jin; Park, Jong Man; Oh, Seok Jin; Jang, Se Jung; Yu, Byung Ok; Lee, Choong Seong; Seo, Chul Gyo; Chae, Hee Taek; Kim, Chang Kyu

2008-01-01

U-Mo/Al dispersion fuel is being developed as advanced fuel for research reactors. Irradiation behavior of U-Mo/Al dispersion fuel has been studied to evaluate its fuel performance. One of the performance limiting factors is a chemical interaction between the U-Mo particle and the Al matrix because the thermal conductivity of fuel meat is decreased with the interaction layer growth. In order to overcome the interaction problem, large-sized U-Mo particles were fabricated by controlling the centrifugal atomization conditions. The fuel performance behavior of U-Mo/Al dispersion fuel was estimated by using empirical models formulated based on the microstructural analyses of the post-irradiation examination (PIE) on U-Mo/Al dispersion fuel irradiated in HANARO reactor. Temperature histories of U-Mo/Al dispersion fuel during irradiation tests were estimated by considering the effect of an interaction layer growth on the thermal conductivity of the fuel meat. When the fuel performances of the dispersion fuel rods containing U-Mo particles with various sizes were compared, fuel temperature was decreased as the average U-Mo particle size was increases. It was found that the dispersion of a larger U-Mo particle was effective for mitigating the thermal degradation which is associated with an interaction layer growth. (author)

A Deformation Model of TRU Metal Dispersion Fuel Rod for HYPER

International Nuclear Information System (INIS)

Lee, Byoung Oon; Hwang, Woan; Park, Won S.

2002-01-01

Deformation analysis in fuel rod design is essential to assure adequate fuel performance and integrity under irradiation conditions. An in-reactor performance computer code for a dispersion fuel rod is being developed in the conceptual design stage of blanket fuel for HYPER. In this paper, a mechanistic deformation model was developed and the model was installed into the DIMAC program. The model was based on the elasto-plasticity theory and power-law creep theory. The preliminary deformation calculation results for (TRU-Zr)-Zr dispersion fuel predicted by DIMAC were compared with those of silicide dispersion fuel predicted by DIFAIR. It appeared that the deformation levels for (TRU-Zr)-Zr dispersion fuel were relatively higher than those of silicide fuel. Some experimental tests including in-pile and out-pile experiments are needed for verifying the predictive capability of the DIMAC code. An in-reactor performance analysis computer code for blanket fuel is being developed at the conceptual design stage of blanket fuel for HYPER. In this paper, a mechanistic deformation model was developed and the model was installed into the DIMAC program. The model was based on the elasto-plasticity theory and power-law creep theory. The preliminary deformation calculation results for (TRUZr)- Zr dispersion fuel predicted by DIMAC were compared with those of silicide dispersion fuel predicted by DIFAIR. It appears that the deformation by swelling within fuel meat is very large for both fuels, and the major deformation mechanism at cladding is creep. The swelling strain is almost constant within the fuel meat, and is assumed to be zero in the cladding made of HT9. It is estimated that the deformation levels for (TRU-Zr)-Zr dispersion fuel were relatively higher than those of silicide fuel, and the dispersion fuel performance may be limited by swelling. But the predicted volume change of the (TRU-Zr)-Zr dispersion fuel models is about 6.1% at 30 at.% burnup. The value of cladding

A general evaluation of the irradiation behaviour of dispersion fuels

International Nuclear Information System (INIS)

Hofman, G.L.

1995-01-01

The irradiation behaviour of aluminum-based dispersion fuels is evaluated with emphasis on metallurgical processes that control the dispersion behaviour. Phase transformations and microstructural changes resulting from fuel-matrix interactions and the effect of fissioning in fuel are discussed. (author)

Reactivity feedbacks of a material test research reactor fueled with various low enriched uranium dispersion fuels

International Nuclear Information System (INIS)

Muhammad, Farhan; Majid, Asad

2009-01-01

The reactivity feedbacks of a material test research reactor using various low enriched uranium fuels, having same uranium density were calculated. For this purpose, the original aluminide fuel (UAl x -Al) containing 4.40 gU/cm 3 of an MTR was replaced with silicide (U 3 Si-Al and U 3 Si 2 -Al) and oxide (U 3 O 8 -Al) dispersion fuels having the same uranium density as of the original fuel. Calculations were carried out to find the fuel temperature reactivity feedback, moderator temperature reactivity feedback, moderator density reactivity feedback and moderator void reactivity feedback. Nuclear reactor analysis codes including WIMS-D4 and CITATION were employed to carry out these calculations. It was observed that the magnitudes all the respective reactivity feedbacks from 38 deg. C to 50 deg. C and 100 deg. C, at the beginning of life, of all the fuels were very close to each other. The fuel temperature reactivity feedback of the U 3 O 8 -Al was about 2% more than the original UAl x -Al fuel. The magnitudes of the moderator temperature, moderator density and moderator void reactivity feedbacks of all the fuels, showed very minor variations from the original aluminide fuel.

A Prediction Study of Aluminum Alloy Oxidation of the Fuel Cladding in Jordan Research and Training Reactor

Energy Technology Data Exchange (ETDEWEB)

Tahk, Y. W.; Oh, J. Y.; Lee, B. H.; Seo, C. G.; Chae, H. T.; Yim, J. S. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2010-10-15

U{sub 3}Si{sub 2}-Al dispersion fuel with Al cladding will be used for Jordan Research and Training Reactor (JRTR). Aluminum alloy cladding experiences the oxidation layer growth on the surface during the reactor operation. The formation of oxides on the cladding affects fuel performance by increasing fuel temperature. According to the current JRTR fuel management scheme and operation strategy for 5 MW power, a fresh fuel is discharged after 900 effective full power days (EFPD) with 18 cycles of 50 days loading. For the proper prediction of the aluminum oxide thickness of fuel cladding during the long residence time, a reliable model is needed. In this work, several oxide thickness prediction models are compared with the measured data from in-pile test by RERTR program. Moreover, specific parametric studies and a preliminary prediction of the aluminum alloy oxidation using the latest model are performed for JRTR fuel

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.

Recovery of UMo alloy from UMo/Al dispersion fuel plates by dissolution

International Nuclear Information System (INIS)

Ren Meng; Li Jia; Liu Jinhong; Zhu Changgui

2011-01-01

Methods for dissolving UMo/Al dispersion fuel plates in the compounded mixed basic aqueous (NaOH and NaNO 3 ) are studied on laboratory scale. After removing the clad and the matrix of the substandard UMo/Al dispersion fuel elements, the U loss ratios are calculated and the granularity distributions of the recovered UMo alloy powder are analyzed by the metallurgical microscope. Besides, the phase structure and the composition of the recovered UMo alloy powder are analyzed by the XRD. The results indicate that as the concentration of NaOH increases, uranium loss ratio increases; but as the concentration of NaNO 3 increases, U loss ration increases firstly and then decreases subsequently; generally, the U recovery ratios are more than 99.3%. The granularity of recovered UMo powders are very small and most parts of γ-U have been oxidated to UO 2 . Therefore, further study is required to determined whether the recovered UMo alloy could be returned to the product line. (authors)

Monte-Carlo simulation of dispersion fuel meat structure

International Nuclear Information System (INIS)

Xing Zhonghu; Ying Shihao

2003-01-01

Under the irradiation conditions in research reactors, the inter-diffusion occurs at the fuel particle and matrix interfaces of U 3 Si 2 -Al dispersion fuel. Because of the inter-diffusion reaction, the U 3 Al 7 Si 2 layer is formed around each U 3 Si 2 particle. The layer thickness grows up with irradiation duration and fission density. The formation of resultant layer causes the consumption of U 3 Si 2 fuel and aluminum matrix. This process leads to the evolution of geometrical structure of fuel meat. According to the stochastic locations of particles in dispersion, the authors developed a simulation method for the evolution of the fuel meat structure by utilizing Monte-Carlo method. Every particle is characterized by its diameter and location. The parameters of meat structure include particle size distribution, as-fabricated fuel volume fraction, resultant layer thickness, layer volume fraction, U 3 Si 2 fuel volume fraction, aluminum volume fraction, contiguity probability and inter-linkage fraction of particles. Particularly for the dispersion with as-fabricated fuel volume fraction of 43% and particle sizes in a well-defined normal distribution, more than 13000 sampling particles are simulated in the meat volume of 6 mm x 6 mm x 0.5 mm. The meat structure parameters are calculated as functions of layer thickness in the range from 0-16 μm. (authors)

Fabrication procedures for manufacturing high uranium concentration dispersion fuel elements

International Nuclear Information System (INIS)

Souza, Jose Antonio Batista de

2011-01-01

IPEN-CNEN/SP developed the technology to produce the dispersion type fuel elements for research reactors and made it available for routine production. Today, the fuel produced in IPEN-CNEN/SP is limited to the uranium concentration of 3.0 gU/cm 3 for U 3 Si 2 -Al dispersion-based and 2.3 gU/cm 3 for U 3 O 8 -Al dispersion. The increase of uranium concentration in fuel plates enables the reactivity of the reactor core reactivity to be higher and extends the fuel life. Concerning technology, it is possible to increase the uranium concentration in the fuel meat up to the limit of 4.8 gU/cm 3 in U 3 Si 2 -Al dispersion and 3.2 gU/cm 3 U 3 O 8 -Al dispersion. These dispersions are well qualified worldwide. This work aims to develop the manufacturing process of both fuel meats with high uranium concentrations, by redefining the manufacturing procedures currently adopted in the Nuclear Fuel Center of IPEN-CNEN/SP. Based on the results, it was concluded that to achieve the desired concentration, it is necessary to make some changes in the established procedures, such as in the particle size of the fuel powder and in the feeding process inside the matrix, before briquette pressing. These studies have also shown that the fuel plates, with a high concentration of U 3 Si 2 -Al, met the used specifications. On the other hand, the appearance of the microstructure obtained from U 3 O 8 -Al dispersion fuel plates with 3.2 gU/cm 3 showed to be unsatisfactory, due to the considerably significant porosity observed. The developed fabrication procedure was applied to U 3 Si 2 production at 4.8 gU/cm 3 , with enriched uranium. The produced plates were used to assemble the fuel element IEA-228, which was irradiated in order to check its performance in the IEA-R1 reactor at IPEN-CNEN/SP. These new fuels have potential to be used in the new Brazilian Multipurpose Reactor - RMB. (author)

Dissolving method for nuclear fuel oxide

International Nuclear Information System (INIS)

Tomiyasu, Hiroshi; Kataoka, Makoto; Asano, Yuichiro; Hasegawa, Shin-ichi; Takashima, Yoichi; Ikeda, Yasuhisa.

1996-01-01

In a method of dissolving oxides of nuclear fuels in an aqueous acid solution, the oxides of the nuclear fuels are dissolved in a state where an oxidizing agent other than the acid is present together in the aqueous acid solution. If chlorate ions (ClO 3 - ) are present together in the aqueous acid solution, the chlorate ions act as a strong oxidizing agent and dissolve nuclear fuels such as UO 2 by oxidation. In addition, a Ce compound which generates Ce(IV) by oxidation is added to the aqueous acid solution, and an ozone (O 3 ) gas is blown thereto to dissolve the oxides of nuclear fuels. Further, the oxides of nuclear fuels are oxidized in a state where ClO 2 is present together in the aqueous acid solution to dissolve the oxides of nuclear fuels. Since oxides of the nuclear fuels are dissolved in a state where the oxidizing agent is present together as described above, the oxides of nuclear fuels can be dissolved even at a room temperature, thereby enabling to use a material such as polytetrafluoroethylene and to dissolve the oxides of nuclear fuels at a reduced cost for dissolution. (T.M.)

Modeling solid-fuel dispersal during slow loss-of-flow-type transients

International Nuclear Information System (INIS)

DiMelfi, R.J.; Fenske, G.R.

1981-01-01

The dispersal, under certain accident conditions, of solid particles of fast-reactor fuel is examined in this paper. In particular, we explore the possibility that solid-fuel fragmentation and dispersal can be driven by expanding fission gas, during a slow LOF-type accident. The consequences of fragmentation are studied in terms of the size and speed of dispersed particles, and the overall quantity of fuel moved. (orig.)

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.)

Development of accident tolerant FeCrAl-ODS steels utilizing Ce-oxide particles dispersion

Science.gov (United States)

Shibata, Hiroki; Ukai, Shigeharu; Oono, Naoko H.; Sakamoto, Kan; Hirai, Mutsumi

2018-04-01

FeCrAl-ODS ferritic steels with Ce-oxide dispersion instead of Y-oxide were produced for the accident tolerant fuel cladding of the light water reactor. Excess oxygen (Ex.O) was added to improve the mechanical property. The tensile strength at Ex.O = 0 is around 200 MPa at 700 °C, mainly owing to dispersed Ce2O3 particles in less than 10 nm size. The formation of the fine Ce2O3 particles is dominated by a coherent interface with ferritic matrix. With increasing Ex.O, an increased of number density of coarser Ce-Al type oxide particles over 10 nm size is responsible for the improvement of the tensile strength. Change of the type of oxide particle, CeO2, Ce2O3, CeAlO3, Al2O3, in FeCrAl-ODS steel was thermodynamically analyzed as a parameter of Ex.O.

A method for determining an effective porosity correction factor for thermal conductivity in fast reactor uranium-plutonium oxide fuel pellets

International Nuclear Information System (INIS)

Inoue, Masaki; Abe, Kazuyuki; Sato, Isamu

2000-01-01

A reliable method has been developed for determining an effective porosity correction factor for calculating a realistic thermal conductivity for fast reactor uranium-plutonium (mixed) oxide fuel pellets. By using image analysis of the ceramographs of transverse sections of mixed-oxide fuel pellets, the fuel morphology could be classified into two basic types. One is a 'two-phase' type that consists of small pores dispersed in the fuel matrix. The other is a 'three-phase' type that has large pores in addition to the small pores dispersed in the fuel matrix. The pore sizes are divided into two categories, large and small, at the 30 μm area equivalent diameter. These classifications lead to an equation for calculating an effective porosity correction factor by accounting for the small and large pore volume fractions and coefficients. This new analytical method for determining the effective porosity correction factor for calculating the realistic thermal conductivity of mixed-oxide fuel was also experimentally confirmed for high-, medium- and low-density fuel pellets

Development of U-Mo/Al dispersion fuel for research reactors

Energy Technology Data Exchange (ETDEWEB)

Park, Jong Man; Ryu, Ho Jin; Yang, Jae Ho; Jeong, Yong Jin; Lee, Yoon Sang [Korea Atomic Energy Research Inst., Research Reactor Fuel Development Division, Daejeon (Korea, Republic of)

2012-03-15

Currently, the KOMO-5 irradiation test for full size U-Mo/Al dispersion fuel rods has been underway since May 23, 2011. The purpose of the KOMO-5 test includes an investigation of the irradiation behaviors of silicide or nitride coated U-7Mo/Al(-Si) dispersion fuels and the effects of pre-formed interaction layers on U-Mo particles. It is expected that the irradiation test will be finished after attaining 60 at% U-235 burnup in May 2012, and the first PIE results of the KOMO-5 will be obtained in September 2012. In addition, an international cooperation program on the qualification of U-Mo dispersion fuels for small and medium size research reactors is going to be proposed in cooperation with the IAEA. Conversion from silicide fuel to U-Mo fuel will increase the cycle length with a smaller number of fuel assemblies and allow more flexible back-end options for spent fuel due to of the reprocessibility of U-Mo. (author)

Fabrication procedures for manufacturing high uranium concentration dispersion fuel elements

International Nuclear Information System (INIS)

Souza, J.A.B.; Durazzo, M.

2010-01-01

IPEN developed and made available for routine production the technology for manufacturing dispersion type fuel elements for use in research reactors. However, the fuel produced at IPEN is limited to the uranium concentration of 3.0 gU/cm 3 by using the U 3 Si 2 -Al dispersion. Increasing the uranium concentration of the fuel is interesting by the possibility of increasing the reactor core reactivity and lifetime of the fuel. It is possible to increase the concentration of uranium in the fuel up to the technological limit of 4.8 gU/cm 3 for the U 3 Si 2 -Al dispersion, which is well placed around the world. This new fuel will be applicable in the new Brazilian-Multipurpose Reactor RMB. This study aimed to develop the manufacturing process of high uranium concentration fuel, redefining the procedures currently used in the manufacture of IPEN. This paper describes the main procedures adjustments that will be necessary. (author)

Fabrication procedures for manufacturing high uranium concentration dispersion fuel elements

Energy Technology Data Exchange (ETDEWEB)

Souza, Jose Antonio Batista de; Durazzo, Michelangelo, E-mail: jasouza@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2011-07-01

IPEN developed and made available for routine production the technology for manufacturing dispersion type fuel elements for use in research reactors. However, the fuel produced at IPEN is limited to the uranium concentration of 3.0 g U/c m3 by using the U{sub 3}Si{sub 2}-Al dispersion. Increasing the uranium concentration of the fuel is interesting by the possibility of increasing the reactor core reactivity and lifetime of the fuel. It is possible to increase the concentration of uranium in the fuel up to the technological limit of 4.8 g U/c m3 for the U{sub 3}Si{sub 2}-Al dispersion, which is well placed around the world. This new fuel will be applicable in the new Brazilian- Multipurpose Reactor RMB. This study aimed to develop the manufacturing process of high uranium concentration fuel, redefining the procedures currently used in the manufacture of IPEN. This paper describes the main procedures adjustments that will be necessary. (author)

Safety aspects of LWR fuel reprocessing and mixed oxide fuel fabrication plants

International Nuclear Information System (INIS)

Fischer, M.; Leichsenring, C.H.; Herrmann, G.W.; Schueller, W.; Hagenberg, W.; Stoll, W.

1977-01-01

The paper is focused on the safety and the control of the consequences of credible accidents in LWR fuel reprocessing plants and in mixed oxide fuel fabrication plants. Each of these plants serve for many power reactor (about 50.000 Mwel) thus the contribution to the overall risk of nuclear energy is correspondingly low. Because of basic functional differences between reprocessing plants, fuel fabrication plants and nuclear power reactors, the structure and safety systems of these plants are different in many respects. The most important differences that influence safety systems are: (1) Both fuel reprocessing and fabrication plants do not have the high system pressure that is associated with power reactors. (2) A considerable amount of the radioactivity of the fuel, which is in the form of short-lived radionuclides has decayed. Therefore, fuel reprocessing plants and mixed oxide fuel fabrication plants are designed with multiple confinement barriers for control of radioactive materials, but do not require the high-pressure containment systems that are used in LWR plants. The consequences of accidents which may lead to the dispersion of radioactive materials such as chemical explosions, nuclear excursions, fires and failure of cooling systems are considered. A reasonable high reliability of the multiple confinement approach can be assured by design. In fuel reprocessing plants, forced cooling is necessary only in systems where fission products are accumulated. However, the control of radioactive materials can be maintained during normal operation and during the above mentioned accidents, if the dissolver off-gas and vessel off-gas treatment systems provide for effective removal of radioactive iodine, radioactive particulates, nitrogen oxides, tritium and krypton 85. In addition, the following incidents in the dissolver off-gas system itself must be controlled: failures of iodine filters, hydrogen explosion in O 2 - and NOsub(x)-reduction component, decomposition of

Homogeneous forming technology of composite materials and its application to dispersion nuclear fuel

International Nuclear Information System (INIS)

Hong, Soon Hyun; Ryu, Ho Jin; Sohn, Woong Hee; Kim, Chang Kyu

1997-01-01

Powder metallurgy processing technique of metal matrix composites is reviewed and its application to process homogeneous dispersion nuclear fuel is considered. The homogeneous mixing of reinforcement with matrix powders is very important step to process metal matrix composites. The reinforcement with matrix powders is very important step to process metal matrix composites. The reinforcement can be ceramic particles, whiskers or chopped fibers having high strength and high modulus. The blended powders are consolidated into billets and followed by various deformation processing, such as extrusion, forging, rolling or spinning into final usable shapes. Dispersion nuclear fuel is a class of metal matrix composite consisted of dispersed U-compound fuel particles and metallic matrix. Dispersion nuclear fuel is fabricated by powder metallurgy process such as hot pressing followed by hot extrusion, which is similar to that of SiC/Al metal matrix composite. The fabrication of homogeneous dispersion nuclear fuel is very difficult mainly due to the inhomogeneous mixing characteristics of the powders from quite different densities between uranium alloy powders and aluminum powders. In order to develop homogeneous dispersion nuclear fuel, it is important to investigate the effect of powder characteristics and mixing techniques on homogeneity of dispersion nuclear fuel. An new quantitative analysis technique of homogeneity is needed to be developed for more accurate analysis of homogeneity in dispersion nuclear fuel. (author). 28 refs., 7 figs., 1tab

Effectiveness of paper-structured catalyst for the operation of biodiesel-fueled solid oxide fuel cell

Science.gov (United States)

Quang-Tuyen, Tran; Kaida, Taku; Sakamoto, Mio; Sasaki, Kazunari; Shiratori, Yusuke

2015-06-01

Mg/Al-hydrotalcite (HDT)-dispersed paper-structured catalyst (PSC) was prepared by a simple paper-making process. The PSC exhibited excellent catalytic activity for the steam reforming of model biodiesel fuel (BDF), pure oleic acid methyl ester (oleic-FAME, C19H36O2) which is a mono-unsaturated component of practical BDFs. The PSC exhibited fuel conversion comparable to a pelletized catalyst material, here, conventional Ni-zirconia cermet anode for solid oxide fuel cell (SOFC) with less than one-hundredth Ni weight. Performance of electrolyte-supported cell connected with the PSC was evaluated in the feed of oleic-FAME, and stable operation was achieved. After 60 h test, coking was not observed in both SOFC anode and PSC.

Copper based anodes for bio-ethanol fueled low-temperature solid oxide fuel cells

Energy Technology Data Exchange (ETDEWEB)

Kondakindi, R.R.; Karan, K. [Queen' s Univ., Kingston, ON (Canada)

2003-07-01

Laboratory studies have been conducted to develop a low-temperature solid oxide fuel cell (SOFC) fueled by bio-ethanol. SOFCs are considered to be a potential source for clean and efficient electricity. The use of bio-ethanol to power the SOFC contributes even further to reducing CO{sub 2} emissions. The main barrier towards the development of the proposed SOFC is the identification of a suitable anode catalyst that prevents coking during electro-oxidation of ethanol while yielding good electrical performance. Copper was selected as the catalyst for this study. Composite anodes consisting of copper catalysts and gadolinium-doped ceria (GDC) electrolytes were prepared using screen printing of GDC and copper oxide on dense GDC electrolytes and by wet impregnation of copper nitrate in porous GDC electrolytes followed by calcination and sintering. The electrical conductivity of the prepared anodes was characterized to determine the percolation threshold. Temperature-programmed reduction and the Brunner Emmett Teller (BET) methods were used to quantify the catalyst dispersion and surface area. Electrochemical performance of the single-cell SOFC with a hydrogen-air system was used to assess the catalytic activities. Electrochemical Impedance Spectroscopy was used to probe the electrode kinetics.

EMISSION REDUCTION FROM A DIESEL ENGINE FUELED BY CERIUM OXIDE NANO-ADDITIVES USING SCR WITH DIFFERENT METAL OXIDES COATED CATALYTIC CONVERTER

Directory of Open Access Journals (Sweden)

B. JOTHI THIRUMAL

2015-11-01

Full Text Available This paper reports the results of experimental investigations on the influence of the addition of cerium oxide in nanoparticle form on the major physiochemical properties and the performance of diesel. The fuel is modified by dispersing the catalytic nanoparticle by ultrasonic agitation. The physiochemical properties of sole diesel fuel and modified fuel are tested with ASTM standard procedures. The effects of the additive nanoparticles on the individual fuel properties, the engine performance, and emissions are studied, and the dosing level of the additive is optimized. Cerium oxide acts as an oxygen-donating catalyst and provides oxygen for the oxidation of CO during combustion. The active energy of cerium oxide acts to burn off carbon deposits within the engine cylinder at the wall temperature and prevents the deposition of non-polar compounds on the cylinder wall which results in reduction in HC emission by 56.5%. Furthermore, a low-cost metal oxide coated SCR (selective catalyst reduction, using urea as a reducing agent, along with different types of CC (catalytic converter, has been implemented in the exhaust pipe to reduce NOx. It was observed that a reduction in NOx emission is 50–60%. The tests revealed that cerium oxide nanoparticles can be used as an additive in diesel to improve complete combustion of the fuel and reduce the exhaust emissions significantly.

Modeling a failure criterion for U-Mo/Al dispersion fuel

Science.gov (United States)

Oh, Jae-Yong; Kim, Yeon Soo; Tahk, Young-Wook; Kim, Hyun-Jung; Kong, Eui-Hyun; Yim, Jeong-Sik

2016-05-01

The breakaway swelling in U-Mo/Al dispersion fuel is known to be caused by large pore formation enhanced by interaction layer (IL) growth between fuel particles and Al matrix. In this study, a critical IL thickness was defined as a criterion for the formation of a large pore in U-Mo/Al dispersion fuel. Specifically, the critical IL thickness is given when two neighboring fuel particles come into contact with each other in the developed IL. The model was verified using the irradiation data from the RERTR tests and KOMO-4 test. The model application to full-sized sample irradiations such as IRISs, FUTURE, E-FUTURE, and AFIP-1 tests resulted in conservative predictions. The parametric study revealed that the fuel particle size and the homogeneity of the fuel particle distribution are influential for fuel performance.

Modeling a failure criterion for U–Mo/Al dispersion fuel

Energy Technology Data Exchange (ETDEWEB)

Oh, Jae-Yong, E-mail: tylor@kaeri.re.kr [Korea Atomic Energy Research Institute, 111, Daedeok-Daero 989 Beon-Gil, Yuseong-Gu, Daejeon 305-353 (Korea, Republic of); Kim, Yeon Soo [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Tahk, Young-Wook; Kim, Hyun-Jung; Kong, Eui-Hyun; Yim, Jeong-Sik [Korea Atomic Energy Research Institute, 111, Daedeok-Daero 989 Beon-Gil, Yuseong-Gu, Daejeon 305-353 (Korea, Republic of)

2016-05-15

The breakaway swelling in U–Mo/Al dispersion fuel is known to be caused by large pore formation enhanced by interaction layer (IL) growth between fuel particles and Al matrix. In this study, a critical IL thickness was defined as a criterion for the formation of a large pore in U–Mo/Al dispersion fuel. Specifically, the critical IL thickness is given when two neighboring fuel particles come into contact with each other in the developed IL. The model was verified using the irradiation data from the RERTR tests and KOMO-4 test. The model application to full-sized sample irradiations such as IRISs, FUTURE, E-FUTURE, and AFIP-1 tests resulted in conservative predictions. The parametric study revealed that the fuel particle size and the homogeneity of the fuel particle distribution are influential for fuel performance.

A complementary and synergistic effect of Fe-Zn binary metal oxide in the process of high-temperature fuel gas desulfurization

Institute of Scientific and Technical Information of China (English)

翁斯灏; 吴幼青

1996-01-01

57Fe Mossbauer spectroscopy was used to investigate the evolution of Fe-Zn binary metal oxide sorbent in the process of high-temperature fuel gas desulfurization. The results of phase analyses show that Fe-Zn binary metal oxide sorbent is rapidly reduced in hot fuel gas and decomposed to new phases of highly dispersed microcrystalline elemental iron and zinc oxide, both of which become the active desulfurization constituents. A complementary and synergistic effect between active iron acting as a high sulfur capacity constituent and active zinc oxide acting as a deep refining desulfurization constituent exists in this type of sorbent for hot fuel gas desulfurization.

Development of technology of high density LEU dispersion fuel fabrication

International Nuclear Information System (INIS)

Wiencek, T.; Totev, T.

2007-01-01

Advanced Materials Fabrication Facilities at Argonne National Laboratory have been involved in development of LEU dispersion fuel for research and test reactors from the beginning of RERTR program. This paper presents development of technology of high density LEU dispersion fuel fabrication for full size plate type fuel elements. A brief description of Advanced Materials Fabrication Facilities where development of the technology was carried out is given. A flow diagram of the manufacturing process is presented. U-Mo powder was manufactured by the rotating electrode process. The atomization produced a U-Mo alloy powder with a relatively uniform size distribution and a nearly spherical shape. Test plates were fabricated using tungsten and depleted U-7 wt.% Mo alloy, 4043 Al and Al-2 wt% Si matrices with Al 6061 aluminum alloy for the cladding. During the development of the technology of manufacturing of full size high density LEU dispersion fuel plates special attention was paid to meet the required homogeneity, bonding, dimensions, fuel out of zone and other mechanical characteristics of the plates.

Calculation simulation of equivalent irradiation swelling for dispersion nuclear fuel

International Nuclear Information System (INIS)

Cai Wei; Zhao Yunmei; Gong Xin; Ding Shurong; Huo Yongzhong

2015-01-01

The dispersion nuclear fuel was regarded as a kind of special particle composites. Assuming that the fuel particles are periodically distributed in the dispersion nuclear fuel meat, the finite element model to calculate its equivalent irradiation swelling was developed with the method of computational micro-mechanics. Considering irradiation swelling in the fuel particles and the irradiation hardening effect in the metal matrix, the stress update algorithms were established respectively for the fuel particles and metal matrix. The corresponding user subroutines were programmed, and the finite element simulation of equivalent irradiation swelling for the fuel meat was performed in Abaqus. The effects of the particle size and volume fraction on the equivalent irradiation swelling were investigated, and the fitting formula of equivalent irradiation swelling was obtained. The results indicate that the main factors to influence equivalent irradiation swelling of the fuel meat are the irradiation swelling and volume fraction of fuel particles. (authors)

Properties of U3Si2-Al dispersion fuel element and its application

International Nuclear Information System (INIS)

Yin Changgeng

2001-01-01

The properties of U 3 Si 2 fuel and U 3 Si 2 -Al dispersion fuel element are introduced, which include U-loading; the banding quality, U-homogeneity and 'dog-bone' phenomenon, the minimum thickness of cladding and the corrosion performances. The fabrication technique of fuel elements, NDT for fuel plates, assemble technique of fuel elements and the application of U 3 Si 2 -Al dispersion fuel elements in the world are introduced

Requirements for materials of dispersion fuel elements

International Nuclear Information System (INIS)

Samojlov, A.G.; Kashtanov, A.I.; Volkov, V.S.

1982-01-01

Requirements for materials of dispersion fuel elements are considered. The necessity of structural and fissile materials compatibility at maximum permissible operation temperatures and temperatures arising in a fuel element during manufacture is pointed out. The fuel element structural material must be ductile, possess high mechanical strength minimum neutron absorption cross section, sufficient heat conductivity, good corrosion resistance in a coolant and radiation resistance. The fissile material must have high fissile isotope concentration, radiation resistance, high thermal conductivity, certain porosity high melting temperature must not change the composition under irradiation

Durable fuel electrode

DEFF Research Database (Denmark)

2017-01-01

the composite. The invention also relates to the use of the composite as a fuel electrode, solid oxide fuel cell, and/or solid oxide electrolyser. The invention discloses a composite for an electrode, comprising a three-dimensional network of dispersed metal particles, stabilised zirconia particles and pores...

Mixed oxide fuel pellet and manufacturing method thereof

International Nuclear Information System (INIS)

Yuda, Ryoichi; Ito, Ken-ichi; Masuda, Hiroshi.

1993-01-01

In a method of manufacturing nuclear fuel pellets which comprises compression molding a mixed oxide powder containing UO 2 and PuO 2 followed by sintering, a sintering agent having a composition comprising about 40 to 80 wt% of SiO 2 and the balance of Al 2 O 3 is mixed to a mixed oxide at a ratio of about 40ppm to about 0.5 wt% based on the total amount of the mixed oxide and the sintering agent, to prepare a mixture. The mixture is molded into a compression product and then sintered at a weakly acidic atmosphere at a temperature of about 1500degC to 1800degC. With such procedures, the sintering agent forms an eutectic product of a single liquid phase, PuO 2 is dispersed over the entire region of the pellet by way of the liquid phase, formation of a solid solution phase is promoted to annihilate a free PuO 2 phase. Further, growth of crystal grains is promoted. Accordingly, since the MOX fuel pellets prepared according to the present invention have a uniform solid solution state, and no free PuO 2 phase remains, increase of FP gas emission due to local nuclear fission of Pu can be avoided. (T.M.)

Modelling of U-Mo/Al Dispersion fuel fission induced swelling and creep

Energy Technology Data Exchange (ETDEWEB)

Jeong, Gwan Yoon; Sohn, Dong Seong [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of); Kim, Yeon Soo [Argonne National Laboratory, Argonne (United States)

2014-05-15

In a Dispersion fuel which U-Mo particles are dispersed in Al metal matrix, a similar phenomenon forming a bulge region was observed but it is difficult to quantify and construct a model for explaining creep and swelling because of its complex microstructure change during irradiation including interaction layer (IL) and porosity formation. In a Dispersion fuel meat, fission product induces fuel particles swelling and it has to be accommodated by the deformation of the Al matrix and newly formed IL during irradiation. Then, it is reasonable that stress from fuel swelling in the complex structure should be relaxed by local adjustments of particles, Al matrix, and IL. For analysis of U-Mo/Al Dispersion fuel creep, the creep of U-Mo particle, Al matrix, and IL should be considered. Moreover, not only fuel particle swelling and IL growth, but also fuel and Al matrix consumptions due to IL formation are accounted in terms of their volume fraction changes during irradiation. In this work, fuel particles, Al matrix and IL are treated in a way of homogenized constituents: Fuel particles, Al matrix and IL consist of an equivalent meat during irradiation. Meat volume swelling of two representative plates was measured: One (Plate A) was a pure Al matrix with 6g/cc uranium loading, the other (Plate B) a silicon added Al matrix with 8g/cc uranium loading. The meat swelling of calculated as a function of burnup. The meat swelling of calculation and measurement was compared and the creep rate coefficients for Al and IL were estimated by repetitions. Based on assumption that only the continuous phase of Al-IL combined matrix accommodated the stress from fuel particle swelling and it was allowed to have creep deformation, the homogenization modeling was performed. The meat swelling of two U-Mo/Al Dispersion fuel plates was modeled by using homogenization model.

Modelling of U-Mo/Al Dispersion fuel fission induced swelling and creep

International Nuclear Information System (INIS)

Jeong, Gwan Yoon; Sohn, Dong Seong; Kim, Yeon Soo

2014-01-01

In a Dispersion fuel which U-Mo particles are dispersed in Al metal matrix, a similar phenomenon forming a bulge region was observed but it is difficult to quantify and construct a model for explaining creep and swelling because of its complex microstructure change during irradiation including interaction layer (IL) and porosity formation. In a Dispersion fuel meat, fission product induces fuel particles swelling and it has to be accommodated by the deformation of the Al matrix and newly formed IL during irradiation. Then, it is reasonable that stress from fuel swelling in the complex structure should be relaxed by local adjustments of particles, Al matrix, and IL. For analysis of U-Mo/Al Dispersion fuel creep, the creep of U-Mo particle, Al matrix, and IL should be considered. Moreover, not only fuel particle swelling and IL growth, but also fuel and Al matrix consumptions due to IL formation are accounted in terms of their volume fraction changes during irradiation. In this work, fuel particles, Al matrix and IL are treated in a way of homogenized constituents: Fuel particles, Al matrix and IL consist of an equivalent meat during irradiation. Meat volume swelling of two representative plates was measured: One (Plate A) was a pure Al matrix with 6g/cc uranium loading, the other (Plate B) a silicon added Al matrix with 8g/cc uranium loading. The meat swelling of calculated as a function of burnup. The meat swelling of calculation and measurement was compared and the creep rate coefficients for Al and IL were estimated by repetitions. Based on assumption that only the continuous phase of Al-IL combined matrix accommodated the stress from fuel particle swelling and it was allowed to have creep deformation, the homogenization modeling was performed. The meat swelling of two U-Mo/Al Dispersion fuel plates was modeled by using homogenization model

Heating subsurface formations by oxidizing fuel on a fuel carrier

Science.gov (United States)

Costello, Michael; Vinegar, Harold J.

2012-10-02

A method of heating a portion of a subsurface formation includes drawing fuel on a fuel carrier through an opening formed in the formation. Oxidant is supplied to the fuel at one or more locations in the opening. The fuel is combusted with the oxidant to provide heat to the formation.

Mechanosynthesis of A Ferritic ODS (Oxide Dispersion Strengthened) Steel Containing 14% Chromium and Its Characterization

Science.gov (United States)

Rivai, A. K.; Dimyati, A.; Adi, W. A.

2017-05-01

One of the advanced materials for application at high temperatures which is aggressively developed in the world is ODS (Oxide Dispersion strengthened) steel. ODS ferritic steels are one of the candidate materials for future nuclear reactors in the world (Generation IV reactors) because it is able to be used in the reactor above 600 °C. ODS ferritic steels have also been developed for the interconnect material of SOFC (Solid Oxide Fuel Cell) which will be exposed to about 800 °C of temperature. The steel is strengthened by dispersing homogeneously of oxide particles (ceramic) in nano-meter sized in the matrix of the steel. Synthesis of a ferritic ODS steel by dispersion of nano-particles of yttrium oxide (yttria: Y2O3) as the dispersion particles, and containing high-chromium i.e. 14% has been conducted. Synthesis of the ODS steels was done mechanically (mechanosynthesis) using HEM (High Energy ball Milling) technique for 40 and 100 hours. The resulted samples were characterized using SEM-EDS (Scanning Electron Microscope-Energy Dispersive Spectroscope), and XRD (X-ray diffraction) to analyze the microstructure characteristics. The results showed that the crystal grains of the sample with 100 hours milling time was much smaller than the sample with 40 hours milling time, and some amount of alloy was formed during the milling process even for 40 hours milling time. Furthermore, the structure analysis revealed that some amount of iron atom substituted by a slight amount of chromium atom as a solid solution. The quantitative analysis showed that the phase mostly consisted of FeCr solid-solution with the structure was BCC (body-centered cubic).

Detailed analysis of uranium silicide dispersion fuel swelling

International Nuclear Information System (INIS)

Hofmann, G.L.; Ryu, Woo-Seog

1991-01-01

Swelling of U 3 Si and U 3 Si 2 is analyzed. The growth of fission gas bubbles appears to be affected by fission rate, fuel loading, and micro structural change taking place in the fuel compounds during irradiation. Several mechanisms are explored to explain the observations. The present work is aimed at a better understanding of the basic swelling phenomenon in order to accurately model irradiation behavior of uranium silicide dispersion fuel. (orig.)

Increased electrochemical properties of ruthenium oxide and graphene/ruthenium oxide hybrid dispersed by polyvinylpyrrolidone

International Nuclear Information System (INIS)

Chen, Yao; Zhang, Xiong; Zhang, Dacheng; Ma, Yanwei

2012-01-01

Highlights: ► A good dispersion of RuO 2 and graphene/RuO 2 is obtained by polyvinylpyrrolidone. ► PVP as a dispersant also can prevent the formation of metal Ru in graphene/RuO 2 . ► The max capacitances of the hybrid and RuO 2 reach 435 and 597 F g −1 at 0.2 A g −1 . ► The hybrid shows the best rate capability of 39% at 50 A g −1 . - Abstract: Ruthenium oxide has been prepared by a sol–gel method. Polyvinylpyrrolidone (PVP) as an excellent polymeric dispersant is adopted to prevent aggregation of ruthenium oxide. In order to enhance the rate capability of ruthenium oxide, graphene with residual oxygen functional groups as a 2D support has been merged into ruthenium oxide. These oxygen functional groups not only favor to form stable few layers of graphene colloids, but also offer the sites to anchor ruthenium oxide nanoparticles. X-ray diffraction infers that PVP can also hinder the partial formation of Ru by blocking the direct contact between the Ru 3+ and the graphene in the sol–gel synthesis of the hybrids. The ruthenium oxide and the graphene/ruthenium oxide hybrids dispersed by PVP have superior electrochemical properties due to good dispersing and protecting ability of PVP. Especially, the hybrids using PVP exhibit the best rate capability, indicating that the composites possess an advanced structure of combining sheets and particles in nano-scale.

Morphology Control of the Electrode for Solid Oxide Fuel Cells by Using Nanoparticles

International Nuclear Information System (INIS)

Fukui, Takehisa; Ohara, Satoshi; Naito, Makio; Nogi, Kiyoshi

2001-01-01

LSM(La(Sr)MnO 3 )/YSZ(Y 2 O 3 stabilized ZrO 2 ) composite cathode for Solid Oxide Fuel Cells (SOFCs) was fabricated by using the composite particle consisting of well-dispersed nano-size grains of LSM and YSZ. The composite cathode had a porous structure as well as uniformly dispersed fine LSM and YSZ grains. Such unique morphology of the composite cathode led high electrochemical activity at 800 deg. C. It suggests that the intermediate temperature (less than 800 o C) operation of SOFCs will be achieved by using composite particles

Randomly dispersed particle fuel model in the PSG Monte Carlo neutron transport code

International Nuclear Information System (INIS)

Leppaenen, J.

2007-01-01

High-temperature gas-cooled reactor fuels are composed of thousands of microscopic fuel particles, randomly dispersed in a graphite matrix. The modelling of such geometry is complicated, especially using continuous-energy Monte Carlo codes, which are unable to apply any deterministic corrections in the calculation. This paper presents the geometry routine developed for modelling randomly dispersed particle fuels using the PSG Monte Carlo reactor physics code. The model is based on the delta-tracking method, and it takes into account the spatial self-shielding effects and the random dispersion of the fuel particles. The calculation routine is validated by comparing the results to reference MCNP4C calculations using uranium and plutonium based fuels. (authors)

Fuel-cycle cost comparisons with oxide and silicide fuels

International Nuclear Information System (INIS)

Matos, J.E.; Freese, K.E.

1982-01-01

This paper addresses fuel cycle cost comparisons for a generic 10 MW reactor with HEU aluminide fuel and with LEU oxide and silicide fuels in several fuel element geometries. The intention of this study is to provide a consistent assessment of various design options from a cost point of view. Fuel cycle cost benefits could result if a number of reactors were to utilize fuel elements with the same number or different numbers of the same standard fuel plate. Data are presented to quantify these potential cost benefits. This analysis shows that there are a number of fuel element designs using LEU oxide or silicide fuels that have either the same or lower total fuel cycle costs than the HEU design. Use of these fuels with the uranium densities considered requires that they are successfully demonstrated and licensed

Spent fuel sabotage test program, characterization of aerosol dispersal : interim final report

International Nuclear Information System (INIS)

Gregson, Michael Warren; Brockmann, John E.; Loiseau, Olivier; Klennert, Lindsay A.; Nolte, Oliver; Molecke, Martin Alan; Autrusson, Bruno A.; Koch, Wolfgang; Pretzsch, Gunter Guido; Brucher, Wenzel; Steyskal, Michele D.

2008-01-01

This multinational, multi-phase spent fuel sabotage test program is quantifying the aerosol particles produced when the products of a high energy density device (HEDD) interact with and explosively particulate test rodlets that contain pellets of either surrogate materials or actual spent fuel. This program provides source-term data that are relevant to plausible sabotage scenarios in relation to spent fuel transport and storage casks and associated risk assessments. We present details and significant results obtained from this program from 2001 through 2007. Measured aerosol results include: respirable fractions produced; amounts, nuclide content, and produced particle size distributions and morphology; measurements of volatile fission product species enhanced sorption--enrichment factors onto respirable particles; and, status on determination of the spent fuel ratio, SFR, needed for scaling studies. Emphasis is provided on recent Phase 3 tests using depleted uranium oxide pellets plus non-radioactive fission product dopants in surrogate spent fuel test rodlets, plus the latest surrogate cerium oxide results and aerosol laboratory supporting calibration work. The DUO 2 , CeO 2 , plus fission product dopant aerosol particle results are compared with available historical data. We also provide a status review on continuing preparations for the final Phase 4 in this program, tests using individual short rodlets containing actual spent fuel from U.S. PWR reactors, with both high- and lower-burnup fuel. The source-term data, aerosol results, and program design have been tailored to support and guide follow-on computer modeling of aerosol dispersal hazards and radiological consequence assessments. This spent fuel sabotage, aerosol test program was performed primarily at Sandia National Laboratories, with support provided by both the U.S. Department of Energy and the Nuclear Regulatory Commission. This program has significant input from, and is cooperatively supported and

Spent fuel sabotage test program, characterization of aerosol dispersal : interim final report.

Energy Technology Data Exchange (ETDEWEB)

Gregson, Michael Warren; Brockmann, John E.; Loiseau, Olivier (Institut de Radioprotection et de Surete Nucleaire, France); Klennert, Lindsay A.; Nolte, Oliver (Fraunhofer Institut fur Toxikologie und Experimentelle Medizin, Germany); Molecke, Martin Alan; Autrusson, Bruno A. (Institut de Radioprotection et de Surete Nucleaire, France); Koch, Wolfgang (Fraunhofer Institut fur Toxikologie und Experimentelle Medizin, Germany); Pretzsch, Gunter Guido (Gesellschaft fur Anlagen- und Reaktorsicherheit, Germany); Brucher, Wenzel (Gesellschaft fur Anlagen- und Reaktorsicherheit, Germany); Steyskal, Michele D.

2008-03-01

This multinational, multi-phase spent fuel sabotage test program is quantifying the aerosol particles produced when the products of a high energy density device (HEDD) interact with and explosively particulate test rodlets that contain pellets of either surrogate materials or actual spent fuel. This program provides source-term data that are relevant to plausible sabotage scenarios in relation to spent fuel transport and storage casks and associated risk assessments. We present details and significant results obtained from this program from 2001 through 2007. Measured aerosol results include: respirable fractions produced; amounts, nuclide content, and produced particle size distributions and morphology; measurements of volatile fission product species enhanced sorption--enrichment factors onto respirable particles; and, status on determination of the spent fuel ratio, SFR, needed for scaling studies. Emphasis is provided on recent Phase 3 tests using depleted uranium oxide pellets plus non-radioactive fission product dopants in surrogate spent fuel test rodlets, plus the latest surrogate cerium oxide results and aerosol laboratory supporting calibration work. The DUO{sub 2}, CeO{sub 2}, plus fission product dopant aerosol particle results are compared with available historical data. We also provide a status review on continuing preparations for the final Phase 4 in this program, tests using individual short rodlets containing actual spent fuel from U.S. PWR reactors, with both high- and lower-burnup fuel. The source-term data, aerosol results, and program design have been tailored to support and guide follow-on computer modeling of aerosol dispersal hazards and radiological consequence assessments. This spent fuel sabotage, aerosol test program was performed primarily at Sandia National Laboratories, with support provided by both the U.S. Department of Energy and the Nuclear Regulatory Commission. This program has significant input from, and is cooperatively

Dispersants in an organic medium: synthesis and physicochemical study of dispersants for fuels and lubricants; Dispersants en milieu organique: synthese et etude physicochimique de dispersants pour carburants et lubrifiants

Energy Technology Data Exchange (ETDEWEB)

Dubois-Clochard, M.C.

1998-11-19

Carbonaceous deposits coming from the fuel and the lubricant are known to form over time at critical locations in an engine. In general, the deposits have an adverse effect on four functional areas which are the fuel metering system, the intake system, the lubrication system and the combustion chambers. These deposits can degrade vehicle performance and drive-ability, reduce fuel economy, increase fuel consumption and pollutant emissions and may lead to the destruction of the engine. In order to remedy these problems, detergent-dispersant additives are used in fuels and lubricants to avoid or decrease deposit adhesion on metallic surfaces and prevent from deposit aggregation. These products are mainly polymer surfactants and in this work, poly-iso-butenyl-succinimide of different structures have been studied. Firstly, 'comb like' polymers have been synthesized. Then they have been compared to classical di-bloc additives in terms of performance and action mechanism. These additives are adsorbed from their hydrophilic polyamine part on the acidic functions of the carbon black surface chosen as an engine deposit model and on the aluminium oxide function of an aluminium powder chosen as an engine wall model. The adsorption increases with temperature on the two solids. Their affinity with the solid surface increases with the length of the hydrophilic part. In the same way, changing the di-bloc structure for a comb like one lead to a better adsorption. At low concentration, it has been shown that the adsorption phenomenon was irreversible, due to the polymer structure of the polar part. Depending on the space required by the hydrophilic part on the solid surface, a more of less dense monolayer is formed. At higher concentrations, an important increase of the adsorbed amount appears. This phenomenon is totally reversible showing that the interactions additive / additive are weak. The dispersing efficiency of a comb like structure is better than a di-bloc one as

The rheology of oxide dispersions and the role of concentrated electrolyte solutions

International Nuclear Information System (INIS)

Biggs, Simon; Tindley, Amy

2007-01-01

Stability control of particulate dispersions is critical to a wide range of industrial processes. In the UK nuclear industry, significant volumes of waste materials arising from the corrosion products of Magnox fuel rods currently require treatment and storage. The majority of this waste is present as aqueous dispersions of oxide particulates. Treatment of these dispersions will require a variety of unit operations including mobilisation, transport and solid- liquid separation. Typically these processes must operate across a narrow optimal range of pH and the dispersions are, almost without exception, found in complex electrolyte conditions of high overall concentration. Knowledge of the behaviour of oxides in various electrolyte conditions and over a large pH range is essential for the efficient design and control of any waste processing approach. The transport properties of particle dispersions are characterised by the rheological properties. It is well known that particle dispersion rheology is strongly influenced by particle-particle interaction forces, and that particle-particle interactions are strongly influenced by adsorbed ions on the particle surfaces. Here we correlate measurements of the shear yield stress and the particle zeta potentials to provide insight as to the role of ions in moderating particle interactions. The zeta potential of model TiO 2 suspensions were determined (Colloidal Dynamics Zeta Probe) over a range of pH for a series of alkali metal halides and quaternary ammonium halides at a range of solution concentrations (0.001 M - 1 M). The results show some surprising co-ion effects at high electrolyte concentrations (>0.5 M) and indicate that even ions generally considered to be indifferent induce a shift in iso-electric point (i.e.p.) which is inferred as being due to specific adsorption of ions. The shear yield stress values of concentrated titania dispersions were measured using a Bohlin C-VOR stress controlled rheometer. The shear

Preparation and study of properties of dispersed graphene oxide

Directory of Open Access Journals (Sweden)

Evgeniya Seliverstova

2015-09-01

Full Text Available Ability of graphene oxide to form stable dispersion in organic solvents was studied in this work. As it was shown, sonication of graphene leads to the decreas of the particle size. Stability of prepared graphene dispersions was studied upon measurements of distribution of number of the particles via size and change of optical density of the solutions with time. It was found that graphene oxide forms a more stable dispersion in tetrahydrofuran and dimethylformamide than in chloroform and acetone.

Study on the irradiation swelling of U3Si2-Al dispersion fuel

International Nuclear Information System (INIS)

Xing Zhonghu; Ying Shihao

2001-01-01

The dominant modeling mechanisms on irradiation swelling of U 3 Si 2 -Al dispersion fuel are introduced. The core of dispersion fuel is looked to as micro-fuel elements of continuous matrix. The formation processes of gas bubbles in the fuel phase are described through the behavior mechanisms of fission gases. The swelling in the fuel phase causes the interaction between fuel particles and metal matrix, and the metal matrix can restrain the irradiation swelling of fuel particles. The developed code can predict irradiation-swelling values according to the parameters of fuel elements and irradiation conditions, and the predicted values are in agreement with the measured results

In-pile irradiation of rock-like oxide fuels

International Nuclear Information System (INIS)

Nitani, N.; Kuramoto, K.; Yamashita, T.; Nakano, Y.; Akie, H.

2001-01-01

Five kinds of ROX fuels were prepared and irradiated using 20% enriched U instead of Pu. Non-destructive and destructive post-irradiation examinations were carried out. FP gas release rates of the particle-dispersed type fuels and homogeneously-blended type fuels were larger than that of the Yttria-stabilized zirconia containing UO 2 single phase fuel. From results of SEM and EPMA, decomposition of the spinel was observed. The decomposition of the spinel is probably avoided by lowering the irradiation temperature, less than 1700 K. The regions suffering the irradiation damage of the particle dispersed type fuels were less than those of the homogeneously-blended type fuels. (author)

In-pile irradiation of rock-like oxide fuels

Energy Technology Data Exchange (ETDEWEB)

Nitani, N.; Kuramoto, K.; Yamashita, T.; Nakano, Y.; Akie, H. [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan)

2001-07-01

Five kinds of ROX fuels were prepared and irradiated using 20% enriched U instead of Pu. Non-destructive and destructive post-irradiation examinations were carried out. FP gas release rates of the particle-dispersed type fuels and homogeneously-blended type fuels were larger than that of the Yttria-stabilized zirconia containing UO{sub 2} single phase fuel. From results of SEM and EPMA, decomposition of the spinel was observed. The decomposition of the spinel is probably avoided by lowering the irradiation temperature, less than 1700 K. The regions suffering the irradiation damage of the particle dispersed type fuels were less than those of the homogeneously-blended type fuels. (author)

Microstructure of the irradiated U 3Si 2/Al silicide dispersion fuel

Science.gov (United States)

Gan, J.; Keiser, D. D.; Miller, B. D.; Jue, J.-F.; Robinson, A. B.; Madden, J. W.; Medvedev, P. G.; Wachs, D. M.

2011-12-01

The silicide dispersion fuel of U 3Si 2/Al is recognized as the best performance fuel for many nuclear research and test reactors with up to 4.8 gU/cm 3 fuel loading. An irradiated U 3Si 2/Al dispersion fuel ( 235U ˜ 75%) from the high-flux side of a fuel plate (U0R040) from the Reduced Enrichment for Research and Test Reactors (RERTR)-8 test was characterized using transmission electron microscopy (TEM). The fuel was irradiated in the Advanced Test Reactor (ATR) for 105 days. The average irradiation temperature and fission density of the U 3Si 2 fuel particles for the TEM sample are estimated to be approximately 110 °C and 5.4 × 10 27 f/m 3. The characterization was performed using a 200-kV TEM. The U/Si ratio for the fuel particle and (Si + Al)/U for the fuel-matrix-interaction layer are approximately 1.1 and 4-10, respectively. The estimated average diameter, number density and volume fraction for small bubbles (<1 μm) in the fuel particle are ˜94 nm, 1.05 × 10 20 m -3 and ˜11%, respectively. The results and their implication on the performance of the U 3Si 2/Al silicide dispersion fuel are discussed.

Jet Fuel Based High Pressure Solid Oxide Fuel Cell System

Science.gov (United States)

Gummalla, Mallika (Inventor); Yamanis, Jean (Inventor); Olsommer, Benoit (Inventor); Dardas, Zissis (Inventor); Bayt, Robert (Inventor); Srinivasan, Hari (Inventor); Dasgupta, Arindam (Inventor); Hardin, Larry (Inventor)

2015-01-01

A power system for an aircraft includes a solid oxide fuel cell system which generates electric power for the aircraft and an exhaust stream; and a heat exchanger for transferring heat from the exhaust stream of the solid oxide fuel cell to a heat requiring system or component of the aircraft. The heat can be transferred to fuel for the primary engine of the aircraft. Further, the same fuel can be used to power both the primary engine and the SOFC. A heat exchanger is positioned to cool reformate before feeding to the fuel cell. SOFC exhaust is treated and used as inerting gas. Finally, oxidant to the SOFC can be obtained from the aircraft cabin, or exterior, or both.

Tungsten and molybdenum with oxide dispersion, production and properties

International Nuclear Information System (INIS)

Haerdtle, S.; Schmidberger, R.

1989-01-01

By the reaction spray process metal powders with dispersed metal oxides can be produced in one step. The systems investigated here are tungsten and molybdenum with 0,5% resp. 5% La 2 O 3 , Y 2 O 3 and ZrO 2 . The oxides with diameters below 0,5μm are finely dispersed within the metal powder particles. The sinterability of the powders depends on the oxide content. Maximum density at an oxide content of 0,5% is about 96% at a sintering temperature of 1600 0 C. The type of oxide influences the densification versus temperature but not the final density. 5 refs., 11 figs. (Author)

A conceptual model for the fuel oxidation of defective fuel

International Nuclear Information System (INIS)

Higgs, J.D.; Lewis, B.J.; Thompson, W.T.; He, Z.

2007-01-01

A mechanistic conceptual model has been developed to predict the fuel oxidation behaviour in operating defective fuel elements for water-cooled nuclear reactors. This theoretical work accounts for gas-phase transport and sheath reactions in the fuel-to-sheath gap to determine the local oxygen potential. An improved thermodynamic analysis has also been incorporated into the model to describe the equilibrium state of the oxidized fuel. The fuel oxidation kinetics treatment accounts for multi-phase transport including normal diffusion and thermodiffusion for interstitial oxygen migration in the solid, as well as gas-phase transport in the fuel pellet cracks. The fuel oxidation treatment is further coupled to a heat conduction equation. A numerical solution of the coupled transport equations is obtained by a finite-element technique with the FEMLAB 3.1 software package. The model is able to provide radial-axial profiles of the oxygen-to-uranium ratio and the fuel temperatures as a function of time in the defective element for a wide range of element powers and defect sizes. The model results are assessed against coulometric titration measurements of the oxygen-to-metal profile for pellet samples taken from ten spent defective elements discharged from the National Research Universal Reactor at the Chalk River Laboratories and commercial reactors

Solid Oxide Fuel Cell Experimental Laboratory

Data.gov (United States)

Federal Laboratory Consortium — NETL’s Solid Oxide Fuel Cell Experimental Laboratory in Morgantown, WV, gives researchers access to models and simulations that predict how solid oxide fuel cells...

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.

TREAT experimental data base regarding fuel dispersals in LMFBR loss-of-flow accidents

International Nuclear Information System (INIS)

Simms, R.; Fink, C.L.; Stanford, G.S.; Regis, J.P.

1981-01-01

The reactivity feedback from fuel relocation is a central issue in the analysis of loss-of-flow (LOF) accidents in LMFBRs. Fuel relocation has been studied in a number of LOF simulations in the TREAT reactor. In this paper the results of these tests are analyzed, using, as the principal figure of merit, the changes in equivalent fuel worth associated with the fuel motion. The equivalent fuel worth was calculated from the measured axial fuel distributions by weighting the data with a typical LMFBR fuel-worth function. At nominal power, the initial fuel relocation resulted in increases in equivalent fuel worth. Above nominal power the fuel motion was dispersive, but the dispersive driving forces could not unequivocally be identified from the experimental data

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)

Electrolytes for solid oxide fuel cells

Science.gov (United States)

Fergus, Jeffrey W.

The high operating temperature of solid oxide fuel cells (SOFCs), as compared to polymer electrolyte membrane fuel cells (PEMFCs), improves tolerance to impurities in the fuel, but also creates challenges in the development of suitable materials for the various fuel cell components. In response to these challenges, intermediate temperature solid oxide fuel cells (IT-SOFCs) are being developed to reduce high-temperature material requirements, which will extend useful lifetime, improve durability and reduce cost, while maintaining good fuel flexibility. A major challenge in reducing the operating temperature of SOFCs is the development of solid electrolyte materials with sufficient conductivity to maintain acceptably low ohmic losses during operation. In this paper, solid electrolytes being developed for solid oxide fuel cells, including zirconia-, ceria- and lanthanum gallate-based materials, are reviewed and compared. The focus is on the conductivity, but other issues, such as compatibility with electrode materials, are also discussed.

Electrolytes for solid oxide fuel cells

Energy Technology Data Exchange (ETDEWEB)

Fergus, Jeffrey W. [Auburn University, Materials Research and Education Center, 275 Wilmore Laboratories, Auburn, AL 36849 (United States)

2006-11-08

The high operating temperature of solid oxide fuel cells (SOFCs), as compared to polymer electrolyte membrane fuel cells (PEMFCs), improves tolerance to impurities in the fuel, but also creates challenges in the development of suitable materials for the various fuel cell components. In response to these challenges, intermediate temperature solid oxide fuel cells (IT-SOFCs) are being developed to reduce high-temperature material requirements, which will extend useful lifetime, improve durability and reduce cost, while maintaining good fuel flexibility. A major challenge in reducing the operating temperature of SOFCs is the development of solid electrolyte materials with sufficient conductivity to maintain acceptably low ohmic losses during operation. In this paper, solid electrolytes being developed for solid oxide fuel cells, including zirconia-, ceria- and lanthanum gallate-based materials, are reviewed and compared. The focus is on the conductivity, but other issues, such as compatibility with electrode materials, are also discussed. (author)

Behaviour of defective CANDU fuel: fuel oxidation kinetic and thermodynamic modelling

International Nuclear Information System (INIS)

Higgs, J.

2005-01-01

The thermal performance of operating CANDU fuel under defect conditions is affected by the ingress of heavy water into the fuel element. A mechanistic model has been developed to predict the extent of fuel oxidation in defective fuel and its affect on fuel thermal performance. A thermodynamic treatment of such oxidized fuel has been performed as a basis for the boundary conditions in the kinetic model. Both the kinetic and thermodynamic models have been benchmarked against recent experimental work. (author)

Constrained Sintering in Fabrication of Solid Oxide Fuel Cells.

Science.gov (United States)

Lee, Hae-Weon; Park, Mansoo; Hong, Jongsup; Kim, Hyoungchul; Yoon, Kyung Joong; Son, Ji-Won; Lee, Jong-Ho; Kim, Byung-Kook

2016-08-09

Solid oxide fuel cells (SOFCs) are inevitably affected by the tensile stress field imposed by the rigid substrate during constrained sintering, which strongly affects microstructural evolution and flaw generation in the fabrication process and subsequent operation. In the case of sintering a composite cathode, one component acts as a continuous matrix phase while the other acts as a dispersed phase depending upon the initial composition and packing structure. The clustering of dispersed particles in the matrix has significant effects on the final microstructure, and strong rigidity of the clusters covering the entire cathode volume is desirable to obtain stable pore structure. The local constraints developed around the dispersed particles and their clusters effectively suppress generation of major process flaws, and microstructural features such as triple phase boundary and porosity could be readily controlled by adjusting the content and size of the dispersed particles. However, in the fabrication of the dense electrolyte layer via the chemical solution deposition route using slow-sintering nanoparticles dispersed in a sol matrix, the rigidity of the cluster should be minimized for the fine matrix to continuously densify, and special care should be taken in selecting the size of the dispersed particles to optimize the thermodynamic stability criteria of the grain size and film thickness. The principles of constrained sintering presented in this paper could be used as basic guidelines for realizing the ideal microstructure of SOFCs.

Quantification of oxide particle composition in model oxide dispersion strengthened steel alloys

Energy Technology Data Exchange (ETDEWEB)

London, A.J., E-mail: andrew.london@materials.ox.ac.uk [Department of Materials, University of Oxford, Parks Road, Oxford OX1 3PH (United Kingdom); Lozano-Perez, S.; Moody, M.P. [Department of Materials, University of Oxford, Parks Road, Oxford OX1 3PH (United Kingdom); Amirthapandian, S.; Panigrahi, B.K.; Sundar, C.S. [Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102, TN (India); Grovenor, C.R.M. [Department of Materials, University of Oxford, Parks Road, Oxford OX1 3PH (United Kingdom)

2015-12-15

Oxide dispersion strengthened ferritic steels (ODS) are being considered for structural components of future designs of fission and fusion reactors because of their impressive high-temperature mechanical properties and resistance to radiation damage, both of which arise from the nanoscale oxide particles they contain. Because of the critical importance of these nanoscale phases, significant research activity has been dedicated to analysing their precise size, shape and composition (Odette et al., Annu. Rev. Mater. Res. 38 (2008) 471–503 [1]; Miller et al., Mater. Sci. Technol. 29(10) (2013) 1174–1178 [2]). As part of a project to develop new fuel cladding alloys in India, model ODS alloys have been produced with the compositions, Fe–0.3Y{sub 2}O{sub 3}, Fe–0.2Ti–0.3Y{sub 2}O{sub 3} and Fe–14Cr–0.2Ti–0.3Y{sub 2}O{sub 3}. The oxide particles in these three model alloys have been studied by APT in their as-received state and following ion irradiation (as a proxy for neutron irradiation) at various temperatures. In order to adequately quantify the composition of the oxide clusters, several difficulties must be managed, including issues relating to the chemical identification (ranging and variable peak-overlaps); trajectory aberrations and chemical structure; and particle sizing. This paper presents how these issues can be addressed by the application of bespoke data analysis tools and correlative microscopy. A discussion follows concerning the achievable precision in these measurements, with reference to the fundamental limiting factors.

Quantification of oxide particle composition in model oxide dispersion strengthened steel alloys.

Science.gov (United States)

London, A J; Lozano-Perez, S; Moody, M P; Amirthapandian, S; Panigrahi, B K; Sundar, C S; Grovenor, C R M

2015-12-01

Oxide dispersion strengthened ferritic steels (ODS) are being considered for structural components of future designs of fission and fusion reactors because of their impressive high-temperature mechanical properties and resistance to radiation damage, both of which arise from the nanoscale oxide particles they contain. Because of the critical importance of these nanoscale phases, significant research activity has been dedicated to analysing their precise size, shape and composition (Odette et al., Annu. Rev. Mater. Res. 38 (2008) 471-503 [1]; Miller et al., Mater. Sci. Technol. 29(10) (2013) 1174-1178 [2]). As part of a project to develop new fuel cladding alloys in India, model ODS alloys have been produced with the compositions, Fe-0.3Y2O3, Fe-0.2Ti-0.3Y2O3 and Fe-14Cr-0.2Ti-0.3Y2O3. The oxide particles in these three model alloys have been studied by APT in their as-received state and following ion irradiation (as a proxy for neutron irradiation) at various temperatures. In order to adequately quantify the composition of the oxide clusters, several difficulties must be managed, including issues relating to the chemical identification (ranging and variable peak-overlaps); trajectory aberrations and chemical structure; and particle sizing. This paper presents how these issues can be addressed by the application of bespoke data analysis tools and correlative microscopy. A discussion follows concerning the achievable precision in these measurements, with reference to the fundamental limiting factors. Copyright © 2015 Elsevier B.V. All rights reserved.

Role of Oxides and Porosity on High-Temperature Oxidation of Liquid-Fueled HVOF Thermal-Sprayed Ni50Cr Coatings

Science.gov (United States)

Song, B.; Bai, M.; Voisey, K. T.; Hussain, T.

2017-02-01

High chromium content in Ni50Cr thermally sprayed coatings can generate a dense and protective scale at the surface of coating. Thus, the Ni50Cr coating is widely used in high-temperature oxidation and corrosion applications. A commercially available gas atomized Ni50Cr powder was sprayed onto a power plant steel (ASME P92) using a liquid-fueled high velocity oxy-fuel thermal spray with three processing parameters in this study. Microstructure of as-sprayed coatings was examined using oxygen content analysis, mercury intrusion porosimetry, scanning electron microscope (SEM), energy-dispersive x-ray spectroscopy (EDX) and x-ray diffraction (XRD). Short-term air oxidation tests (4 h) of freestanding coatings (without boiler steel substrate) in a thermogravimetric analyzer at 700 °C were performed to obtain the kinetics of oxidation of the as-sprayed coating. Long-term air oxidation tests (100 h) of the coated substrates were performed at same temperature to obtain the oxidation products for further characterization in detail using SEM/EDX and XRD. In all samples, oxides of various morphologies developed on top of the Ni50Cr coatings. Cr2O3 was the main oxidation product on the surface of all three coatings. The coating with medium porosity and medium oxygen content has the best high-temperature oxidation performance in this study.

Highly Selective TiN-Supported Highly Dispersed Pt Catalyst: Ultra Active toward Hydrogen Oxidation and Inactive toward Oxygen Reduction.

Science.gov (United States)

Luo, Junming; Tang, Haibo; Tian, Xinlong; Hou, Sanying; Li, Xiuhua; Du, Li; Liao, Shijun

2018-01-31

The severe dissolution of the cathode catalyst, caused by an undesired oxygen reduction reaction at the anode during startup and shutdown, is a fatal challenge to practical applications of polymer electrolyte membrane fuel cells. To address this important issue, according to the distinct structure-sensitivity between the σ-type bond in H 2 and the π-type bond in O 2 , we design a HD-Pt/TiN material by highly dispersing Pt on the TiN surface to inhibit the unwanted oxygen reduction reaction. The highly dispersed Pt/TiN catalyst exhibits excellent selectivity toward hydrogen oxidation and oxygen reduction reactions. With a Pt loading of 0.88 wt %, our catalyst shows excellent hydrogen oxidation reaction activity, close to that of commercial 20 wt % Pt/C catalyst, and much lower oxygen reduction reaction activity than the commercial 20 wt % Pt/C catalyst. The lack of well-ordered Pt facets is responsible for the excellent selectivity of the HD-Pt/TiN materials toward hydrogen oxidation and oxygen reduction reactions. Our work provides a new and cost-effective solution to design selective catalysts toward hydrogen oxidation and oxygen reduction reactions, making the strategy of using oxygen-tolerant anode catalyst to improve the stability of polymer electrolyte membrane fuel cells during startup and shutdown more affordable and practical.

Thermochemical aspects of fuel-cladding and fuel-coolant interactions in LMFBR oxide fuel pins

International Nuclear Information System (INIS)

Adamson, M.G.; Aitken, E.A.; Caputi, R.W.; Potter, P.E.; Mignanelli, M.A.

1979-01-01

This paper examines several thermochemical aspects of the fuel-cladding, fuel-coolant and fuel-fission product interactions that occur in LMFBR austenitic stainless steel-clad mixed (U,Pu)-oxide fuel pins during irradiation under normal operating conditions. Results are reported from a variety of high temperature EMF cell experiments in which continuous oxygen activity measurements on reacting and equilibrium mixtures of metal oxides and (excess) liquid alkali metal (Na, K, Cs) were performed. Oxygen potential and 0:M thresholds for Na-fuel reactions are re-evaluated in the light of new measurements and newly-assessed thermochemical data, and the influence on oxygen potential of possible U-Pu segregation between oxide and urano-plutonate (equilibrium) phases has been analyzed. (orig./RW) [de

Transient survivability of LMR oxide fuel pins

International Nuclear Information System (INIS)

Weber, E.T.; Pitner, A.L.; Bard, F.E.; Culley, G.E.; Hunter, C.W.

1986-01-01

Fuel pin integrity during transient events must be assessed for both the core design and safety analysis phases of a reactor project. A significant increase in the experience related to limits of integrity for oxide fuel pins in transient overpower events has been realized from testing of fuel pins irradiated in FFTF and PFR. Fourteen FFTF irradiated fuel pins were tested in TREAT, representing a range of burnups, overpower ramp rates and maximum overpower conditions. Results of these tests along with similar testing in the PFR/TREAT program, provide a demonstration of significant safety margins for oxide fuel pins. Useful information applied in analytical extrapolation of fuel pin test data have been developed from laboratory transient tests on irradiated fuel cladding (FCTT) and on unirradiated fuel pellet deformation. These refinements in oxide fuel transient performance are being applied in assessment of transient capabilities of long lifetime fuel designs using ferritic cladding

Transmission electron microscopy characterization of irradiated U-7Mo/Al-2Si dispersion fuel

International Nuclear Information System (INIS)

Gan, J.; Keiser, D.D.; Wachs, D.M.; Robinson, A.B.; Miller, B.D.; Allen, T.R.

2010-01-01

The plate-type dispersion fuels, with the atomized U(Mo) fuel particles dispersed in the Al or Al alloy matrix, are being developed for use in research and test reactors worldwide. It is found that the irradiation performance of a plate-type dispersion fuel depends on the radiation stability of the various phases in a fuel plate. Transmission electron microscopy was performed on a sample (peak fuel mid-plane temperature ∼109 deg. C and fission density ∼4.5 x 10 27 f m -3 ) taken from an irradiated U-7Mo dispersion fuel plate with Al-2Si alloy matrix to investigate the role of Si addition in the matrix on the radiation stability of the phase(s) in the U-7Mo fuel/matrix interaction layer. A similar interaction layer that forms in irradiated U-7Mo dispersion fuels with pure Al matrix has been found to exhibit poor irradiation stability, likely as a result of poor fission gas retention. The interaction layer for both U-7Mo/Al-2Si and U-7Mo/Al fuels is observed to be amorphous. However, unlike the latter, the amorphous layer for the former was found to effectively retain fission gases in areas with high Si concentration. When the Si concentration becomes relatively low, the fission gas bubbles agglomerate into fewer large pores. Within the U-7Mo fuel particles, a bubble superlattice ordered as fcc structure and oriented parallel to the bcc metal lattice was observed where the average bubble size and the superlattice constant are 3.5 nm and 11.5 nm, respectively. The estimated fission gas inventory in the bubble superlattice correlates well with the fission density in the fuel.

Modeling and preliminary analysis on the temperature profile of the (TRU-Zr)-Zr dispersion fuel rod for HYPER

International Nuclear Information System (INIS)

Lee, B. W.; Hwang, W.; Lee, B. S.; Park, W. S.

2000-01-01

Either TRU-Zr metal alloy or (TRU-Zr)-Zr dispersion fuel is considered as a blanket fuel for HYPER(Hybrid Power Extraction Reactor). In order to develop the code for dispersion fuel rod performance analysis under steady state condition, the fuel temperature distribution model which is the one of the most important factors in a fuel performance code has been developed in this paper,. This developed model computes the one dimensional radial temperature distribution of a cylindrical fuel rod. The temperature profile results by this model are compared with the temperature distributions of U 3 Si-A1 dispersion fuel and TRU-Zr metal alloy fuel. This model will be installed in performance analysis code for dispersion fuel

Predicted irradiation behavior of U3O8-Al dispersion fuels for production reactor applications

International Nuclear Information System (INIS)

Cronenberg, A.W.; Rest, J.

1990-01-01

Candidate fuels for the new heavy-water production reactor include uranium/aluminum alloy and U 3 O 8 -Al dispersion fuels. The U 3 O 8 -Al dispersion fuel would make possible higher uranium loadings and would facilitate uranium recycle. Research efforts on U 3 O 8 -Al fuel include in-pile irradiation studies and development of analytical tools to characterize the behavior of dispersion fuels at high-burnup. In this paper the irradiation performance of U 3 O 8 -Al is assessed using the mechanistic Dispersion Analysis Research Tool (DART) code. Predictions of fuel swelling and alteration of thermal conductivity are presented and compared with experimental data. Calculational results indicate good agreement with available data where the effects of as-fabricated porosity and U 3 O 8 -Al oxygen exchange reactions are shown to exert a controlling influence on irradiation behavior. The DART code is judged to be a useful tool for assessing U 3 O 8 -Al performance over a wide range of irradiation conditions

POST-IRRADIATION ANALYSES OF U-MO DISPERSION FUEL RODS OF KOMO TESTS AT HANARO

Directory of Open Access Journals (Sweden)

H.J. RYU

2013-12-01

Full Text Available Since 2001, a series of five irradiation test campaigns for atomized U-Mo dispersion fuel rods, KOMO-1, -2, -3, -4, and -5, has been conducted at HANARO (Korea in order to develop high performance low enriched uranium dispersion fuel for research reactors. The KOMO irradiation tests provided valuable information on the irradiation behavior of U-Mo fuel that results from the distinct fuel design and irradiation conditions of the rod fuel for HANARO. Full size U-Mo dispersion fuel rods of 4–5 g-U/cm3 were irradiated at a maximum linear power of approximately 105 kW/m up to 85% of the initial U-235 depletion burnup without breakaway swelling or fuel cladding failure. Electron probe microanalyses of the irradiated samples showed localized distribution of the silicon that was added in the matrix during fuel fabrication and confirmed its beneficial effect on interaction layer growth during irradiation. The modifications of U-Mo fuel particles by the addition of a ternary alloying element (Ti or Zr, additional protective coatings (silicide or nitride, and the use of larger fuel particles resulted in significantly reduced interaction layers between fuel particles and Al.

Fuel cycle cost comparisons with oxide and silicide fuels

Energy Technology Data Exchange (ETDEWEB)

Matos, J E; Freese, K E [RERTR Program, Argonne National Laboratory (United States)

1983-09-01

This paper addresses fuel cycle cost comparisons for a generic 10 MW reactor with HEU aluminide fuel and with LEU oxide and silicide fuels in several fuel element geometries. The intention of this study is to provide a consistent assessment of various design options from a cost point of view. The status of the development and demonstration of the oxide and silicide fuels are presented in several papers in these proceedings. Routine utilization of these fuels with the uranium densities considered here requires that they are successfully demonstrated and licensed. Thermal-hydraulic safety margins, shutdown margins, mixed cores, and transient analyses are not addressed here, but analyses of these safety issues are in progress for a limited number of the most promising design options. Fuel cycle cost benefits could result if a number of reactors were to utilize fuel elements with the same number or different numbers of the same standard fuel plate. Data is presented to quantify these potential cost benefits. This analysis shows that there are a number of fuel element designs using LEU oxide or silicide fuels that have either the same or lower total fuel cycle costs than the HEU design. Use of these fuels with the uranium densities considered requires that they are successfully demonstrated and licensed. All safety criteria for the reactor with these fuel element designs need to be satisfied as well. With LEU oxide fuel, 31 g U/cm{sup 3} 1 and 0.76 mm--thick fuel meat, elements with 18-22 plates 320-391 g {sup 235}U) result in the same or lower total costs than with the HEU element 23 plates, 280 g {sup 235}U). Higher LEU loadings (more plates per element) are needed for larger excess reactivity requirements. However, there is little cost advantage to using more than 20 of these plates per element. Increasing the fuel meat thickness from 0.76 mm to 1.0 mm with 3.1 g U/cm{sup 3} in the design with 20 plates per element could result in significant cost reductions if the

A comparison of the metallurgical behaviour of dispersion fuels with uranium silicides and U6Fe as dispersants

International Nuclear Information System (INIS)

Nazare, S.

1984-01-01

In the past few years metallurgical studies have been carried out to develop fuel dispersions with U-densities up to 7.0 Mg U m -3 . Uranium silicides have been considered to be the prime candidates as dispersants; U 6 Fe being a potential alternative on account of its higher U-density. The objective of this paper is to compare the metallurgical behaviour of these two material combinations with regard to the following aspects: (1) preparation of the compounds U 3 Si, U 3 Si 2 and U 6 Fe; (2) powder metallurgical processing to miniature fuel element plates; (3) reaction behaviour under equilibrium conditions in the relevant portions of the ternary U-Si-Al and U-Fe-Al systems; (4) dimensional stability of the fuel plates after prolonged thermal treatment; (5) thermochemical behaviour of fuel plates at temperatures near the melting point of the cladding. Based on this data, the possible advantages of each fuel combination are discussed. (author)

Internal fuel pin oxidizer

International Nuclear Information System (INIS)

Andrews, M.G.

1978-01-01

A nuclear fuel pin has positioned within it material which will decompose to release an oxidizing agent which will react with the cladding of the pin and form a protective oxide film on the internal surface of the cladding

Reliability analysis of dispersion nuclear fuel elements

Science.gov (United States)

Ding, Shurong; Jiang, Xin; Huo, Yongzhong; Li, Lin an

2008-03-01

Taking a dispersion fuel element as a special particle composite, the representative volume element is chosen to act as the research object. The fuel swelling is simulated through temperature increase. The large strain elastoplastic analysis is carried out for the mechanical behaviors using FEM. The results indicate that the fission swelling is simulated successfully; the thickness increments grow linearly with burnup; with increasing of burnup: (1) the first principal stresses at fuel particles change from tensile ones to compression ones, (2) the maximum Mises stresses at the particles transfer from the centers of fuel particles to the location close to the interfaces between the matrix and the particles, their values increase with burnup; the maximum Mises stresses at the matrix exist in the middle location between the two particles near the mid-plane along the length (or width) direction, and the maximum plastic strains are also at the above region.

Reliability analysis of dispersion nuclear fuel elements

Energy Technology Data Exchange (ETDEWEB)

Ding Shurong [Department of Mechanics and Engineering Science, Fudan University, Shanghai 200433 (China)], E-mail: dsr1971@163.com; Jiang Xin [Department of Mechanics and Engineering Science, Fudan University, Shanghai 200433 (China); Huo Yongzhong [Department of Mechanics and Engineering Science, Fudan University, Shanghai 200433 (China)], E-mail: yzhuo@fudan.edu.cn; Li Linan [Department of Mechanics, Tianjin University, Tianjin 300072 (China)

2008-03-15

Taking a dispersion fuel element as a special particle composite, the representative volume element is chosen to act as the research object. The fuel swelling is simulated through temperature increase. The large strain elastoplastic analysis is carried out for the mechanical behaviors using FEM. The results indicate that the fission swelling is simulated successfully; the thickness increments grow linearly with burnup; with increasing of burnup: (1) the first principal stresses at fuel particles change from tensile ones to compression ones, (2) the maximum Mises stresses at the particles transfer from the centers of fuel particles to the location close to the interfaces between the matrix and the particles, their values increase with burnup; the maximum Mises stresses at the matrix exist in the middle location between the two particles near the mid-plane along the length (or width) direction, and the maximum plastic strains are also at the above region.

Modeling the influence of interaction layer formation on thermal conductivity of U–Mo dispersion fuel

International Nuclear Information System (INIS)

Burkes, Douglas E.; Casella, Andrew M.; Huber, Tanja K.

2015-01-01

Highlights: • Hsu equation provides best thermal conductivity estimate of U–Mo dispersion fuel. • Simple model considering interaction layer formation was coupled with Hsu equation. • Interaction layer thermal conductivity is not the most important attribute. • Effective thermal conductivity is mostly influenced by interaction layer formation. • Fuel particle distribution also influences the effective thermal conductivity. - Abstract: The Global Threat Reduction Initiative Program continues to develop existing and new test reactor fuels to achieve the maximum attainable uranium loadings to support the conversion of a number of the world’s remaining high-enriched uranium fueled reactors to low-enriched uranium fuel. Currently, the program is focused on assisting with the development and qualification of a fuel design that consists of a uranium–molybdenum (U–Mo) alloy dispersed in an aluminum matrix. Thermal conductivity is an important consideration in determining the operational temperature of the fuel and can be influenced by interaction layer formation between the dispersed phase and matrix, porosity that forms during fabrication of the fuel plates or rods, and upon the concentration of the dispersed phase within the matrix. This paper develops and validates a simple model to study the influence of interaction layer formation, dispersed particle size, and volume fraction of dispersed phase in the matrix on the effective conductivity of the composite. The model shows excellent agreement with results previously presented in the literature. In particular, the thermal conductivity of the interaction layer does not appear to be as important in determining the effective conductivity of the composite, while formation of the interaction layer and subsequent consumption of the matrix reveals a rather significant effect. The effective thermal conductivity of the composite can be influenced by the dispersed particle distribution by minimizing interaction

A model to predict failure of irradiated U–Mo dispersion fuel

Energy Technology Data Exchange (ETDEWEB)

Burkes, Douglas E., E-mail: Douglas.Burkes@pnnl.gov; Senor, David J.; Casella, Andrew M.

2016-12-15

Highlights: • Simple model to predict failure of dispersion fuel meat designs. • Evaluated as a function of fabrication parameters and irradiation conditions. • Predictions compare well with experimental measurements of miniature fuel plates. • Interaction layer formation reduces matrix strength and increases temperature. • Si additions to the matrix appear effective only at moderate heat flux and burnup. - Abstract: Numerous global programs are focused on the continued development of existing and new research and test reactor fuels to achieve maximum attainable uranium loadings to support the conversion of a number of the world’s remaining high-enriched uranium fueled reactors to low-enriched uranium fuel. Some of these programs are focused on development and qualification of a fuel design that consists of a uranium–molybdenum (U–Mo) alloy dispersed in an aluminum matrix as one option for reactor conversion. The current paper extends a failure model originally developed for UO{sub 2}-stainless steel dispersion fuels and uses currently available thermal–mechanical property information for the materials of interest in the currently proposed design. A number of fabrication and irradiation parameters were investigated to understand the conditions at which failure of the matrix, classified as onset of pore formation in the matrix, might occur. The results compared well with experimental observations published as part of the Reduced Enrichment for Research and Test Reactors (RERTR)-6 and -7 mini-plate experiments. Fission rate, a function of the {sup 235}U enrichment, appeared to be the most influential parameter in premature failure, mainly as a result of increased interaction layer formation and operational temperature, which coincidentally decreased the strength of the matrix and caused more rapid fission gas production and recoil into the surrounding matrix material. Addition of silicon to the matrix appeared effective at reducing the rate of

Irradiation performance of uranium-molybdenum alloy dispersion fuels

International Nuclear Information System (INIS)

Almeida, Cirila Tacconi de

2005-01-01

The U-Mo-Al dispersion fuels of Material Test Reactors (MTR) are analyzed in terms of their irradiation performance. The irradiation performance aspects are associated to the neutronic and thermal hydraulics aspects to propose a new core configuration to the IEA-R1 reactor of IPEN-CNEN/SP using U-Mo-Al fuels. Core configurations using U-10Mo-Al fuels with uranium densities variable from 3 to 8 gU/cm 3 were analyzed with the computational programs Citation and MTRCR-IEA R1. Core configurations for fuels with uranium densities variable from 3 to 5 gU/cm 3 showed to be adequate to use in IEA-R1 reactor e should present a stable in reactor performance even at high burn-up. (author)

Progress in development of low-enriched U-Mo dispersion fuels

International Nuclear Information System (INIS)

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

2002-01-01

Results from post irradiation examinations and analyses of U-Mo/Al dispersion mini plates are presented. Irradiation test RERTR-5 contained mini- fuel plates with fuel loadings of 6 and 8 g U cm -3 . The fuel material consisted of 6, 7 and 10 wt. % Mo-uranium-alloy powders in atomized and machined form. The swelling behavior of the various fuel types is analyzed, indicating athermal swelling of the U-Mo alloy and temperature-dependent swelling owing to U-Mo/Al interdiffusion. (author)

Constrained Sintering in Fabrication of Solid Oxide Fuel Cells

Science.gov (United States)

Lee, Hae-Weon; Park, Mansoo; Hong, Jongsup; Kim, Hyoungchul; Yoon, Kyung Joong; Son, Ji-Won; Lee, Jong-Ho; Kim, Byung-Kook

2016-01-01

Solid oxide fuel cells (SOFCs) are inevitably affected by the tensile stress field imposed by the rigid substrate during constrained sintering, which strongly affects microstructural evolution and flaw generation in the fabrication process and subsequent operation. In the case of sintering a composite cathode, one component acts as a continuous matrix phase while the other acts as a dispersed phase depending upon the initial composition and packing structure. The clustering of dispersed particles in the matrix has significant effects on the final microstructure, and strong rigidity of the clusters covering the entire cathode volume is desirable to obtain stable pore structure. The local constraints developed around the dispersed particles and their clusters effectively suppress generation of major process flaws, and microstructural features such as triple phase boundary and porosity could be readily controlled by adjusting the content and size of the dispersed particles. However, in the fabrication of the dense electrolyte layer via the chemical solution deposition route using slow-sintering nanoparticles dispersed in a sol matrix, the rigidity of the cluster should be minimized for the fine matrix to continuously densify, and special care should be taken in selecting the size of the dispersed particles to optimize the thermodynamic stability criteria of the grain size and film thickness. The principles of constrained sintering presented in this paper could be used as basic guidelines for realizing the ideal microstructure of SOFCs. PMID:28773795

Fuel cracking in relation to fuel oxidation in support of an out-reactor instrumented defected fuel experiment

Energy Technology Data Exchange (ETDEWEB)

Quastel, A.; Thiriet, C. [Atomic Energy of Canada Limited, Chalk River, ON (Canada); Lewis, B., E-mail: brent.lewis@uoit.ca [Univ. of Ontario Inst. of Tech., Oshawa, ON (Canada); Corcoran, E., E-mail: emily.corcoran@rmc.ca [Royal Military College of Canada, Kingston, ON (Canada)

2014-07-01

An experimental program funded by the CANDU Owners Group (COG) is studying an out-reactor instrumented defected fuel experiment in Stern Laboratories (Hamilton, Ontario) with guidance from Atomic Energy of Canada Limited (AECL). The objective of this test is to provide experimental data for validation of a mechanistic fuel oxidation model. In this experiment a defected fuel element with UO{sub 2} pellets will be internally heated with an electrical heater element, causing the fuel to crack. By defecting the sheath in-situ the fuel will be exposed to light water coolant near normal reactor operating conditions (pressure 10 MPa and temperature 265-310{sup o}C) causing fuel oxidation, especially near the hotter regions of the fuel in the cracks. The fuel thermal conductivity will change, resulting in a change in the temperature distribution of the fuel element. This paper provides 2D r-θ plane strain solid mechanics models to simulate fuel thermal expansion, where conditions for fuel crack propagation are investigated with the thermal J integral to predict fuel crack stress intensity factors. Finally since fuel crack geometry can affect fuel oxidation this paper shows that the solid mechanics model with pre-set radial cracks can be coupled to a 2D r-θ fuel oxidation model. (author)

UK experience on fuel and cladding interaction in oxide fuels

Energy Technology Data Exchange (ETDEWEB)

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

1977-04-01

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

UK experience on fuel and cladding interaction in oxide fuels

International Nuclear Information System (INIS)

Batey, W.; Findlay, J.R.

1977-01-01

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

Gas transport in solid oxide fuel cells

CERN Document Server

He, Weidong; Dickerson, James

2014-01-01

This book provides a comprehensive overview of contemporary research and emerging measurement technologies associated with gas transport in solid oxide fuel cells. Within these pages, an introduction to the concept of gas diffusion in solid oxide fuel cells is presented. This book also discusses the history and underlying fundamental mechanisms of gas diffusion in solid oxide fuel cells, general theoretical mathematical models for gas diffusion, and traditional and advanced techniques for gas diffusivity measurement.

An investigation on the irradiation behavior of atomized U-Mo/Al dispersion rod fuels

International Nuclear Information System (INIS)

Park, J.M.; Ryu, H.J.; Lee, Y.S.; Lee, D.B.; Oh, S.J.; Yoo, B.O.; Jung, Y.H.; Sohn, D.S.; Kim, C.K.

2005-01-01

The second irradiation fuel experiment, KOMO-2, for the qualification test of atomized U-Mo dispersion rod fuels with U-loadings of 4-4.5 gU/cc at KAERI was finished after an irradiation up to 70 at% U 235 peak burn-up and subjected to the IMEF (Irradiation material Examination Facility) for a post-irradiation analysis in order to understand the fuel irradiation performance of the U-Mo dispersion fuel. Current results for PIE of KOMO-2 revealed that the U-Mo/Al dispersion fuel rods exhibited a sound performance without any break-away swelling, but most of the fuel rods irradiated at a high linear power showed an extensive formation of the interaction phase between the U-Mo particle and the Al matrix. In this paper, the analysis of the PIE results, which focused on the diffusion related microstructures obtained from the optical and EPMA (Electron Probe Micro Analysis) observations, will be presented in detail. And a thermal modeling will be carried out to calculate the temperature of the fuel rod during an irradiation. (author)

Development status of metallic, dispersion and non-oxide advanced and alternative fuels for power and research reactors

International Nuclear Information System (INIS)

2003-09-01

eighties until the present days. The aspects of HTGR fuels, as well as partitioning and transmutation (P and T) of minor actinides and relative specific fuels have not been addressed. The International Atomic Energy Agency's (IAEA) Division of Nuclear Fuel Cycle and Waste Technology has been closely involved for many years in the above mentioned activities in the framework of the Advisory Group on Advanced Fuel Technology and Performance (fast reactor fuels) and Technical Working Group on Water Reactor Fuel Performance and Technology (thermal power reactor fuels). Apart from the progress made during the last decade, this report summarizes technological approaches, out-of-pile and in-pile properties of many types of advanced non-oxide fuels. It is expected that the report will provide IAEA Member States and their nuclear engineers with useful information and will preserve knowledge in the area for future developments. The review was prepared by a group of experts in the field from Germany, India and the Russian Federation and supported by information from specialists in Japan, Switzerland and the IAEA engaged in non-oxide fuel developments and related subjects

Oxide dispersion-strengthened ferritic alloys

International Nuclear Information System (INIS)

Asbroeck, P. van.

1976-10-01

The publication gives the available data on the DTO2 dispersion-strengthened ferritic alloy developed at C.E.N./S.C.K. Mol, Belgium. DTO2 is a Fe-Cr-Mo ferritic alloy, strengthened by addition of titanium oxide and of titanium leading to the formation of Chi phase. It was developed for use as canning material for fast breeder reactors. (author)

Fission induced swelling of U–Mo/Al dispersion fuel

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); Jeong, G.Y. [Ulsan National Institute of Science and Technology, 50 UNIST-gil, Eonyang-eup, Uljoo-gun, Ulsan 689-798 (Korea, Republic of); Park, J.M. [Korea Atomic Energy Research Institute, 989-111 Daedeok-daero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Robinson, A.B. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States)

2015-10-15

Fission-induced swelling of U–Mo/Al dispersion fuel meat was measured using microscopy images obtained from post-irradiation examination. The data of reduced-size plate-type test samples and rod-type test samples were employed for this work. A model to predict the meat swelling of U–Mo/Al dispersion fuel was developed. This model is composed of several submodels including a model for interaction layer (IL) growth between U–Mo and Al matrix, a model for IL thickness to IL volume conversion, a correlation for the fission-induced swelling of U–Mo alloy particles, a correlation for the fission-induced swelling of IL, and models of U–Mo and Al consumption by IL growth. The model was validated using full-size plate data that were not included in the model development.

Electrometallurgical treatment of oxide spent fuels

International Nuclear Information System (INIS)

Karell, E. J.

1999-01-01

The Department of Energy (DOE) inventory of spent nuclear fuel contains a wide variety of oxide fuel types that may be unsuitable for direct repository disposal in their current form. The molten-salt electrometallurgical treatment technique developed by Argonne National Laboratory (ANL) has the potential to simplify preparing and qualifying these fuels for disposal by converting them into three uniform product streams: uranium metal, a metal waste form, and a ceramic waste form. This paper describes the major steps in the electrometallurgical treatment process for oxide fuels and provides the results of recent experiments performed to develop and scale up the process

Oxidation behavior of fuel cladding tube in spent fuel pool accident condition

International Nuclear Information System (INIS)

Nemoto, Yoshiyuki; Kaji, Yoshiyuki; Ogawa, Chihiro; Nakashima, Kazuo; Tojo, Masayuki

2017-01-01

In spent fuel pool (SFP) under loss-of-cooling or loss-of-coolant severe accident condition, the spent fuels will be exposed to air and heated by their own residual decay heat. Integrity of fuel cladding is crucial for SFP safety therefore study on cladding oxidation in air at high temperature is important. Zircaloy-2 (Zry2) and zircaloy-4 (Zry4) were applied for thermogravimetric analyses (TGA) in different temperatures in air at different flow rates to evaluate oxidation behavior. Oxidation rate increased with testing temperature. In a range of flow rate of air which is predictable in spent fuel lack during a hypothetical SFP accident, influence of flow rate was not clearly observed below 950degC for the Zry2, or below 1050degC for Zry4. In higher temperature, oxidation rate was higher in high rate condition, and this trend was seen clearer when temperature increased. Oxide layers were carefully examined after the TGA analyses and compared with mass gain data to investigate detail of oxidation process in air. It was revealed that the mass gain data in pre-breakaway regime reflects growth of dense oxide film on specimen surface, meanwhile in post-breakaway regime, it reflects growth of porous oxide layer beneath fracture of the dense oxide film. (author)

Irradiation behavior of low-enriched U/sub 6/Fe-Al dispersion fuel elements

Energy Technology Data Exchange (ETDEWEB)

Hofman, G.L.; Domagala, R.F.; Copeland, G.L.

1987-10-01

An irradiation test of miniature fuel plates containing low-enriched (20% /sup 235/U)U/sub 6/Fe dispersed and clad in Al was performed. The postirradiation examination shows U/sub 6/Fe to form extensive fission gas bubbles at burnups of only approx. = 20% of the original 20% fuel enrichment. Plate failure by fission gas-driven pillowing occurred at approx. = 40% burnup. This places U/sub 6/FE at the lowest burnup capability among low enriched dispersion fuels that have been tested for use in research and test reactors

Uranium dispersion in the coating of weak-acid-resin-deprived HTGR fuel microspheres

International Nuclear Information System (INIS)

Weber, G.W.; Beatty, R.L.; Tennery, V.J.; Lackey, W.J. Jr.

1976-02-01

The current reference HTGR recycle fuel particle is a UO 2 /UC 2 kernel with a Triso coating comprising a low-density pyrocarbon (PyC) buffer, a high-density PyC inner LTI coating, SiC, and a high-density PyC outer LTI. The kernel is fabricated from a weak-acid ion exchange resin (WAR). Microradiographic examination of coated WAR particles has demonstrated that considerable U can be transferred from the kernel to the buffer coating during fabrication. Investigation of causes of fuel dispersion has indicated several different factors that contribute to fuel redistribution if not properly controlled. The presence of a nonequilibrium UC/sub 1-x/O/sub x/ (0 less than or equal to x less than or equal to 0.3) phase had no significant effect on initiating fuel dispersion. Gross exposure of the completed fuel kernel to ambient atmosphere or to water vapor at room temperature produced very minimal levels of dispersion. Exposure of the fuel to perchloroethylene during buffer and inner LTI deposition produced massive redistribution. Fuel redistribution observed in Triso-coated particles results from permeation of the inner LTI by HCl during SiC deposition. As the decomposition of CH 3 Cl 3 Si is used to deposit SiC, chlorine is readily available during this process. The permeability of the inner LTI coating has a marked effect on the extent of this mode of fuel dispersion. LTI permeability was determined by chlorine leaching studies to be a strong function of density, coating gas dilution, and coating temperature but relatively unaffected by application of a seal coat, variations in coating thickness, and annealing at 1800 0 C. Mechanical attrition of the kernels during processing was identified as a potential source of U-bearing fines that may be incorporated into the coating in some circumstances

Study of the residual porosity in fuel plate cores based on U3O8 - Al dispersions

International Nuclear Information System (INIS)

Durazzo, M.

2005-01-01

The residual porosity in the meat of nuclear dispersion fuel plates, the fabrication voids, explains the corrosion behaviour of the meats when exposed to the water used as coolant and moderator of MTR type research reactors. The fabrication voids also explain variations in irradiation performance of many fuel dispersion for nuclear reactors. To obtain improved corrosion and irradiation performance, we must understand the fabrication factors that control the amount of void volume in fuel plate meats. The purpose of this study was to investigate the void content of aluminum-base dispersion-type U 3 O 8 -Al fuel plates depending on the characteristics of the starting fuel dispersion used to produce the fuel meat, which is fabricated by pressing. The void content depends on the U 3 O 8 concentration. For a particular U 3 O 8 content, the rolling process establishes a constant void concentration, which is called equilibrium porosity. The equilibrium quantity of voids is insensitive to the initial density of the fuel compact. (author)

Mechanical analysis of UMo/Al dispersion fuel

International Nuclear Information System (INIS)

Jeong, Gwan Yoon; Kim, Yeon Soo; Sohn, Dong-Seong

2015-01-01

Deformation of fuel particles and mass transfer from the transverse end of fuel meat toward the meat center was observed. This caused plate thickness peaking at a location between the meat edge and the meat center. The underlying mechanism for this fuel volume transport is believed to be fission induced creep of the U–Mo/Al meat. Fuel meat swelling was measured using optical microscopy images of the cross sections of the irradiated test plates. The time-dependent meat swelling was modeled for use in numerical simulation. A distinctive discrepancy between the predicted and measured meat thickness was found at the meat ends, which was assumed to be due to creep-induced mass relocation from the meat end to the meat center region that was not considered in the meat swelling model. ABAQUS FEA simulation was performed to reproduce the observed phenomenon at the meat ends. Through the simulation, we obtained the effective creep rate constants for the interaction layers (IL) and aluminum matrix. In addition, we obtained the corresponding stress and strain analysis results that can be used to understand mechanical behavior of U–Mo/Al dispersion fuel.

Mechanical analysis of UMo/Al dispersion fuel

Energy Technology Data Exchange (ETDEWEB)

Jeong, Gwan Yoon [Ulsan National Institute of Science and Technology, Department of Nuclear Engineering, 50 UNIST-gil, Eonyang-eup, Ulju-gun, Ulsan, 689-798 (Korea, Republic of); Kim, Yeon Soo [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Sohn, Dong-Seong, E-mail: dssohn@unist.ac.kr [Ulsan National Institute of Science and Technology, Department of Nuclear Engineering, 50 UNIST-gil, Eonyang-eup, Ulju-gun, Ulsan, 689-798 (Korea, Republic of)

2015-11-15

Deformation of fuel particles and mass transfer from the transverse end of fuel meat toward the meat center was observed. This caused plate thickness peaking at a location between the meat edge and the meat center. The underlying mechanism for this fuel volume transport is believed to be fission induced creep of the U–Mo/Al meat. Fuel meat swelling was measured using optical microscopy images of the cross sections of the irradiated test plates. The time-dependent meat swelling was modeled for use in numerical simulation. A distinctive discrepancy between the predicted and measured meat thickness was found at the meat ends, which was assumed to be due to creep-induced mass relocation from the meat end to the meat center region that was not considered in the meat swelling model. ABAQUS FEA simulation was performed to reproduce the observed phenomenon at the meat ends. Through the simulation, we obtained the effective creep rate constants for the interaction layers (IL) and aluminum matrix. In addition, we obtained the corresponding stress and strain analysis results that can be used to understand mechanical behavior of U–Mo/Al dispersion fuel.

Fate of dispersed marine fuel oil in sediment under pre-spill application strategy

International Nuclear Information System (INIS)

Jian Hua

2004-01-01

A comparison of the movement of dispersed oil in marine sediment under two dispersant application scenarios, applied prior to and after oil being spilled overboard, was examined. The pre-spill application scenario caused much less oil to be retained in the top sediment than post-spill scenario. The difference in oil retention in the top sediment between pre- and post-spill application scenario increased with increase in fuel oil temperature. For fuel oil above 40 o C, the difference in the effect of pre-spill application strategy under various water temperatures was negligible. When soap water was used as replacement for chemical dispersant, almost one-half as much oil was retained in the top sediment as that when using chemical dispersant. The adsorption of dispersed oil to the top sediment was almost proportionally decreased with doubling of soap dosage. (Author)

Powder fabrication of U-Mo alloys for nuclear dispersion fuels

Energy Technology Data Exchange (ETDEWEB)

Durazzo, Michelangelo; Rocha, Claudio Jose da; Mestnik Filho, Jose; Leal Neto, Ricardo Mendes, E-mail: mdurazzo@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2011-07-01

For the last 30 years high uranium density dispersion fuels have been developed in order to accomplish the low enrichment goals of the Reduced Enrichment for Research and Test Reactors (RERTR) Program. Gamma U-Mo alloys, particularly with 7 to 10 wt% Mo, as a fuel phase dispersed in aluminum matrix, have shown good results concerning its performance under irradiation tests. That's why this fissile phase is considered to be used in the nuclear fuel of the Brazilian Multipurpose Research Reactor (RMB), currently being designed. Powder production from these ductile alloys has been attained by atomization, mechanical (machining, grinding, cryogenic milling) and chemical (hydriding-de hydriding) methods. This work is a part of the efforts presently under way at IPEN to investigate the feasibility of these methods. Results on alloy fabrication by induction melting and gamma-stabilization of U-10Mo alloys are presented. Some results on powder production and characterization are also discussed. (author)

Powder fabrication of U-Mo alloys for nuclear dispersion fuels

International Nuclear Information System (INIS)

Durazzo, Michelangelo; Rocha, Claudio Jose da; Mestnik Filho, Jose; Leal Neto, Ricardo Mendes

2011-01-01

For the last 30 years high uranium density dispersion fuels have been developed in order to accomplish the low enrichment goals of the Reduced Enrichment for Research and Test Reactors (RERTR) Program. Gamma U-Mo alloys, particularly with 7 to 10 wt% Mo, as a fuel phase dispersed in aluminum matrix, have shown good results concerning its performance under irradiation tests. That's why this fissile phase is considered to be used in the nuclear fuel of the Brazilian Multipurpose Research Reactor (RMB), currently being designed. Powder production from these ductile alloys has been attained by atomization, mechanical (machining, grinding, cryogenic milling) and chemical (hydriding-de hydriding) methods. This work is a part of the efforts presently under way at IPEN to investigate the feasibility of these methods. Results on alloy fabrication by induction melting and gamma-stabilization of U-10Mo alloys are presented. Some results on powder production and characterization are also discussed. (author)

Qualification process of dispersion fuels in the IEAR1 research reactor

International Nuclear Information System (INIS)

Domingos, D.B.; Silva, A.T.; Silva, J.E.R.

2010-01-01

Neutronic, thermal-hydraulics and accident analysis calculations were developed to estimate the safety of a miniplate irradiation device (MID) to be placed in the IEA-R1 reactor core. The irradiation device will be used to receive miniplates of U 3 O 8 -Al and U 3 Si 2 -Al dispersion fuels, LEU type (19,9% of 235 U) with uranium densities of, respectively, 3.0 gU/cm 3 and 4.8 gU/cm 3 . The fuel miniplates will be irradiated to nominal 235 U burnup levels of 50% and 80%, in order to qualify the above high-density dispersion fuels to be used in the Brazilian Multipurpose Reactor (RMB), now in the conception phase. For the neutronic calculation, the computer codes CITATION and TWODB were utilized. The computer code FLOW was used to calculate the coolant flow rate in the irradiation device, allowing the determination of the fuel miniplate temperatures with the computer model MTRCR-IEA-R1. A postulated Loss of Coolant Accident (LOCA) was analyzed with the computer code LOSS and TEMPLOCA, allowing the calculation of the fuel miniplate temperatures after the reactor pool draining. This paper also presents a system designed for fuel swelling evaluation. The determination of the fuel swelling will be performed by means of the fuel miniplate thickness measurements along the irradiation time. (author)

A Prediction Study on Oxidation of Aluminum Alloy Cladding of U{sub 3}Si{sub 2}-Al Fuel Plate

Energy Technology Data Exchange (ETDEWEB)

Tahk, Y.W.; Lee, B.H.; Oh, J.Y.; Park, J.H.; Yim, J.S. [Research Reactor Design and Engineering Div., Korea Atomic Energy Research Institute, 1045 Daedeokdaero, Yuseong, Daejeon 305-353 (Korea, Republic of)

2011-07-01

U{sub 3}Si{sub 2}-Al dispersion fuel with aluminum alloy cladding will be used for the Jordan Research and Training Reactor (JRTR). Aluminum alloy cladding undergoes corrosion at slow rates under operational status. This causes thinning of the cladding walls and impairs heat transfer to the coolant. Predictions of the aluminum oxide thickness of the fuel cladding and the maximum temperature difference across the oxide film are needed for reliability evaluation based on the design criteria and limits which prohibit spallation of oxide film. In this work, several oxide thickness prediction models were compared with the measured data of in-pile test results from RERTR program. Moreover, specific parametric studies and a preliminary prediction of the aluminum alloy oxidation using the latest model were performed for JRTR fuel. According to the current JRTR fuel management scheme and operation strategy for 5 MW power, fresh fuel is discharged after 900 effective full power days (EFPD), which is too long a span to predict oxidation properly without an elaborate model. The latest model developed by Kim et al. is in good agreement with the recent in-pile test data as well as with the out-of-pile test data available in the literature, and is one of the best predictors for the oxidation of aluminum alloy cladding in various operating condition. Accordingly, this model was chosen for estimating the oxide film thickness. Through the preliminarily evaluation, water pH level is to be controlled lower than 6.2 for the conservativeness in the case of including the effect of anticipated operational occurrences and the spent fuel residence time in the storage rack after discharging. (author)

Development of oxide dispersion strengthened 2205 duplex stainless steel composite

Directory of Open Access Journals (Sweden)

Oladayo OLANIRAN

2015-05-01

Full Text Available Composites of duplex stainless steel were produced by oxide dispersion strengthening with comparatively improved mechanical properties by hot press sintering of partially stabilized Zirconia (PSZ, 3% yttria, mole fraction dispersion in 2205 duplex stainless steels. Ceramic oxide was added as reinforcement, while chromium (Cr and Nickel (Ni were incorporated to maintain the austenitic/ferritic phase balance of the duplex stainless steel. The powders and sintered were characterized in detail using scanning electron microscopy (SEM and X-ray diffraction (XRD. The microstructural evolution and phase formation during oxide dispersion strengthening of duplex stainless steel composites were investigated. The influence of composition variation of the reinforcements on the microstructural and corrosion behaviour in simulated mine water of the composites were investigated. In this manuscript, it was established that composition has great influence on the structure/properties relationship of the composites developed.

The burnup dependence of light water reactor spent fuel oxidation

International Nuclear Information System (INIS)

Hanson, B.D.

1998-07-01

Over the temperature range of interest for dry storage or for placement of spent fuel in a permanent repository under the conditions now being considered, UO 2 is thermodynamically unstable with respect to oxidation to higher oxides. The multiple valence states of uranium allow for the accommodation of interstitial oxygen atoms in the fuel matrix. A variety of stoichiometric and nonstoichiometric phases is therefore possible as the fuel oxidizers from UO 2 to higher oxides. The oxidation of UO 2 has been studied extensively for over 40 years. It has been shown that spent fuel and unirradiated UO 2 oxidize via different mechanisms and at different rates. The oxidation of LWR spent fuel from UO 2 to UO 2.4 was studied previously and is reasonably well understood. The study presented here was initiated to determine the mechanism and rate of oxidation from UO 2.4 to higher oxides. During the early stages of this work, a large variability in the oxidation behavior of samples oxidized under nearly identical conditions was found. Based on previous work on the effect of dopants on UO 2 oxidation and this initial variability, it was hypothesized that the substitution of fission product and actinide impurities for uranium atoms in the spent fuel matrix was the cause of the variable oxidation behavior. Since the impurity concentration is roughly proportional to the burnup of a specimen, the oxidation behavior of spent fuel was expected to be a function of both temperature and burnup. This report (1) summarizes the previous oxidation work for both unirradiated UO 2 and spent fuel (Section 2.2) and presents the theoretical basis for the burnup (i.e., impurity concentration) dependence of the rate of oxidation (Sections 2.3, 2.4, and 2.5), (2) describes the experimental approach (Section 3) and results (Section 4) for the current oxidation tests on spent fuel, and (3) establishes a simple model to determine the activation energies associated with spent fuel oxidation (Section 5)

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

Discrete element method study of fuel relocation and dispersal during loss-of-coolant accidents

International Nuclear Information System (INIS)

Govers, K.; Verwerft, M.

2016-01-01

The fuel fragmentation, relocation and dispersal (FFRD) during LOCA transients today retain the attention of the nuclear safety community. The fine fragmentation observed at high burnup may, indeed, affect the Emergency Core Cooling System performance: accumulation of fuel debris in the cladding ballooned zone leads to a redistribution of the temperature profile, while dispersal of debris might lead to coolant blockage or to debris circulation through the primary circuit. This work presents a contribution, by discrete element method, towards a mechanistic description of the various stages of FFRD. The fuel fragments are described as a set of interacting particles, behaving as a granular medium. The model shows qualitative and quantitative agreement with experimental observations, such as the packing efficiency in the balloon, which is shown to stabilize at about 55%. The model is then applied to study fuel dispersal, for which experimental parametric studies are both difficult and expensive. - Highlights: • We performed Discrete Element Methods simulation for fuel relocation and dispersal during LOCA transients. • The approach provides a mechanistic description of these phenomena. • The approach shows the ability of the technique to reproduce experimental observations. • The packing fraction in the balloon is shown to stabilize at 50–60%.

Discrete element method study of fuel relocation and dispersal during loss-of-coolant accidents

Science.gov (United States)

Govers, K.; Verwerft, M.

2016-09-01

The fuel fragmentation, relocation and dispersal (FFRD) during LOCA transients today retain the attention of the nuclear safety community. The fine fragmentation observed at high burnup may, indeed, affect the Emergency Core Cooling System performance: accumulation of fuel debris in the cladding ballooned zone leads to a redistribution of the temperature profile, while dispersal of debris might lead to coolant blockage or to debris circulation through the primary circuit. This work presents a contribution, by discrete element method, towards a mechanistic description of the various stages of FFRD. The fuel fragments are described as a set of interacting particles, behaving as a granular medium. The model shows qualitative and quantitative agreement with experimental observations, such as the packing efficiency in the balloon, which is shown to stabilize at about 55%. The model is then applied to study fuel dispersal, for which experimental parametric studies are both difficult and expensive.

Study on characteristics of U-Mo/Al-Si interaction layers of dispersion fuel plates

International Nuclear Information System (INIS)

Liu Lijian; Yin Changgeng; Chen Jiangang; Sun Changlong; Liu Yunming

2014-01-01

In this paper, we analyzed the characteristics of U-Mo/Al-Si interaction layers of dispersion fuel plates. The results show that the interaction layers (IL) are with irregular morphology and uneven thickness, and are mainly formed in the internal micro cracks of the dispersion fuel particles or at the interface between the particles and the substrates. The diffusion mechanism of U-Mo/Al-Si is the vacancy diffusion, Al and Si are migrating elements, and the diffusion reaction is that Al and Si diffuse to U-Mo alloy. Inside the interaction layers, the Al content keeps constant basically, but the Si content gradually increases with the substrate-fuel direction, and the maximum content of Si appears interaction layers near the U-Mo side. Adding about 5 wt% Si into Al matrix can restrain the diffusion reaction, and improve the performance of dispersion fuel plates finally. (authors)

An oxidation-resistant indium tin oxide catalyst support for proton exchange membrane fuel cells

Energy Technology Data Exchange (ETDEWEB)

Chhina, H.; Campbell, S. [Ballard Power Systems Inc., 9000 Glenlyon Parkway, Burnaby, BC V5J 5J8 (Canada); Kesler, O. [Department of Mechanical Engineering, University of British Columbia, Vancouver, BC, V6T 1Z4 (Canada)

2006-10-27

The oxidation of carbon catalyst supports causes degradation in catalyst performance in proton exchange membrane fuel cells (PEMFCs). Indium tin oxide (ITO) is considered as a candidate for an alternative catalyst support. The electrochemical stability of ITO was studied by use of a rotating disk electrode (RDE). Oxidation cycles between +0.6 and +1.8V were applied to ITO supporting a Pt catalyst. Cyclic voltammograms (CVs) both before and after the oxidation cycles were obtained for Pt on ITO, Hispec 4000 (a commercially available catalyst), and 40wt.% Pt dispersed in-house on Vulcan XC-72R. Pt on ITO showed significantly better electrochemical stability, as determined by the relative change in electrochemically active surface area after cycling. Hydrogen desorption peaks in the CVs existed even after 100 cycles from 0.6 to 1.8V for Pt on ITO. On the other hand, most of the active surface area was lost after 100 cycles of the Hispec 4000 catalyst. The 40wt.% Pt on Vulcan made in-house also lost most of its active area after only 50 cycles. Pt on ITO was significantly more electrochemically stable than both Hispec 4000 and Pt on Vulcan XC-72R. In this study, it was found that the Pt on ITO had average crystallite sizes of 13nm for Pt and 38nm for ITO. Pt on ITO showed extremely high thermal stability, with only {approx}1wt.% loss of material for ITO versus {approx}57wt.% for Hispec 4000 on heating to 1000{sup o}C. The TEM data show Pt clusters dispersed on small crystalline ITO particles. The SEM data show octahedral shaped ITO particles supporting Pt. (author)

Possibilities of practical usage of dispersed aluminim oxidation by liquid water

Science.gov (United States)

Larichev, M. N.; Laricheva, O. O.; Shaitura, N. S.; Shkolnikov, E. I.

2012-12-01

The goal of this work is to show the possibility of practical usage of the environmentally pure oxidation process of preliminarily dispersed aluminum (aluminum powders of the ASD or PAD grade according to TU (Technical Specifications) 48-5-226-87, which are serially produced in industry) with liquid water in order to obtain gaseous hydrogen in volumes sufficient to provide the operation of energizers based on airhydrogen fuel cells (AHFC) for portable and stationary devices (up to 3 kW). It is shown that the synthesis of aluminum oxides-hydroxides with the specified phase and chemical compositions as well as the particle shape and size can be provided simultaneously with producing commercial hydrogen. The practical usage of hydrogen, which is formed in the oxidation reaction of metallic aluminum with liquid water at pressures close to atmospheric (particularly, to service AHFCs), requires reaction intensification to increase the oxidation rate of aluminum. In this work, we considered the aspects of practical implementation of thermal, ultrasonic, and chemical activation as well as their combinations for this purpose. As the chemical activator of oxidation, we used the additives of calcium oxide (<5% of the mass of oxidized aluminum). Application of each activation method affects the phase and chemical compositions as well as the structure of formed aluminum hydroxides, which provides the possibility of their reproducible production. Thus, simultaneously with the production of commercial hydrogen, solid oxidation products satisfying the needs of industry in aluminum oxides and having the specified composition, purity, and particle shape and size can be synthesized. The acquired experimental results and elements of the design of specially developed industrial apparatuses, which were used when performing this work, can be applied when designing the model of the hydrogen generator—the prototype of the hydrogen generator for portable and stationary devices or devices

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

International Nuclear Information System (INIS)

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

1969-01-01

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

Solid Oxide Fuel Cells Operating on Alternative and Renewable Fuels

Energy Technology Data Exchange (ETDEWEB)

Wang, Xiaoxing; Quan, Wenying; Xiao, Jing; Peduzzi, Emanuela; Fujii, Mamoru; Sun, Funxia; Shalaby, Cigdem; Li, Yan; Xie, Chao; Ma, Xiaoliang; Johnson, David; Lee, Jeong; Fedkin, Mark; LaBarbera, Mark; Das, Debanjan; Thompson, David; Lvov, Serguei; Song, Chunshan

2014-09-30

This DOE project at the Pennsylvania State University (Penn State) initially involved Siemens Energy, Inc. to (1) develop new fuel processing approaches for using selected alternative and renewable fuels – anaerobic digester gas (ADG) and commercial diesel fuel (with 15 ppm sulfur) – in solid oxide fuel cell (SOFC) power generation systems; and (2) conduct integrated fuel processor – SOFC system tests to evaluate the performance of the fuel processors and overall systems. Siemens Energy Inc. was to provide SOFC system to Penn State for testing. The Siemens work was carried out at Siemens Energy Inc. in Pittsburgh, PA. The unexpected restructuring in Siemens organization, however, led to the elimination of the Siemens Stationary Fuel Cell Division within the company. Unfortunately, this led to the Siemens subcontract with Penn State ending on September 23rd, 2010. SOFC system was never delivered to Penn State. With the assistance of NETL project manager, the Penn State team has since developed a collaborative research with Delphi as the new subcontractor and this work involved the testing of a stack of planar solid oxide fuel cells from Delphi.

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

Evolution of dispersion fuel meat structure caused by interface reaction

International Nuclear Information System (INIS)

Xing Zhonghu; Ying Shihao

2000-01-01

In reactor operation, the resultant layers are formed by interdiffusion at the fuel particle-matrix interfaces of U 3 Si 2 -Al dispersion fuel. This results in the evolution of meat structure. On the basis of Monte-Carlo method, the author developed simulation method of fuel meat, and simulated the stochastic space locations of spherical fuel particles in the meat. The fuel volume fraction is 43%, and the particles are in definite size distribution. For the 13551 simulated particle samples, the evolution of meat structure is calculated with layer thickness ranging from 0 to 16 μm. The parameters of meat structure include the U 3 Si 2 fuel volume fraction, resultant layer volume fraction, Al matrix volume fraction, particle contact probability and overlap degree as functions of layer thickness

Release behavior of fission products from irradiated dispersion fuels at high temperatures

International Nuclear Information System (INIS)

Iwai, Takashi; Shimizu, Michio; Nakagawa, Tetsuya

1990-02-01

As a framework of reduced enrichment fuel program of JMTR Project, the measurements of fission products release rates at high temperatures (600degC - 1100degC) were performed in order to take the data to use for safety evaluation of LEU fuel. Three type miniplates of dispersion silicide and aluminide fuel, 20% enrichment LEU fuel with 4.8 gU/cc (U 3 Si 2 90 %, USi 10 % and U 3 Si 2 50 %, U 3 Si 50 % dispersed in aluminium) and 45 % enrichment MEU fuel with 1.6 gU/cc, were irradiated in JMTR. The burnups attained by one cycle (22 days) irradiation were within 21.6 % - 22.5 % of initial 235 U. The specimens cut down from miniplates were measured on fission products release rates by means of new apparatus specially designed for this experiment. The specimens were heated up within 600degC - 1100degC in dry air. Then fission products such as 85 Kr, 133 Xe, 131 I, 137 Cs, 103 Ru, 129m Te were collected at each temperature and measured on release rates. In the results of measurement, the release rates of 85 Kr, 133 Xe, 131 I, 129m Te from all specimens were slightly less than that of G.W. Parker's data on U-Al alloy fuel. For 137 Cs and 103 Ru from a silicide specimen (U 3 Si 2 90 %, USi 10 % dispersed in aluminium) and 137 Cs from an aluminide specimen, the release rates were slightly higher than that of G.W. Parker's. (author)

Dispersion strengthening of aluminium-aluminium-oxide products

DEFF Research Database (Denmark)

Hansen, Niels

1970-01-01

The true stress-true strain curves at room temperature and at 400°C were determined for various types of aluminium-aluminium-oxide products containing from 0.2 to 4.7 weight per cent of aluminium oxide. The effect of particles on the initial flow stress and the flow stress for 0.2% offset at room...... temperature and at 400°C is in agreement with Orowan's theory. The increase in flow stress at room temperature for strain values below 3 per cent was related to the plastic strain by the equation σ-σoy=k1ε 1/2, where σoy is the initial flow stress and where k1 increases for increasing volume fraction...... and decreasing particle size of the dispersed particles. A general expression for k1 was derived for the relationship between the dislocation density and the strain in dispersion-strengthened products...

Modélisation de la combustion de fuels lourds prenant en compte la dispersion des asphaltènes Modeling Heavy Fuel-Oil Combustion (While Considering Or Including Asphaltene Dispersion

Directory of Open Access Journals (Sweden)

Audibert F.

2006-11-01

Full Text Available Divers modèles, ayant pour but de prédire le taux d'imbrûlés solides lors de la combustion du fuel lourd, ont été mis au point dans le passé. Les paramètres entrant en ligne de compte sont le plus souvent les teneurs en résidus lourds hydrocarbonés (asphaltènes précipités au pentane ou à l'heptane et carbone Conradson et en métaux : c'est le cas des modèles Exxon et Shell développés respectivement en 1979 et 1981. D'autres modèles tiennent compte, en plus de la composition du fuel, de son mode d'atomisation, de son mode de diffusion dans le foyer et de la cinétique de combustion : on peut citer les travaux du Laboratoire Energie du MIT publiés en 1986. Néanmoins, ces facteurs ne sont pas les seuls à intervenir : l'expérience a montré que l'état de dispersion des asphaltènes peut jouer également un grand rôle, notamment dans le cas d'installations de combustion à injection mécanique, pour lesquelles la dispersion des gouttelettes n'est pas aussi fine que pour des installations munies d'une injection assistée par la vapeur. Cette influence de la dispersion des asphaltènes sur la combustion a été mise en évidence dans le passé par l'utilisation d'additifs dispersants et également par la combustion de fuels lourds constitués par dilution d'asphaltes précipités au pentane avec un gas-oil de cracking catalytique de raffinerie (LCO. Ce sont ces fuels que l'on a considérés dans la présente étude. L'effet de ce facteur dispersion n'a pas été quantifié jusqu'alors, la difficulté étant de définir une grandeur mesurable représentant la répartition des agglomérats d'asphaltènes. Dans cette étude, on a essayé en un premier temps de faire une approche fractale de la répartition des asphaltènes à partir de clichés (préparés par la société Total, cette méthode ayant déjà été utilisée avec succès pour décrire des structures d'aspects comparables. Malheureusement, on s'est heurté à des

History of Resistance Welding Oxide Dispersion Strengthened Cladding and other High Temperature Materials at Center for Advanced Energy Studies

International Nuclear Information System (INIS)

Zirker, Larry; Jerred, Nathan; Charit, Indrajit; Cole, James

2012-01-01

Research proposal 08-1079, 'A Comparative Study of Welded ODS Cladding Materials for AFCI/GNEP,' was funded in 2008 under an Advanced Fuel Cycle Initiative (AFCI) Research and Development Funding Opportunity, number DE-PS07-08ID14906. Th proposal sought to conduct research on joining oxide dispersion strengthen (ODS) tubing material to a solid end plug. This document summarizes the scientific and technical progress achieved during the project, which ran from 2008 to 2011.

An investigation on fuel meats extruded with atomized U-10wt% Mo powder for uranium high-density dispersion fuel

International Nuclear Information System (INIS)

Kim, Chang-Kyu; Kim, Ki-Hwan; Park, Jong-Man; Lee, Don-Bae; Sohn, Dong-Seong

1997-01-01

The RERTR program has been making an effort to develop dispersion fuels with uranium densities of 8 to 9 g U/cm3 for research and test reactors. Using atomized U-10wt%Mo powder, fuel meats have been fabricated successfully up to 55 volume % of fuel powder. The uranium density of an extruded meat with a 55 volume % of fuel powder was obtained to be 7.7 g/cm3. A relatively high porosity of 7.3% was formed due to cracking of particles, presumably induced by the impingement among agglomerated particles. Tensile test results indicated that the strength of fuel meats with 55% volume fraction decreased some and a little of ductility was maintained. Examination on the fracture surface revealed that some U-10%Mo particles appeared to be broken by the tensile force in brittle rupture mode. The increase of broken particles in high fuel fraction is considered to be induced mainly by the impingement among agglomerated particles. Uranium loading density is assumed to be improved through the development of the better homogeneous dispersion technology. (author)

Progress in qualifying low-enriched U-Mo dispersion fuels

International Nuclear Information System (INIS)

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

2001-01-01

The U.S. Reduced Enrichment for Research and Test Reactors program is working to qualify dispersions of U-Mo alloys in aluminum with fuel-meat densities of 8 to 9 gU cm -3 . Post irradiation examinations of the small fuel plates irradiated in the Advanced Test Reactor during the high-temperature RERTR-3 tests are virtually complete, and analysis of the large quantity of data obtained is underway. We have observed that the swelling of the fuel plates is stable and modest and that the swelling is dominated by the temperature-dependent interaction of the U-Mo fuel and the aluminum matrix. In order to extract detailed information about the behavior of these fuels from the data, a complex fuel-plate thermal model is being developed to account for the effects of the changing fission rate and thermal conductivity of the fuel meat during irradiation. This paper summarizes the empirical results of the post irradiation examinations and the preliminary results of the model development. In addition, the schedule for irradiation of full-sized elements in the HFR-Petten is briefly discussed. (author)

Advanced methods of solid oxide fuel cell modeling

CERN Document Server

Milewski, Jaroslaw; Santarelli, Massimo; Leone, Pierluigi

2011-01-01

Fuel cells are widely regarded as the future of the power and transportation industries. Intensive research in this area now requires new methods of fuel cell operation modeling and cell design. Typical mathematical models are based on the physical process description of fuel cells and require a detailed knowledge of the microscopic properties that govern both chemical and electrochemical reactions. ""Advanced Methods of Solid Oxide Fuel Cell Modeling"" proposes the alternative methodology of generalized artificial neural networks (ANN) solid oxide fuel cell (SOFC) modeling. ""Advanced Methods

Irradiation behavior of uranium-silicide dispersion fuels

International Nuclear Information System (INIS)

Hofman, G.L.; Neimark, L.A.

1984-01-01

This paper describes and analyzes the irradiation behavior of experimental fuel plates containing U 3 Si, U 3 Si-1.5 w/o Al, and U 3 Si 2 particulate fuel dispersed and clad in aluminum. The fuel is nominally 19.9%-enriched 235 U and the fuel volume fraction in the central ''meat'' section of the plates is approximately 33%. Sets of fuel plates were removed from the Oak Ridge Research reactor at burnup levels of 35, 83, and 94% 235 U depletion and examined at the Alpha-Gamma Hot-Cell Facility at Argonne National Laboratory. The results of the examination may be summarized as follows. The dimensional stability of the U 3 Si 2 and pure U 3 Si fuel was excellent throughout the entire burnup range, with uniform plate thickness increases up to a maximum of 4 mils at the highest burnup level (94% 235 U depletion). This corresponds to a meat volume increase of 11%. The swelling was partially due to solid fission products but to a larger extent to fission gas bubbles. The fission gas bubbles in U 3 Si 2 were small (submicrometer size) and very uniformly distributed, indicating great stability. To a large extent this was also the case for U 3 Si; however, larger bubbles ( 3 Si-1.5 w/o Al fuel became unstable at the higher burnup levels. Fission gas bubbles were larger than in the other two fuels and were present throughout the fuel particles. At 94% 235 U depletion, the formation of fission gas bubbles with diameters up to 20 mils caused the plates to pillow. It is proposed that aluminum in U 3 Si destabilizes fission gas bubble formation to the point of severe breakaway swelling in the prealloyed silicide fuel. (author)

Behavior of molybdenum in mixed-oxide fuel

International Nuclear Information System (INIS)

Giacchetti, G.; Sari, C.

1976-01-01

Metallic molybdenum, Mo--Ru--Rh--Pd alloys, barium, zirconium, and tungsten were added to uranium and uranium--plutonium oxides by coprecipitation and mechanical mixture techniques. This material was treated in a thermal gradient similar to that existing in fuel during irradiation to study the behavior of molybdenum in an oxide matrix as a function of the O/(U + Pu) ratio and some added elements. Result of ceramographic and microprobe analysis shows that when the overall O/(U + Pu) ratio is less than 2, molybdenum and Mo--Ru--Rh--Pd alloy inclusions are present in the uranium--plutonium oxide matrix. If the O/(U + Pu) ratio is greater than 2, molybdenum oxidizes to MoO 2 , which is gaseous at a temperature approximately 1000 0 C. Molybdenum oxide vapor reacts with barium oxide and forms a compound that exists as a liquid phase in the columnar grain region. Molybdenum oxide also reacts with tungsten oxide (tungsten is often present as an impurity in the fuel) and forms a compound that contains approximately 40 wt percent of actinide metals. The apparent solubility of molybdenum in uranium and uranium--plutonium oxides, determined by electron microprobe, was found to be less than 250 ppM both for hypo- and hyperstoichiometric fuels

Well-dispersed NiO nanoparticles supported on nitrogen-doped carbon nanotube for methanol electrocatalytic oxidation in alkaline media

Energy Technology Data Exchange (ETDEWEB)

Wang, Pengcheng; Zhou, Yingke, E-mail: zhouyk888@hotmail.com; Hu, Min; Chen, Jian

2017-01-15

Highlights: • Nitrogen-doped carbon nanotube supporting ultrafine NiO nanoparticles with high dispersity are facile synthesized. • The nitrogen doping, calcination temperature and NiO loading present great effects on the catalyst morphology, structure and electrochemical performance. • NiO-NCNT-3x-400 demonstrates remarkable catalytic activity and stability for the methanol electrolytic oxidation reaction. - Abstract: Nitrogen-doped carbon nanotube supporting NiO nanoparticles were synthesized by a chemical precipitation process coupled with subsequent calcination. The morphology and structure of the composites were characterized by transmission electron microscopy (TEM), X-ray diffraction (XRD) and X-ray photoelectron spectroscopy (XPS), and the electrochemical performance was evaluated using cyclic voltammetry and chronoamperometric technique. The effects of nitrogen doping, calcination temperature and content of NiO nanoparticles on the electrocatalytic activity toward methanol oxidation were systematically studied. The results show that the uniformly dispersed ultrafine NiO nanoparticles supported on nitrogen-doped carbon nanotube are obtained after calcination at 400 °C. The optimized composite catalysts present high electrocatalytic activity, fast charge-transfer process, excellent accessibility and stability for methanol oxidation reaction, which are promising for application in the alkaline direct methanol fuel cells.

Irradiation behavior of U 6Mn-Al dispersion fuel elements

Science.gov (United States)

Meyer, M. K.; Wiencek, T. C.; Hayes, S. L.; Hofman, G. L.

2000-02-01

Irradiation testing of U 6Mn-Al dispersion fuel miniplates was conducted in the Oak Ridge Research Reactor (ORR). Post-irradiation examination showed that U 6Mn in an unrestrained plate configuration performs similarly to U 6Fe under irradiation, forming extensive and interlinked fission gas bubbles at a fission density of approximately 3×10 27 m-3. Fuel plate failure occurs by fission gas pressure driven `pillowing' on continued irradiation.

Development of PRIME for irradiation performance analysis of U-Mo/Al dispersion fuel

Science.gov (United States)

Jeong, Gwan Yoon; Kim, Yeon Soo; Jeong, Yong Jin; Park, Jong Man; Sohn, Dong-Seong

2018-04-01

A prediction code for the thermo-mechanical performance of research reactor fuel (PRIME) has been developed with the implementation of developed models to analyze the irradiation behavior of U-Mo dispersion fuel. The code is capable of predicting the two-dimensional thermal and mechanical performance of U-Mo dispersion fuel during irradiation. A finite element method was employed to solve the governing equations for thermal and mechanical equilibria. Temperature- and burnup-dependent material properties of the fuel meat constituents and cladding were used. The numerical solution schemes in PRIME were verified by benchmarking solutions obtained using a commercial finite element analysis program (ABAQUS). The code was validated using irradiation data from RERTR, HAMP-1, and E-FUTURE tests. The measured irradiation data used in the validation were IL thickness, volume fractions of fuel meat constituents for the thermal analysis, and profiles of the plate thickness changes and fuel meat swelling for the mechanical analysis. The prediction results were in good agreement with the measurement data for both thermal and mechanical analyses, confirming the validity of the code.

Physicochemical analysis of interaction of oxide fuel with pyrocarbon coatings of fuel particles

International Nuclear Information System (INIS)

Lyutikov, R.A.; Khromov, Yu.F.; Chernikov, A.S.

1990-01-01

Equilibrium pressure of (CO+Kr,Xe) gases inside fuel particle with oxide kern depending on design features of fuel particle, on temperature. on (O/U) initial composition and fuel burnup is calculated using the suggested model. Analysis of possibility for gas pressure reduction by means of uranium carbide alloying of kern and degree increase of solid fission product retention (Cs for example) during alumosilicate alloying of uranium oxide is conducted

Phase stability of oxide dispersion-strengthened ferritic steels in neutron irradiation

International Nuclear Information System (INIS)

Yamashita, S.; Oka, K.; Ohnuki, S.; Akasaka, N.; Ukai, S.

2002-01-01

Oxide dispersion-strengthened ferritic steels were irradiated by neutrons up to 21 dpa and studied by microstructural observation and microchemical analysis. The original high dislocation density did not change after neutron irradiation, indicating that the dispersed oxide particles have high stability under neutron irradiation. However, there is potential for recoil resolution of the oxide particles due to ballistic ejection at high dose. From the microchemical analysis, it was implied that some of the complex oxides have a double-layer structure, such that TiO 2 occupied the core region and Y 2 O 3 the outer layer. Such a structure may be more stable than the simple mono-oxides. Under high-temperature irradiation, Laves phase was the predominant precipitate occurring at grain boundaries α phase and χ phase were not observed in this study

Forging Oxide-Dispersion-Strengthened Superalloys

Science.gov (United States)

Harf, F. H.; Glasgow, T. K.; Moracz, D. J.; Austin, C. M.

1986-01-01

Cladding of mild steel prevents surface cracking when alloy contacts die. Continual need for improvements in properties of alloys capable of withstanding elevated temperatures. Accomplished by using oxide-dispersion-strengthed superalloys such as Inconel Alloy MA 6000. Elevated tensile properties of forged alloy equal those of hot-rolled MA 6000 bar. Stress-rupture properties somewhat lower than those of bar stock but, at 1,100 degrees C, exceed those of strongest commercial single crystal, directionally solidified and conventionally cast superalloys.

History of Resistance Welding Oxide Dispersion Strengthened Cladding and other High Temperature Materials at Center for Advanced Energy Studies

Energy Technology Data Exchange (ETDEWEB)

Larry Zirker; Nathan Jerred; Dr. Indrajit Charit; James Cole

2012-03-01

Research proposal 08-1079, 'A Comparative Study of Welded ODS Cladding Materials for AFCI/GNEP,' was funded in 2008 under an Advanced Fuel Cycle Initiative (AFCI) Research and Development Funding Opportunity, number DE-PS07-08ID14906. Th proposal sought to conduct research on joining oxide dispersion strengthen (ODS) tubing material to a solid end plug. This document summarizes the scientific and technical progress achieved during the project, which ran from 2008 to 2011.

Development of examination technique for oxide layer thickness measurement of irradiated fuel rods

International Nuclear Information System (INIS)

Koo, D. S.; Park, S. W.; Kim, J. H.; Seo, H. S.; Min, D. K.; Kim, E. K.; Chun, Y. B.; Bang, K. S.

1999-06-01

Technique for oxide layer thickness measurement of irradiated fuel rods was developed to measure oxide layer thickness and study characteristic of fuel rods. Oxide layer thickness of irradiated fuels were measured, analyzed. Outer oxide layer thickness of 3 cycle-irradiated fuel rods were 20 - 30 μm, inner oxide layer thickness 0 - 10 μm and inner oxide layer thickness on cracked cladding about 30 μm. Oxide layer thickness of 4 cycle-irradiated fuel rods were about 2 times as thick as those of 1 cycle-irradiated fuel rods. Oxide layer on lower region of irradiated fuel rods was thin and oxide layer from lower region to upper region indicated gradual increase in thickness. Oxide layer thickness from 2500 to 3000 mm showed maximum and oxide layer thickness from 3000 to top region of irradiated fuel rods showed decreasing trend. Inner oxide layer thicknesses of 4 cycle-irradiated fuel rod were about 8 μm at 750 - 3500 mm from the bottom end of fuel rod. Outer oxide layer thickness were about 8 μm at 750 - 1000 mm from the bottom end of fuel rod. These indicated gradual increase up to upper region from the bottom end of fuel rod. These indicated gradual increase up to upper region from the bottom end of fuel. Oxide layer thickness technique will apply safety evaluation and study of reactor fuels. (author). 6 refs., 14 figs

Spent fuel. Dissolution and oxidation

International Nuclear Information System (INIS)

Grambow, B.

1989-03-01

Data from studies of the low temperature air oxidation of spent fuel were retrieved in order to provide a basis for comparison between the mechanism of oxidation in air and corrosion in water. U 3 O 7 is formed by diffusion of oxygen into the UO 2 lattice. A diffusion coefficient of oxygen in the fuel matric was calculated for 25 degree C to be in the range of 10 -23 to 10 -25 m 2 /s. The initial rates of U release from spent fuel and from UO 2 appear to be similar. The lowest rates (at 25 degree c >10 -4 g/(m 2 d)) were observed under reducing conditions. Under oxidizing conditions the rates depend mainly of the nature and concentraion of the oxidant and/or on corbonate. In contact with air, typical initial rates at room temperature were in the range between 0.001 and 0.1 g/(m 2 d). A study of apparent U solubility under oxidizing conditions was performed and it was suggested that the controlling factor is the redox potential at the UO 2 surface rather than the E h of the bulk solution. Electrochemical arguments were used to predict that at saturation, the surface potential will eventually reach a value given by the boundaries at either the U 3 O 7 /U 3 O 8 or the U 3 O 7 /schoepite stability field, and a comparison with spent fuel leach data showed that the solution concentration of uranium is close to the calculated U solubility at the U 3 O 7 /U 3 O 8 boundary. The difference in the cumulative Sr and U release was calculated from data from Studsvik laboratory. The results reveal that the rate of Sr release decreases with the square root of time under U-saturated conditions. This time dependence may be rationalized either by grain boundary diffusion or by diffusion into the fuel matrix. Hence, there seems to be a possibility of an agreement between the Sr release data, structural information and data for oxygen diffusion in UO 2 . (G.B.)

Pore growth in U-Mo/Al dispersion fuel

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); Jeong, G.Y.; Sohn, D.-S. [Ulsan National Institute of Science and Technology, 50 UNIST-gil, Eonyang-eup, Ulju-gun, Ulsan, 689-798 (Korea, Republic of); Jamison, L.M. [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States)

2016-09-15

U-Mo/Al dispersion fuel is currently under development in the DOE’s Material Management and Minimization program to convert HEU-fueled research reactors to LEU-fueled reactors. In some demanding conditions in high-power and high-performance reactors, large pores form in the interaction layers between the U-Mo fuel particles and the Al matrix, which pose a potential to cause fuel failure. In this study, comprehension of the formation and growth of these pores was explored. As a product, a model to predict pore growth and porosity increase was developed. The model includes three major topics: fission gas release from the U-Mo and the IL to the pores, stress evolution in the fuel meat, and the effect of amorphous IL growth. Well-characterized in-pile data from reduced-size plates were used to fit the model parameters. A data set from full-sized plates, independent and distinctively different from those used to fit the model parameters, was used to examine the accuracy of the model. The model showed fair agreement with the measured data. The model suggested that the growth of the IL has a critical effect on pore growth, as both its material properties and energetics are favorable to pore formation. Therefore, one area of the current effort, focused on suppressing IL growth, appears to be on the right track to improve the performance of this fuel.

Alternative anode materials for solid oxide fuel cells

Energy Technology Data Exchange (ETDEWEB)

Goodenough, John B.; Huang, Yun-Hui [Texas Materials Institute, ETC 9.102, 1 University Station, C2200, The University of Texas at Austin, Austin, TX 78712 (United States)

2007-11-08

The electrolyte of a solid oxide fuel cell (SOFC) is an O{sup 2-}-ion conductor. The anode must oxidize the fuel with O{sup 2-} ions received from the electrolyte and it must deliver electrons of the fuel chemisorption reaction to a current collector. Cells operating on H{sub 2} and CO generally use a porous Ni/electrolyte cermet that supports a thin, dense electrolyte. Ni acts as both the electronic conductor and the catalyst for splitting the H{sub 2} bond; the oxidation of H{sub 2} to H{sub 2}O occurs at the Ni/electrolyte/H{sub 2} triple-phase boundary (TPB). The CO is oxidized at the oxide component of the cermet, which may be the electrolyte, yttria-stabilized zirconia, or a mixed oxide-ion/electron conductor (MIEC). The MIEC is commonly a Gd-doped ceria. The design and fabrication of these anodes are evaluated. Use of natural gas as the fuel requires another strategy, and MIECs are being explored for this application. The several constraints on these MIECs are outlined, and preliminary results of this on-going investigation are reviewed. (author)

Interim results from UO2 fuel oxidation tests in air

International Nuclear Information System (INIS)

Campbell, T.K.; Gilbert, E.R.; Thornhill, C.K.; White, G.D.; Piepel, G.F.; Griffin, C.W.j.

1987-08-01

An experimental program is being conducted at Pacific Northwest Laboratory (PNL) to extend the characterization of spent fuel oxidation in air. To characterize oxidation behavior of irradiated UO 2 , fuel oxidation tests were performed on declad light-water reactor spent fuel and nonirradited UO 2 pellets in the temperature range of 135 to 250 0 C. These tests were designed to determine the important independent variables that might affect spent fuel oxidation behavior. The data from this program, when combined with the test results from other programs, will be used to develop recommended spent fuel dry-storage temperature limits in air. This report describes interim test results. The initial PNL investigations of nonirradiated and spent fuels identified the important testing variables as temperature, fuel burnup, radiolysis of the air, fuel microstructure, and moisture in the air. Based on these initial results, a more extensive statistically designed test matrix was developed to study the effects of temperature, burnup, and moisture on the oxidation behavior of spent fuel. Oxidation tests were initiated using both boiling-water reactor and pressurized-water reactor fuels from several different reactors with burnups from 8 to 34 GWd/MTU. A 10 5 R/h gamma field was applied to the test ovens to simulate dry storage cask conditions. Nonirradiated fuel was included as a control. This report describes experimental results from the initial tests on both the spent and nonirradiated fuels and results to date on the tests in a 10 5 R/h gamma field. 33 refs., 51 figs., 6 tabs

Aluminum cladding oxidation of prefilmed in-pile fueled experiments

Energy Technology Data Exchange (ETDEWEB)

Marcum, W.R., E-mail: marcumw@engr.orst.edu [Oregon State University, School of Nuclear Science and Engineering, 116 Radiation Center, Corvallis, OR 97331 (United States); Wachs, D.M.; Robinson, A.B.; Lillo, M.A. [Idaho National Laboratory, Nuclear Fuels & Materials Department, 2525 Fremont Ave., Idaho Falls, ID 83415 (United States)

2016-04-01

A series of fueled irradiation experiments were recently completed within the Advanced Test Reactor Full size plate In center flux trap Position (AFIP) and Gas Test Loop (GTL) campaigns. The conduct of the AFIP experiments supports ongoing efforts within the global threat reduction initiative (GTRI) to qualify a new ultra-high loading density low enriched uranium-molybdenum fuel. This study details the characterization of oxide growth on the fueled AFIP experiments and cross-correlates the empirically measured oxide thickness values to existing oxide growth correlations and convective heat transfer correlations that have traditionally been utilized for such an application. This study adds new and valuable empirical data to the scientific community with respect to oxide growth measurements of highly irradiated experiments, of which there is presently very limited data. Additionally, the predicted oxide thickness values are reconstructed to produce an oxide thickness distribution across the length of each fueled experiment (a new application and presentation of information that has not previously been obtainable in open literature); the predicted distributions are compared against experimental data and in general agree well with the exception of select outliers. - Highlights: • New experimental data is presented on oxide layer thickness of irradiated aluminum fuel. • Five oxide growth correlations and four convective heat transfer correlations are used to compute the oxide layer thickness. • The oxide layer thickness distribution is predicted via correlation for each respective experiment. • The measured experiment and predicted distributions correlate well, with few outliers.

Oxidative Stability of Dispersions Prepared from Purified Marine Phospholipid and the Role of α-Tocopherol

DEFF Research Database (Denmark)

Lu, Henna Fung Sieng; Nielsen, Nina Skall; Baron, Caroline P.

2012-01-01

, respectively, during 32 days of storage at 2 °C. Nonenzymatic browning was investigated through measurement of Strecker aldehydes, color changes, and pyrrole content. Dispersions containing α-tocopherol or higher levels of purified marine PL showed a lower increment of volatiles after 32 days storage......The objective of this study was to investigate the oxidative stability of dispersions prepared from different levels of purified marine phospholipid (PL) obtained by acetone precipitation, with particular focus on the interaction between α-tocopherol and PL in dispersions. This also included...... the investigation of nonenzymatic browning in purified marine PL dispersions. Dispersions were prepared by high-pressure homogenizer. The oxidative and hydrolytic stabilities of dispersions were investigated by determination of hydroperoxides, secondary volatile oxidation products, and free fatty acids...

Cobalt oxide-based catalysts deposited by cold plasma for proton exchange membrane fuel cells

Energy Technology Data Exchange (ETDEWEB)

Kazimierski, P.; Jozwiak, L.; Sielski, J.; Tyczkowski, J., E-mail: jacek.tyczkowski@p.lodz.pl

2015-11-02

In proton exchange membrane fuel cells (PEMFC), both the anodic hydrogen oxidation reaction and the cathodic oxygen reduction reaction (ORR) require appropriate catalysts. So far, platinum-based catalysts are still the best option for this purpose. However, because these catalysts are too expensive for making commercially viable fuel cells, extensive research over the past decade has focused on developing noble metal-free alternative catalysts. In this paper, an approach based on cobalt oxide films fabricated by plasma-enhanced metal-organic chemical vapor deposition is presented. Such a material can be used to prepare catalysts for ORR in PEMFC. The films containing CoO{sub X} were deposited on a carbon paper thereby forming the electrode. Morphology and atomic composition of the films were investigated by scanning electron microscopy and energy-dispersive X-ray spectroscopy, respectively. The possibility of their application as the electro-catalyst for ORR in PEMFC was investigated and the electro-catalytic activities were evaluated by the electrochemical measurements and single cell tests. It was found that the fuel cell with Pt as the anode catalyst and CoO{sub X} deposit as the cathode catalyst was characterized by the open circuit voltage of 635 mV, Tafel slope of approx. 130 mV/dec and the maximum power density of 5.3 W/m{sup 2}. - Highlights: • Cobalt oxide catalyst for proton exchange membrane fuel cells was plasma deposited. • The catalyst exhibits activity for the oxygen reduction reaction. • Morphology and atomic composition of the catalyst were determined.

Development of oxide dispersion strengthened 9Cr ferritic-martensitic steel clad tube for fast reactor

International Nuclear Information System (INIS)

Laha, K.; Saroja, S.; Mathew, M.D.; Jayakumar, T.; Vijay, R.; Venugopal Reddy, A.; Lakshminarayana, B.; Kapoor, Komal; Jha, S.K.; Tonpe, S.S.

2012-01-01

One of the key issues in the economical operation of FBR is to achieve high burn-up of fuel (200-250 GWd/t) which considerably reduces the fuel cycle cost. This imposes stringent requirements of void swelling resistance upto 200 dpa for the core structural materials. Presently used alloy 09 (a modified austenitic stainless steel, 15Cr-15Ni-Ti) for PFBR has void swelling limit less than 150 dpa. Because of the inherent void swelling resistance, 9-12Cr steels ferritic/martensitic steels are qualified for irradiation upto 200 dpa but their low creep strength at temperatures above 600 deg C restricts their application as a clad material. Oxide dispersion strengthening is found to be promising means of extending the creep resistance of ferritic/martensitic steels beyond 650 deg C without sacrificing the inherent advantages of high thermal conductivity and low swelling of ferritic steels

DART model for thermal conductivity of U3Si2 aluminum dispersion fuel

International Nuclear Information System (INIS)

Rest, J.; Snelgrove, J.L.; Hofman, G.L.

1995-09-01

This paper describes the primary physical models that form the basis of the DART model for calculating irradiation-induced changes in the thermal conductivity of aluminium dispersion fuel. DART calculations of fuel swelling, pore closure, and thermal conductivity are compared with measured values

In-reactor behaviour of centrifugally atomized U3Si dispersion fuel irradiated at high temperature in HANARO

International Nuclear Information System (INIS)

Kim, Ki Hwan; Park, Jong Man; Yoo, Byeong Ok; Park, Dae Kyu; Lee, Choong Sung; Kim, Chang Kyu

2002-01-01

The irradiation test on full-size U 3 Si dispersion fuel elements, prepared by centrifugal atomization and conventional comminution method, has been performed up to about 77 at.% U-235 in maximum burn-up at CT hole position having the highest power condition in the HANARO reactor, in order to examine the irradiation performance of the atomized U 3 Si for the driver fuels of HANARO. The in-reactor interaction of the atomized U 3 Si dispersion fuel meats is generally assumed to be acceptable with the range of 5-15 μm in average thickness. The atomized spherical particles have more uniform and thinner reaction layer than the comminuted irregular particles. The U 3 Si particles have relatively fine and uniform size distribution of fission gas bubbles, irrespective of the powdering method. The bubble population in the atomized particles appears to be finer and more homogeneous with the characteristics of narrower bubble size distribution than that of the comminuted fuel. The atomized U 3 Si dispersion fuel elements exhibit sound swelling behaviours of 5 % in ΔV/V m even at ∼77 at.% U-235 burn-up, which meets with the safety criterion of the fuel rod, 20vol.% for HANARO. The atomized U3Si dispersion fuel elements show smaller swelling than the comminuted fuel elements

A novel direct carbon fuel cell by approach of tubular solid oxide fuel cells

Energy Technology Data Exchange (ETDEWEB)

Liu, Renzhu; Zhao, Chunhua; Li, Junliang; Zeng, Fanrong; Wang, Shaorong; Wen, Tinglian; Wen, Zhaoyin [CAS Key Laboratory of Materials for Energy Conversion, Shanghai Inorganic Energy Materials and Power Source Engineering Center, Shanghai Institute of Ceramics, Chinese Academy of Sciences (SICCAS), 1295 Dingxi Road, Shanghai 200050 (China)

2010-01-15

A direct carbon fuel cell based on a conventional anode-supported tubular solid oxide fuel cell, which consisted of a NiO-YSZ anode support tube, a NiO-ScSZ anode functional layer, a ScSZ electrolyte film, and a LSM-ScSZ cathode, has been successfully achieved. It used the carbon black as fuel and oxygen as the oxidant, and a preliminary examination of the DCFC has been carried out. The cell generated an acceptable performance with the maximum power densities of 104, 75, and 47 mW cm{sup -2} at 850, 800, and 750 C, respectively. These results demonstrate the feasibility for carbon directly converting to electricity in tubular solid oxide fuel cells. (author)

Electrocontact material based on silver dispersion-strengthened by nickel, titanium, and zinc oxides

Science.gov (United States)

Zeer, G. M.; Zelenkova, E. G.; Belousov, O. V.; Beletskii, V. V.; Nikolaev, S. V.; Ledyaeva, O. N.

2017-09-01

Samples of a composite electrocontact material based on silver strengthened by the dispersed phases of zinc and titanium oxides have been investigated by the electron microscopy and energy dispersive X-ray spectroscopy. A uniform distribution of the oxide phases containing 2 wt % zinc oxide in the initial charge has been revealed. The increase in the amount of zinc oxide leads to an increase of the size of the oxide phases. It has been shown that at the zinc oxide content of 2 wt %, the minimum wear is observed in the process of electroerosion tests; at 3 wt %, an overheating and welding of the contacts are observed.

Uranium plutonium oxide fuels

International Nuclear Information System (INIS)

Cox, C.M.; Leggett, R.D.; Weber, E.T.

1981-01-01

Uranium plutonium oxide is the principal fuel material for liquid metal fast breeder reactors (LMFBR's) throughout the world. Development of this material has been a reasonably straightforward evolution from the UO 2 used routinely in the light water reactor (LWR's); but, because of the lower neutron capture cross sections and much lower coolant pressures in the sodium cooled LMFBR's, the fuel is operated to much higher discharge exposures than that of a LWR. A typical LMFBR fuel assembly is shown. Depending on the required power output and the configuration of the reactor, some 70 to 400 such fuel assemblies are clustered to form the core. There is a wide variation in cross section and length of the assemblies where the increasing size reflects a chronological increase in plant size and power output as well as considerations of decreasing the net fuel cycle cost. Design and performance characteristics are described

The Effect of Uncertainties on the Operating Temperature of U-Mo/Al Dispersion Fuel

Energy Technology Data Exchange (ETDEWEB)

Sweidana, Faris B.; Mistarihia, Qusai M.; Ryu Ho Jin [KAIST, Daejeon (Korea, Republic of); Yim, Jeong Sik [KAERI, Daejeon (Korea, Republic of)

2016-05-15

In this study, uncertainty and combined uncertainty studies have been carried out to evaluate the uncertainty of the parameters affecting the operational temperature of U-Mo/Al fuel. The uncertainties related to the thermal conductivity of fuel meat, which consists of the effects of thermal diffusivity, density and specific heat capacity, the interaction layer (IL) that forms between the dispersed fuel and the matrix, fuel plate dimensions, heat flux, heat transfer coefficient and the outer cladding temperature were considered. As the development of low-enriched uranium (LEU) fuels has been pursued for research reactors to replace the use of highly-enriched uranium (HEU) for the improvement of proliferation resistance of fuels and fuel cycle, U-Mo particles dispersed in an Al matrix (UMo/Al) is a promising fuel for conversion of the research reactors that currently use HEU fuels to LEUfueled reactors due to its high density and good irradiation stability. Several models have been developed for the estimation of the thermal conductivity of U–Mo fuel, mainly based on the best fit of the very few measured data without providing uncertainty ranges. The purpose of this study is to provide a reasonable estimation of the upper bounds and lower bounds of fuel temperatures with burnup through the evaluation of the uncertainties in the thermal conductivity of irradiated U-Mo/Al dispersion fuel. The combined uncertainty study using RSS method evaluated the effect of applying all the uncertainty values of all the parameters on the operational temperature of U-Mo/Al fuel. The overall influence on the value of the operational temperature is 16.58 .deg. C at the beginning of life and it increases as the burnup increases to reach 18.74 .deg. C at a fuel meat fission density of 3.50E+21 fission/cm{sup 3}. Further studies are needed to evaluate the behavior more accurately by including other parameters uncertainties such as the interaction layer thermal conductivity.

Desulfurization of Diesel Fuel by Oxidation and Solvent Extraction

Directory of Open Access Journals (Sweden)

Wadood Taher Mohammed

2015-02-01

Full Text Available This research presents a study in ultra-desulfurization of diesel fuel produced from conventional hydro desulfurization process, using oxidation and solvent extraction techniques. Dibenzothiophene (DBT was the organosulfur compound that had been detected in sulfur removal. The oxidation process used hydrogen peroxide as an oxidant and acetic acid as homogeneous catalyst . The solvent extraction process used acetonitrile (ACN and N-methyl – 2 - pyrrolidone (NMP as extractants . Also the effect of five parameters (stirring speed :150 , 250 , 350 , and 450 rpm, temperature (30 , 40 , 45 , and 50 oC, oxidant/simulated diesel fuel ratio (0.5 , 0.75 , 1 , and 1.5 , catalyst/oxidant ratio(0.125,0.25,0.5,and0.75 , and solvent/simulated diesel fuel ratio(0.5,0.6,0.75,and1 were examined as well as solvent type. The results exhibit that the highest removal of sulfur is 98.5% using NMP solvent while it is 95.8% for ACN solvent. The set of conditions that show the highest sulfur removal is: stirring speed of 350 rpm , temperature 50oC , oxidant/simulated diesel fuel ratio 1 , catalyst/oxidant ratio 0.5 , solvent/simulated diesel fuel ratio 1. These best conditions were applied upon real diesel fuel (produced from Al-Dora refinerywith 1000 ppm sulfur content . It was found that sulfur removal was 64.4% using ACN solvent and 75% using NMP solvent.

Effect of a surface oxide-dispersion-strengthened layer on mechanical strength of zircaloy-4 tubes

Directory of Open Access Journals (Sweden)

Yang-Il Jung

2018-03-01

Full Text Available An oxide-dispersion-strengthened (ODS layer was formed on Zircaloy-4 tubes by a laser beam scanning process to increase mechanical strength. Laser beam was used to scan the yttrium oxide (Y2O3–coated Zircaloy-4 tube to induce the penetration of Y2O3 particles into Zircaloy-4. Laser surface treatment resulted in the formation of an ODS layer as well as microstructural phase transformation at the surface of the tube. The mechanical strength of Zircaloy-4 increased with the formation of the ODS layer. The ring-tensile strength of Zircaloy-4 increased from 790 to 870 MPa at room temperature, from 500 to 575 MPa at 380°C, and from 385 to 470 MPa at 500°C. Strengthening became more effective as the test temperature increased. It was noted that brittle fracture occurred at room temperature, which was not observed at elevated temperatures. Resistance to dynamic high-temperature bursting improved. The burst temperature increased from 760 to 830°C at a heating rate of 5°C/s and internal pressure of 8.3 MPa. The burst opening was also smaller than those in fresh Zircaloy-4 tubes. This method is expected to enhance the safety of Zr fuel cladding tubes owing to the improvement of their mechanical properties. Keywords: Laser Surface Treatment, Microstructure, Oxide Dispersion Strengthened Alloy, Tensile Strength, Zirconium Alloy

Postirradiation analysis of experimental uranium-silicide dispersion fuel plates

International Nuclear Information System (INIS)

Hofman, G.L.; Neimark, L.A.

1985-01-01

Low-enriched uranium silicide dispersion fuel plates were irradiated to maximum burnups of 96% of 235 U. Fuel plates containing 33 v/o U 3 Si and U 3 Si 2 behaved very well up to this burnup. Plates containing 33 v/o U 3 Si-Al pillowed between 90 and 96% burnup of the fissile atoms. More highly loaded U 3 Si-Al plates, up to 50 v/o were found to pillow at lower burnups. Plates containing 40 v/o U 3 Si showed an increase swelling rate around 85% burnup. 5 refs., 10 figs

Monolithic Approach to Oxide Dispersion Strengthened Aluminum, Phase I

Data.gov (United States)

National Aeronautics and Space Administration — Nassau Stern Company is investigating an approach for manufacturing oxide dispersion strengthened (ODS) aluminum in bulk rather than powder form. The approach...

DART model for thermal conductivity of U3Si2 Aluminum dispersion fuel

International Nuclear Information System (INIS)

Rest, J.; Snelgrove, J.L.; Hofman, G.L.

2004-01-01

This paper describes the primary physical models that form the basis of the DART model for calculating irradiation-induced changes in the thermal conductivity of aluminum dispersion fuel. DART calculations of fuel swelling, pore closure, and thermal conductivity are compared with measured values. (author)

High resolution Transmission Electron Microscopy characterization of a milled oxide dispersion strengthened steel powder

Energy Technology Data Exchange (ETDEWEB)

Loyer-Prost, M., E-mail: marie.loyer-prost@cea.fr [DEN-Service de Recherches de Métallurgie Physique, CEA, Université Paris-Saclay, F-91191, Gif-sur-Yvette (France); Merot, J.-S. [Laboratoire d’Etudes des Microstructures – UMR 104, CNRS/ONERA, BP72-29, Avenue de la Division Leclerc, 92 322, Châtillon (France); Ribis, J. [DEN-Service de Recherches de Métallurgie Appliquée, CEA, Université Paris-Saclay, F-91191, Gif-sur-Yvette (France); Le Bouar, Y. [Laboratoire d’Etudes des Microstructures – UMR 104, CNRS/ONERA, BP72-29, Avenue de la Division Leclerc, 92 322, Châtillon (France); Chaffron, L. [DEN-Service de Recherches de Métallurgie Appliquée, CEA, Université Paris-Saclay, F-91191, Gif-sur-Yvette (France); Legendre, F. [DEN-Service de la Corrosion et du Comportement des Matériaux dans leur Environnement, CEA, Université Paris-Saclay, F-91191, Gif-sur-Yvette (France)

2016-10-15

Oxide Dispersion Strengthened (ODS) steels are promising materials for generation IV fuel claddings as their dense nano-oxide dispersion provides good creep and irradiation resistance. Even if they have been studied for years, the formation mechanism of these nano-oxides is still unclear. Here we report for the first time a High Resolution Transmission Electron Microscopy and Energy Filtered Transmission Electron Microscopy characterization of an ODS milled powder. It provides clear evidence of the presence of small crystalline nanoclusters (NCs) enriched in titanium directly after milling. Small NCs (<5 nm) have a crystalline structure and seem partly coherent with the matrix. They have an interplanar spacing close to the (011) {sub bcc} iron structure. They coexist with larger crystalline spherical precipitates of 15–20 nm in size. Their crystalline structure may be metastable as they are not consistent with any Y-Ti-O or Ti-O structure. Such detailed observations in the as-milled grain powder confirm a mechanism of Y, Ti, O dissolution in the ferritic matrix followed by a NC precipitation during the mechanical alloying process of ODS materials. - Highlights: • We observed an ODS ball-milled powder by high resolution transmission microscopy. • The ODS ball-milled powder exhibits a lamellar microstructure. • Small crystalline nanoclusters were detected in the milled ODS powder. • The nanoclusters in the ODS milled powder are enriched in titanium. • Larger NCs of 15–20 nm in size are, at least, partly coherent with the matrix.

Assessment of bio-fuel options for solid oxide fuel cell applications

Science.gov (United States)

Lin, Jiefeng

Rising concerns of inadequate petroleum supply, volatile crude oil price, and adverse environmental impacts from using fossil fuels have spurred the United States to promote bio-fuel domestic production and develop advanced energy systems such as fuel cells. The present dissertation analyzed the bio-fuel applications in a solid oxide fuel cell-based auxiliary power unit from environmental, economic, and technological perspectives. Life cycle assessment integrated with thermodynamics was applied to evaluate the environmental impacts (e.g., greenhouse gas emission, fossil energy consumption) of producing bio-fuels from waste biomass. Landfill gas from municipal solid wastes and biodiesel from waste cooking oil are both suggested as the promising bio-fuel options. A nonlinear optimization model was developed with a multi-objective optimization technique to analyze the economic aspect of biodiesel-ethanol-diesel ternary blends used in transportation sectors and capture the dynamic variables affecting bio-fuel productions and applications (e.g., market disturbances, bio-fuel tax credit, policy changes, fuel specification, and technological innovation). A single-tube catalytic reformer with rhodium/ceria-zirconia catalyst was used for autothermal reformation of various heavy hydrocarbon fuels (e.g., diesel, biodiesel, biodiesel-diesel, and biodiesel-ethanol-diesel) to produce a hydrogen-rich stream reformates suitable for use in solid oxide fuel cell systems. A customized mixing chamber was designed and integrated with the reformer to overcome the technical challenges of heavy hydrocarbon reformation. A thermodynamic analysis, based on total Gibbs free energy minimization, was implemented to optimize the operating environment for the reformations of various fuels. This was complimented by experimental investigations of fuel autothermal reformation. 25% biodiesel blended with 10% ethanol and 65% diesel was determined to be viable fuel for use on a truck travelling with

Thermodynamic and kinetic modelling of fuel oxidation behaviour in operating defective fuel

International Nuclear Information System (INIS)

Lewis, B.J.; Thompson, W.T.; Akbari, F.; Thompson, D.M.; Thurgood, C.; Higgs, J.

2004-01-01

A theoretical treatment has been developed to predict the fuel oxidation behaviour in operating defective nuclear fuel elements. The equilibrium stoichiometry deviation in the hyper-stoichiometric fuel has been derived from thermodynamic considerations using a self-consistent set of thermodynamic properties for the U-O system, which emphasizes replication of solubilities and three-phase invariant conditions displayed in the U-O binary phase diagram. The kinetics model accounts for multi-phase transport including interstitial oxygen diffusion in the solid and gas-phase transport of hydrogen and steam in the fuel cracks. The fuel oxidation model is further coupled to a heat conduction model to account for the feedback effect of a reduced thermal conductivity in the hyper-stoichiometric fuel. A numerical solution has been developed using a finite-element technique with the FEMLAB software package. The model has been compared to available data from several in-reactor X-2 loop experiments with defective fuel conducted at the Chalk River Laboratories. The model has also been benchmarked against an O/U profile measurement for a spent defective fuel element discharged from a commercial reactor

Oxide nanoparticles in an Al-alloyed oxide dispersion strengthened steel: crystallographic structure and interface with ferrite matrix

DEFF Research Database (Denmark)

Zhang, Zhenbo; Pantleon, Wolfgang

2017-01-01

Oxide nanoparticles are quintessential for ensuring the extraordinary properties of oxide dispersion strengthened (ODS) steels. In this study, the crystallographic structure of oxide nanoparticles, and their interface with the ferritic steel matrix in an Al-alloyed ODS steel, i.e. PM2000, were...

The TMI regenerable solid oxide fuel cell

Science.gov (United States)

Cable, Thomas L.

1995-04-01

Energy storage and production in space requires rugged, reliable hardware which minimizes weight, volume, and maintenance while maximizing power output and usable energy storage. These systems generally consist of photovoltaic solar arrays which operate during sunlight cycles to provide system power and regenerate fuel (hydrogen) via water electrolysis; during dark cycles, hydrogen is converted by the fuel cell into system. The currently preferred configuration uses two separate systems (fuel cell and electrolyzer) in conjunction with photovoltaic cells. Fuel cell/electrolyzer system simplicity, reliability, and power-to-weight and power-to-volume ratios could be greatly improved if both power production (fuel cell) and power storage (electrolysis) functions can be integrated into a single unit. The Technology Management, Inc. (TMI), solid oxide fuel cell-based system offers the opportunity to both integrate fuel cell and electrolyzer functions into one unit and potentially simplify system requirements. Based an the TMI solid oxide fuel cell (SOPC) technology, the TMI integrated fuel cell/electrolyzer utilizes innovative gas storage and operational concepts and operates like a rechargeable 'hydrogen-oxygen battery'. Preliminary research has been completed on improved H2/H2O electrode (SOFC anode/electrolyzer cathode) materials for solid oxide, regenerative fuel cells. Improved H2/H2O electrode materials showed improved cell performance in both fuel cell and electrolysis modes in reversible cell tests. ln reversible fuel cell/electrolyzer mode, regenerative fuel cell efficiencies (ratio of power out (fuel cell mode) to power in (electrolyzer model)) improved from 50 percent (using conventional electrode materials) to over 80 percent. The new materials will allow the TMI SOFC system to operate as both the electrolyzer and fuel cell in a single unit. Preliminary system designs have also been developed which indicate the technical feasibility of using the TMI SOFC

Observation on the irradiation behavior of U-Mo alloy dispersion fuel

International Nuclear Information System (INIS)

Hofman, Gerard L.; Meyer, Mitchell K.; Park, Jong-Man

2000-01-01

Initial results from the postirradiation examination of high-density dispersion fuel test RERTR-3 are discussed. The U-Mo alloy fuels in this test were irradiated to 40% U-235 burnup at temperature ranging from 140 0 C to 240 0 C. Temperature has a significant effect on overall swelling of the test plates. The magnitude of the swelling appears acceptable and no unstable irradiation behavior is evident. (author)

Development of PRIME for irradiation performance analysis of U-Mo/Al dispersion fuel

Energy Technology Data Exchange (ETDEWEB)

Jeong, Gwan Yoon; Kim, Yeon Soo; Jeong, Yong Jin; Park, Jong Man; Sohn, Dong-Seong

2018-04-01

A prediction code for the thermo-mechanical performance of research reactor fuel (PRIME) has been developed with the implementation of developed models to analyze the irradiation behavior of U-Mo dispersion fuel. The code is capable of predicting the two-dimensional thermal and mechanical performance of U-Mo dispersion fuel during irradiation. A finite element method was employed to solve the governing equations for thermal and mechanical equilibria. Temperature-and burnup-dependent material properties of the fuel meat constituents and cladding were used. The numerical solution schemes in PRIME were verified by benchmarking solutions obtained using a commercial finite element analysis program (ABAQUS).The code was validated using irradiation data from RERTR, HAMP-1, and E-FUTURE tests. The measured irradiation data used in the validation were IL thickness, volume fractions of fuel meat constituents for the thermal analysis, and profiles of the plate thickness changes and fuel meat swelling for the mechanical analysis. The prediction results were in good agreement with the measurement data for both thermal and mechanical analyses, confirming the validity of the code. (c) 2018 Elsevier B.V. All rights reserved.

Summary of the physical chemical analyses of mixed oxide nuclear fuel as they might influence biological behavior and internal dose

International Nuclear Information System (INIS)

Eidson, A.F.; Mewhinney, J.A.

1987-01-01

Twelve representative materials that might be accidentally released during the fabrication of mixed-oxide nuclear fuel pellets were studied using x-ray diffraction, infrared spectroscopy, energy dispersive x-ray fluorescence, alpha spectroscopy and in vitro dissolution methods. The results are related to a postulated exposure accident and to inhalation experiments using laboratory animals. 19 refs., 5 figs., 19 tabs

Comparison of irradiation behavior of different uranium silicide dispersion fuel element designs

International Nuclear Information System (INIS)

Hofman, G.L.; Rest, J.; Snelgrove, J.L.

1995-01-01

Calculations of fuel swelling of U 3 SiAl-Al and U 3 Si 2 were performed for various dispersion fuel element designs. Breakaway swelling criteria in the form of critical fuel volume fractions were derived with data obtained from U 3 SiAl-Al plate irradiations. The results of the analysis show that rod-type elements remain well below the pillowing threshold. However, tubular fuel elements, which behave essentially like plates, will likely develop pillows or blisters at around 90% 235 U burnup. The U 3 Si 2 -Al compounds demonstrate stable swelling behavior throughout the entire burnup range for all fuel element designs

Performance of artificially defected LWR fuel rods in an unlimited air dry storage atmosphere

International Nuclear Information System (INIS)

Einziger, R.E.; Knecht, R.L.; Cantley, D.A.; Cook, J.A.

1983-09-01

Thus far the tests are inconclusive as to whether breached LWR fuel can be stored at 230 0 C for long periods of time in air without fuel oxidation and dispersion. There is every indication, as expected, that there is no oxidation problem in an inert atmosphere. Only one of four defects exposed to unlimited air gave any indication of fuel oxidation. It has been suggested that this might be an incubation effect and continued operation would result in oxidation occurring at all four defects. As yet the destructive examination of the BWR rod has not been completed, so it is not possible to determine if cladding splitting was due to an anomoly in this test rod or something that can be expected in LWR rods in general. Thus far there is no indication of respirable particle dispersal even if fuel oxidation does occur

Reaction layer growth and reaction heat of U-Mo/Al dispersion fuels using centrifugally atomized powders

International Nuclear Information System (INIS)

Ryu, Ho Jin; Han, Young Soo; Park, Jong Man; Park, Soon Dal; Kim, Chang Kyu

2003-01-01

The growth behavior of reaction layers and heat generation during the reaction between U-Mo powders and the Al matrix in U-Mo/Al dispersion fuels were investigated. Annealing of 10 vol.% U-10Mo/Al dispersion fuels at temperatures from 500 to 550 deg. C was carried out for 10 min to 36 h to measure the growth rate and the activation energy for the growth of reaction layers. The concentration profiles of reaction layers between the U-10Mo vs. Al diffusion couples were measured and the integrated interdiffusion coefficients were calculated for the U and Al in the reaction layers. Heat generation of U-Mo/Al dispersion fuels with 10-50 vol.% of U-Mo fuel during the thermal cycle from room temperature to 700 deg. C was measured employing the differential scanning calorimetry. Exothermic heat from the reaction between U-Mo and the Al matrix is the largest when the volume fraction of U-Mo fuel is about 30 vol.%. The unreacted fraction in the U-Mo powders increases as the volume fraction of U-Mo fuel increases from 30 to 50 vol.%

UV stabilization of wood by nano metal oxides dispersed in propylene glycol.

Science.gov (United States)

Nair, Sreeja; Nagarajappa, Giridhar B; Pandey, Krishna K

2018-06-01

Nanoparticles of some of the metal oxides are known to have high UV protective efficiency. The UV filtering efficiency of nanoparticles invariably depends on their size and stability in the dispersion. In the present work, a stable dispersion of nanoparticles of three metal oxides, zinc oxide (ZnO), cerium oxide (CeO 2 ) and titanium dioxide (TiO 2 ), was prepared in propylene glycol (PG) using ultrasonication. The method is easy and useful as no additional surfactant or dispersant is needed. The particle size and its distribution was confirmed by Scanning Electron Microscopy and Dynamic Light Scattering. The stability of dispersion was assessed by UV-visible absorption spectroscopy. The UV stability of wood surfaces of Wrightia tinctoria coated with nanodispersions of ZnO, CeO 2 and TiO 2 was evaluated under laboratory conditions in an accelerated weathering tester. Changes in the colour and FTIR spectra of exposed specimens were measured periodically. Rapid colour darkening (yellowing) was observed in uncoated and PG coated specimens. In contrast, nanodispersion coated specimens prevented photo-yellowing considerably with significant reduction in colour changes examined by CIE L*, a*, b* and ΔE*. Increase in concentration of nanoparticles in the dispersion imparted higher resistance to UV induced degradation. However, increased concentration of nanoparticles reduced the transparency of the coating. FTIR analysis indicated rapid degradation of lignin in uncoated and PG coated specimens due to UV exposure. Coating of wood surfaces with nanodispersions restricted lignin degradation. The study also demonstrates the potential of propylene glycol as a dispersant for developing stable and efficient UV protective nanodispersions for wood coating. Copyright © 2018 Elsevier B.V. All rights reserved.

Micro-Solid Oxide Fuel Cell: A multi-fuel approach for portable applications

International Nuclear Information System (INIS)

Patil, Tarkeshwar C.; Duttagupta, Siddhartha P.

2016-01-01

Highlights: • We report the oxygen ion transport properties at the electrode–electrolyte interface (EEI) of the SOFC for the first time. • This ion transport plays a key role in the overall performance of SOFCs with different fuels. • The GIIB mechanism is also studied for the first time. • GIIB is assumed to be the prime reason for low power density and ion conductivity at the EEI when using hydrocarbon fuels. • Due to its scalability, a fuel cell can serve as a power source for on-chip applications and all portable equipment. - Abstract: The impact of oxygen ion transport at the electrolyte–electrode interface of a micro-solid oxide fuel cell using different fuels is investigated. Model validation is performed to verify the results versus the reported values. Furthermore, as the hydrogen-to-carbon ratio decreases, the diffusivity of the oxygen ion increases. This increase in diffusivity is observed because the number of hydrogen atoms available as the reacting species increases in fuels with lower hydrogen-to-carbon ratios. The oxygen ion conductivity and output power density decrease as the hydrogen-to-carbon ratio of the fuels decreases. The reason behind this impact is the formation of a gas-induced ion barrier at the electrode–electrolyte interface by the CO_2 molecules formed during the reaction at the interface, thus blocking the flow of oxygen ions. As the oxygen ions become blocked, the output current contribution from the reaction also decreases and thereby affects the overall performance of the micro-solid oxide fuel cell. The experimental verification confirms this because of a significant decrease in the output power density. Furthermore, as per the application in portable devices, the appropriate choice of fuel can be chosen so that the micro-solid oxide fuel cell operates at the maximum power density.

Method of manufacturing gadolinium oxide-incorporated nuclear fuel sintering products

International Nuclear Information System (INIS)

Komono, Akira; Seki, Makoto; Omori, Sadayuki.

1987-01-01

Purpose: To manufacture nuclear fuel sintering products excellent in burning property and mechanical property. Constitution: In the manufacturing step for nuclear fuel sintering products, specific metal oxides are added for promoting the growth of crystal grains in the sintering. Those metal oxides melted at a temperature lower than the sintering temperature of a mixture of nuclear fuel oxide powder and oxide power, or those metal oxides causing eutectic reaction are used as the metal oxide. Particularly, those compounds having oxygen atom - metal atom ratio (O/M) of not less than 2 are preferably used. As such metal oxides usable herein transition metal oxides, e.g., Nb 2 O 5 , TiO 2 , MoO 3 and WO 3 are preferred, with Nb 2 O 3 and TiO 2 being preferred particularly. (Seki, T.)

The improvement of technology for high-uranium-density Al-base dispersion fuel plates

International Nuclear Information System (INIS)

Shouhui, Dai; Rongxian, Sun; Hejian, Mao; Baosheng, Zhao; Changgen, Yin

1987-01-01

An improved rolling process was developed for manufacturing Al-base dispersion fuel plates. When the fuel content in the meat increased up to 50 vol%, the non-uniformity of uranium is not more than ± 7.2%, and the minimum cladding thickness is not less than 0.32 mm. (Author)

Unirradiated characteristics of U-Si alloys as dispersed-phase fuels

International Nuclear Information System (INIS)

Domagala, R.F.; Wiencek, T.C.

1987-06-01

To satisfy the power demands of many research reactors, a new LEU fuel with a high density and U content was needed. Any fuel must be compatible with Al and its alloys so that it may be fabricable as a dispersed-phase in Al alloy and Al matrix plate-type elements following, as nearly as possible, established commercial manufacturing techniques. U-Si and U-Si-Al alloys at or near the composition of U 3 Si were immediately attractive because of work documented by the Canadians. 8 refs., 2 figs

Development of a FBR fuel pin bundle deformation analysis code 'BAMBOO' . Development of a dispersion model and its validation

International Nuclear Information System (INIS)

Uwaba, Tomoyuki; Ukai, Shigeharu; Asaga, Takeo

2002-03-01

Bundle Duct Interaction (BDI) is one of the life limiting factors of a FBR fuel subassembly. Under the BDI condition, the fuel pin dispersion would occur mainly by the deviation of the wire position due to the irradiation. In this study the effect of the dispersion on the bundle deformation was evaluated by using the BAMBOO code and following results were obtained. (1) A new contact analysis model was introduced in BAMBOO code. This model considers the contact condition at the axial position other than the nodal point of the beam element that composes the fuel pin. This improvement made it possible in the bundle deformation analysis to cause fuel pin dispersion due to the deviations of the wire position. (2) This model was validated with the results of the out-of-pile compression test with the wire deviation. The calculated pin-to-duct and pin-to-pin clearances with the dispersion model almost agreed with the test results. Therefore it was confirmed that the BAMBOO code reasonably predicts the bundle deformation with the dispersion. (3) In the dispersion bundle the pin-to-pin clearances widely scattered. And the minimum pin-to-duct clearance increased or decreased depending on the dispersion condition compared to the no-dispersion bundle. This result suggests the possibility that the considerable dispersion would affect the thermal integrity of the bundle. (author)

Solid Oxide Fuel Cell

DEFF Research Database (Denmark)

2010-01-01

The solid oxide fuel cell comprising a metallic support material, an active anode layer consisting of a good hydrocarbon cracking catalyst, an electrolyte layer, an active cathode layer, and a transition layer consisting of preferably a mixture of LSM and a ferrite to the cathode current collector...

Fuel elements based on U3O8 dispersions in aluminium

International Nuclear Information System (INIS)

Haydt, H.M.

1990-01-01

A review of the Nuclear Metallurgy Division, now Nuclear and Energetic Research Institute, to the national Know-how development of fuel elements fabrication from U 3 O 8 dispersions in aluminium, during 1962 to 1977, are described. (C.G.C.)

Irradiation behavior of U{sub 6}Mn-Al dispersion fuel elements

Energy Technology Data Exchange (ETDEWEB)

Meyer, M.K. E-mail: mitchell.meyer@anl.gov; Wiencek, T.C.; Hayes, S.L.; Hofman, G.L

2000-04-01

Irradiation testing of U{sub 6}Mn-Al dispersion fuel miniplates was conducted in the Oak Ridge Research Reactor (ORR). Post-irradiation examination showed that U{sub 6}Mn in an unrestrained plate configuration performs similarly to U{sub 6}Fe under irradiation, forming extensive and interlinked fission gas bubbles at a fission density of approximately 3x10{sup 27} m{sup -3}. Fuel plate failure occurs by fission gas pressure driven 'pillowing' on continued irradiation.

Cuprous oxide nanoparticles dispersed on reduced graphene oxide as an efficient electrocatalyst for oxygen reduction reaction.

Science.gov (United States)

Yan, Xiao-Yan; Tong, Xi-Li; Zhang, Yue-Fei; Han, Xiao-Dong; Wang, Ying-Yong; Jin, Guo-Qiang; Qin, Yong; Guo, Xiang-Yun

2012-02-11

Cuprous oxide (Cu(2)O) nanoparticles dispersed on reduced graphene oxide (RGO) were prepared by reducing copper acetate supported on graphite oxide using diethylene glycol as both solvent and reducing agent. The Cu(2)O/RGO composite exhibits excellent catalytic activity and remarkable tolerance to methanol and CO in the oxygen reduction reaction. This journal is © The Royal Society of Chemistry 2012

Redox Stable Anodes for Solid Oxide Fuel Cells

Directory of Open Access Journals (Sweden)

Guoliang eXiao

2014-06-01

Full Text Available Solid oxide fuel cells (SOFCs can convert chemical energy from the fuel directly to electrical energy with high efficiency and fuel flexibility. Ni-based cermets have been the most widely adopted anode for SOFCs. However, the conventional Ni-based anode has low tolerance to sulfur-contamination, is vulnerable to deactivation by carbon build-up (coking from direct oxidation of hydrocarbon fuels, and suffers volume instability upon redox cycling. Among these limitations, the redox instability of the anode is particularly important and has been intensively studied since the SOFC anode may experience redox cycling during fuel cell operations even with the ideal pure hydrogen as the fuel. This review aims to highlight recent progresses on improving redox stability of the conventional Ni-based anode through microstructure optimization and exploration of alternative ceramic-based anode materials.

When do oxide precipitates form during consolidation of oxide dispersion strengthened steels?

Energy Technology Data Exchange (ETDEWEB)

Deschamps, A., E-mail: alexis.deschamps@grenoble-inp.fr [Univ. Grenoble Alpes, SIMAP, F-38000 Grenoble (France); CNRS, SIMAP, F-38000 Grenoble (France); De Geuser, F. [Univ. Grenoble Alpes, SIMAP, F-38000 Grenoble (France); CNRS, SIMAP, F-38000 Grenoble (France); Malaplate, J.; Sornin, D. [DEN, DANS, DMN, Service de Recherches Métallurgiques Appliquées, CEA, Université Paris-Saclay, 91191 Gif-Sur-Yvette (France)

2016-12-15

The processing of oxide dispersion strengthened (ODS) steels involves ball milling, where the oxide forming species are driven in solid solution. Precipitation of the nanometre-scale oxides occurs during subsequent annealing and consolidation. This paper reports in-situ Small-Angle X-ray Scattering measurements of the formation of these precipitates during heating of cold-compressed as-milled powders. Clusters are already initially present, and precipitation starts at 300 °C. The maximum precipitate density is achieved at 600 °C, followed by very slow coarsening at higher temperature. These results open the way to understand the coupled evolution of precipitation and crystalline defects during heating and consolidation of ODS steels.

Structures of the particles of the condensed dispersed phase in solid fuel combustion products plasma

International Nuclear Information System (INIS)

Samaryan, A.A.; Chernyshev, A.V.; Nefedov, A.P.; Petrov, O.F.; Fortov, V.E.; Mikhailov, Yu.M.; Mintsev, V.B.

2000-01-01

The results of experimental investigations of a type of dusty plasma which has been least studied--the plasma of solid fuel combustion products--were presented. Experiments to determine the parameters of the plasma of the combustion products of synthetic solid fuels with various compositions together with simultaneous diagnostics of the degree of ordering of the structures of the particles of the dispersed condensed phase were performed. The measurements showed that the charge composition of the plasma of the solid fuels combustion products depends strongly on the easily ionized alkali-metal impurities which are always present in synthetic fuel in one or another amount. An ordered arrangement of the particles of a condensed dispersed phase in structures that form in a boundary region between the high-temperature and condensation zones was observed for samples of aluminum-coated solid fuels with a low content of alkali-metal impurities

performance calculations of gadolinium oxide and boron nitride coated fuel

International Nuclear Information System (INIS)

Tanker, E.; Uslu, I.; Disbudak, H.; Guenduez, G.

1997-01-01

A comparative study was performed on the behaviour of natural uranium dioxide-gadolinium oxide mixture fuel and boron nitride coated low enriched fuel in a pressurized water reactor. A fuel element containing one burnable poison fuel pins was modeled with the computer code WIMS, and burn-up dependent critically, fissile isotope inventory and two dimensional power distribution were obtained. Calculations were performed for burnable poison fuels containing 5% and 10% gadolinium oxide and for those coated with 1μ,5μ and 10μ of boron nitride. Boron nitride coating was found superior to gadolinium oxide on account of its smoother criticality curve, lower power peaks and insignificant change in fissile isotope content

Irradiation testing of high-density uranium alloy dispersion fuels

International Nuclear Information System (INIS)

Hayes, S.L.; Trybus, C.L.; Meyer, M.K.

1997-01-01

Two irradiation test vehicles have been designed, fabricated, and inserted into the Advanced Test Reactor in Idaho. Irradiation of these experiments began in August 1997. These irradiation tests were designed to obtain irradiation performance information on a variety of potential new, high-density dispersion fuels. Each of the two irradiation vehicles contains 32 'microplates'. Each microplate is aluminum clad, having an aluminum matrix phase and containing one of the following compositions as the fuel phase: U-10Mo, U-8Mo, U-6Mo, U-4Mo, U-9Nb-3Zr, U-6Nb-4Zr, U-5Nb-3Zr, U-6Mo-1Pt, U-6Mo-0.6Ru, U10Mo-0.05Sn, U2Mo, or U 3 Si 2 . These experiments will be discharged at peak fuel burnups of approximately 40 and 80 at.% U 235 . Of particular interest are the extent of reaction of the fuel and matrix phases and the fission gas retention/swelling characteristics of these new fuel alloys. This paper presents the design of the irradiation vehicles and the irradiation conditions. (author)

Development oxide dispersion strengthened ferritic steels for fusion

Energy Technology Data Exchange (ETDEWEB)

Mukhopadhyay, D.K.; Froes, F.H.; Gelles, D.S. [Pacific Northwest National Lab., Richland, WA (United States)

1997-04-01

Uniaxial tension creep response is reported for an oxide dispersion strengthened (ODS) steel, Fe-13.5Cr-2W-0.5Ti-0.25 Y{sub 2}O{sub 3} (in weight percent) manufactured using the mechanical alloying process. Acceptable creep response is obtained at 900{degrees}C.

Electrode Design for Low Temperature Direct-Hydrocarbon Solid Oxide Fuel Cells

Science.gov (United States)

Chen, Fanglin (Inventor); Zhao, Fei (Inventor); Liu, Qiang (Inventor)

2015-01-01

In certain embodiments of the present disclosure, a solid oxide fuel cell is described. The solid oxide fuel cell includes a hierarchically porous cathode support having an impregnated cobaltite cathode deposited thereon, an electrolyte, and an anode support. The anode support includes hydrocarbon oxidation catalyst deposited thereon, wherein the cathode support, electrolyte, and anode support are joined together and wherein the solid oxide fuel cell operates a temperature of 600.degree. C. or less.

Electrode design for low temperature direct-hydrocarbon solid oxide fuel cells

Science.gov (United States)

Chen, Fanglin; Zhao, Fei; Liu, Qiang

2015-10-06

In certain embodiments of the present disclosure, a solid oxide fuel cell is described. The solid oxide fuel cell includes a hierarchically porous cathode support having an impregnated cobaltite cathode deposited thereon, an electrolyte, and an anode support. The anode support includes hydrocarbon oxidation catalyst deposited thereon, wherein the cathode support, electrolyte, and anode support are joined together and wherein the solid oxide fuel cell operates a temperature of 600.degree. C. or less.

Development of advanced mixed oxide fuels for plutonium management

International Nuclear Information System (INIS)

Eaton, S.; Beard, C.; Buksa, J.; Butt, D.; Chidester, K.; Havrilla, G.; Ramsey, K.

1997-01-01

A number of advanced Mixed Oxide (MOX) fuel forms are currently being investigated at Los Alamos National Laboratory that have the potential to be effective plutonium management tools. Evolutionary Mixed Oxide (EMOX) fuel is a slight perturbation on standard MOX fuel, but achieves greater plutonium destruction rates by employing a fractional nonfertile component. A pure nonfertile fuel is also being studied. Initial calculations show that the fuel can be utilized in existing light water reactors and tailored to address different plutonium management goals (i.e., stabilization or reduction of plutonium inventories residing in spent nuclear fuel). In parallel, experiments are being performed to determine the feasibility of fabrication of such fuels. Initial EMOX pellets have successfully been fabricated using weapons-grade plutonium. (author)

Development of advanced mixed oxide fuels for plutonium management

International Nuclear Information System (INIS)

Eaton, S.; Beard, C.; Buksa, J.; Butt, D.; Chidester, K.; Havrilla, G.; Ramsey, K.

1997-06-01

A number of advanced Mixed Oxide (MOX) fuel forms are currently being investigated at Los Alamos National Laboratory that have the potential to be effective plutonium management tools. Evolutionary Mixed Oxide (EMOX) fuel is a slight perturbation on standard MOX fuel, but achieves greater plutonium destruction rates by employing a fractional nonfertile component. A pure nonfertile fuel is also being studied. Initial calculations show that the fuel can be utilized in existing light water reactors and tailored to address different plutonium management goals (i.e., stabilization or reduction of plutonium inventories residing in spent nuclear fuel). In parallel, experiments are being performed to determine the feasibility of fabrication of such fuels. Initial EMOX pellets have successfully been fabricated using weapons-grade plutonium

Plasticity of oxide dispersion strengthened ferritic alloys

International Nuclear Information System (INIS)

Zakine, C.; Prioul, C.; Alamo, A.; Francois, D.

1993-01-01

Two 13%Cr oxide dispersion strengthened (ODS) ferritic alloys, DT and DY, exhibiting different oxide particle size distribution and a χ phase precipitation were studied. Their tensile properties have been tested from 20 to 700 C. Experimental observations during room temperature tensile tests performed in a scanning electronic microscope have shown that the main damage mechanism consists in microcracking of the χ phase precipitates on grain boundaries. These alloys are high tensile and creep resistant between 500 and 700 C. Their strongly stress-sensitive creep behaviour can be described by usual creep laws and incorporating a threshold stress below which the creep rate is negligible. (orig.)

Core-shell rhodium sulfide catalyst for hydrogen evolution reaction / hydrogen oxidation reaction in hydrogen-bromine reversible fuel cell

Science.gov (United States)

Li, Yuanchao; Nguyen, Trung Van

2018-04-01

Synthesis and characterization of high electrochemical active surface area (ECSA) core-shell RhxSy catalysts for hydrogen evolution oxidation (HER)/hydrogen oxidation reaction (HOR) in H2-Br2 fuel cell are discussed. Catalysts with RhxSy as shell and different percentages (5%, 10%, and 20%) of platinum on carbon as core materials are synthesized. Cyclic voltammetry is used to evaluate the Pt-equivalent mass specific ECSA and durability of these catalysts. Transmission electron microscopy (TEM), X-ray Photoelectron spectroscopy (XPS) and Energy-dispersive X-ray spectroscopy (EDX) techniques are utilized to characterize the bulk and surface compositions and to confirm the core-shell structure of the catalysts, respectively. Cycling test and polarization curve measurements in the H2-Br2 fuel cell are used to assess the catalyst stability and performance in a fuel cell. The results show that the catalysts with core-shell structure have higher mass specific ECSA (50 m2 gm-Rh-1) compared to a commercial catalyst (RhxSy/C catalyst from BASF, 6.9 m2 gm-Rh-1). It also shows better HOR/HER performance in the fuel cell. Compared to the platinum catalyst, the core-shell catalysts show more stable performance in the fuel cell cycling test.

Characterization of fuel miniplates fabricated with U(Mo) particles dispersed in Al-Si matrices

International Nuclear Information System (INIS)

Arico, S F; Mirandou, M I; Balart, S N; Fabro, J O

2012-01-01

In 2011 ECRI facility (Depto. ECRI, GCCN, CNEA) restarted the development for the fabrication of dispersion miniplates fuel elements in Al-Si matrix. This miniplates are fabricated with atomized U-7wt%Mo particles dispersed in a matrix formed by a mixture of pure Al and pure Si powders. The first results for an Al-4wt%Si matrix were presented at the AATN 2011 Annual Meeting. In this work, new results from the microstructural characterization of the meat in Al- 2wt%Si and pure Al miniplates are presented and compared with the previous ones. It is the intention to study the influence of the fabrication parameters as well as different Si concentration in the matrix, on the formation and characteristics of the interaction layer formed between the particles and the matrix at the end of the fabrication process. According to the results presented in this work an improvement can be observed on miniplates with Al-Si matrix respect to the one with pure Al. On the miniplates with Al- Si matrix, almost 100 % of the U(Mo) particles presented, at least in some fraction of its surface, an interaction layer composed by phases that contain Si. Moreover its morphological characteristics are independent of the crystallographic state of the U(Mo) particles. However, the oxide layer formed on the U(Mo) during the hot rolling acts as a barrier to the formation of the interaction layer. As a consequence, it is then mandatory to introduce some changes on the fabrication parameters to avoid, or at least minimize, this oxide layer (author)

POST CRITICAL HEAT TRANSFER AND FUEL CLADDING OXIDATION

Directory of Open Access Journals (Sweden)

Vojtěch Caha

2016-12-01

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

Irradiation testing of high density uranium alloy dispersion fuels

International Nuclear Information System (INIS)

Hayes, S.L.; Trybus, C.L.; Meyer, M.K.

1997-10-01

Two irradiation test vehicles have been designed, fabricated, and inserted into the Advanced Test Reactor in Idaho. Irradiation of these experiments began in August 1997. These irradiation tests were designed to obtain irradiation performance information on a variety of potential new, high-density dispersion fuels. Each of the two irradiation vehicles contains 32 microplates. Each microplate is aluminum clad, having an aluminum matrix phase and containing one of the following compositions as the fuel phase: U-10Mo, U-8Mo, U-6Mo, U-4Mo, U-9Nb-3Zr, U-6Nb-4Zr, U-5Nb-3Zr, U-6Mo-1Pt, U-6Mo-0.6Ru, U-10Mo-0.05Sn, U 2 Mo, or U 3 Si 2 . These experiments will be discharged at peak fuel burnups of 40% and 80%. Of particular interest is the fission gas retention/swelling characteristics of these new fuel alloys. This paper presents the design of the irradiation vehicles and the irradiation conditions

Contribution to the study of second phases particles dispersion in polycrystalline uranium dioxide

International Nuclear Information System (INIS)

Peres, V.

1994-06-01

To reduce fission gas release of irradiated polycrystalline uranium dioxide, the dispersion of intragranular nanometric particles of second phase necessary to pin gas bubbles may complete the advantage of a large-grained fuel microstructure. Moreover, intergranular glass films may improve high temperatures mechanical properties of UO 2 . In this study, mixtures of additives composed of ''corindon'' structure oxides that enhance the fuel grain growth and composed of different oxides with variable solid solubilities in the UO 2 matrix were achieved. Additives with a negligible solubility inhibit grain boundaries motion except those, such as silica, that involve a liquid phase at the sintering temperature. Rare earth oxides that form stable solid solutions with UO 2 cannot lead to precipitation, but have no effect on the fuel grain growth doped with ''corindon'' type oxides. A chromium oxide excess allows the creation of a fuel microstructure described by large grains and intragranular spherical Cr 2 O 3 inclusions observed by scanning electron microscopy. Values for the bulk lattice diffusion coefficient of Cr 3+ cations in UO 2 can be deduced from the experimental growth of those dispersed particles by an Ostwald ripening mechanism. The formation of small precipitated metal particles inside the uranium dioxide matrix induced by the internal reduction of a solid solution has not been performed. However, direct reduction of insoluble chromium oxide particles is easy and produces metallic intragranular precipitates. (author). 119 refs., 112 figs., 33 tabs., 5 annexes

Preparation of Dispersion-Hardened Copper by Internal Oxidation

DEFF Research Database (Denmark)

Brøndsted, Povl; Sørensen, Ole Toft

1978-01-01

Internal oxidation experiments in CO2/CO atmospheres on Cu-Al alloys for preparation of dispersion-hardened Cu are described. The oxygen pressures of the atmospheres used in the experiments were controlled with a solid electrolyte oxygen cell based on ZrO2 (CaO). The particle size distributions o...

Fuel-sodium reaction product formation in breached mixed-oxide fuel

International Nuclear Information System (INIS)

Bottcher, J.H.; Lambert, J.D.B.; Strain, R.V.; Ukai, S.; Shibahara, S.

1988-01-01

The run-beyond-cladding-breach (RBCB) operation of mixed-oxide LMR fuel pins has been studied for six years in the Experimental Breeder Reactor-II (EBR-II) as part of a joint program between the US Department of Energy and the Power Reactor and Nuclear Fuel Development Corporation of Japan. The formation of fuel-sodium reaction product (FSRP), Na 3 MO 4 , where M = U/sub 1-y/Pu/sub y/, in the outer fuel regions is the major phenomenon governing RBCB behavior. It increases fuel volume, decreases fuel stoichiometry, modifies fission-product distributions, and alters thermal performance of a pin. This paper describes the morphology of Na 3 MO 4 observed in 5.84-mm diameter pins covering a variety of conditions and RBCB times up to 150 EFPD's. 8 refs., 1 fig

Relative neutronic performance of proposed high-density dispersion fuels in water-moderated and D2O-reflected research reactors

International Nuclear Information System (INIS)

Bretscher, M.M.; Matos, J.E.; Snelgrove, J.L.

1996-01-01

This paper provides an overview of the neutronic performance of an idealized research reactor using several high density LEU fuels that are being developed by the RERTR program. High-density LEU dispersion fuels are needed for new and existing high-performance research reactors and to extend the lifetime of fuel elements in other research reactors. This paper discusses the anticipated neutronic behavior of proposed advanced fuels containing dispersions of U 3 Si 2 , UN, U 2 Mo and several uranium alloys with Mo, or Zr and Nb. These advanced fuels are ranked based on the results of equilibrium depletion calculations for a simplified reactor model having a small H 2 O-cooled core and a D 2 O reflector. Plans have been developed to fabricate and irradiate several uranium alloy dispersion fuels in order to test their stability and compatibility with the matrix material and to establish practical loading limits

Relative neutronic performance of proposed high-density dispersion fuels in water-moderated and D2O-reflected research reactors

International Nuclear Information System (INIS)

Bretscher, M.M.; Matos, J.E.; Snelgrove, J.L.

1996-01-01

This paper provides an overview of the neutronic performance of an idealized research reactor using several high density Leu fuels that are being developed by the Rarita program. High-density Leu dispersion fuels are needed for new and existing high-performance research reactors and to extend the lifetime of fuel elements in other research reactors. This paper discusses the anticipated neutronic behavior of proposed advanced fuels containing dispersions of U 3 Si 2 , UN, U 2 Mo and several uranium alloys with Mo, or Zr and Nb. These advanced fuels are ranked based on the results of equilibrium depletion calculations for a simplified reactor model having a small H 2 O-cooled core and a D 2 O reflector. Plans have been developed to fabricate and irradiate several uranium alloy dispersion fuels in order to test their stability and compatibility with the matrix material and to establish practical loading limits. (author)

Durability of solid oxide fuel cells using sulfur containing fuels

DEFF Research Database (Denmark)

Hagen, Anke; Rasmussen, Jens Foldager Bregnballe; Thydén, Karl Tor Sune

2011-01-01

The usability of hydrogen and also carbon containing fuels is one of the important advantages of solid oxide fuel cells (SOFCs), which opens the possibility to use fuels derived from conventional sources such as natural gas and from renewable sources such as biogas. Impurities like sulfur compounds...... are critical in this respect. State-of-the-art Ni/YSZ SOFC anodes suffer from being rather sensitive towards sulfur impurities. In the current study, anode supported SOFCs with Ni/YSZ or Ni/ScYSZ anodes were exposed to H2S in the ppm range both for short periods of 24h and for a few hundred hours. In a fuel...

Addressing fuel recycling in solid oxide fuel cell systems fed by alternative fuels

DEFF Research Database (Denmark)

Rokni, Masoud

2017-01-01

An innovative study on anode recirculation in solid oxide fuel cell systems with alternative fuels is carried out and investigated. Alternative fuels under study are ammonia, pure hydrogen, methanol, ethanol, DME and biogas from biomass gasification. It is shown that the amount of anode off......%. Furthermore, it is founded that for the case with methanol, ethanol and DME then at high utilization factors, low anode recirculation is recommended while at low utilization factors, high anode recirculation is recommended. If the plant is fed by biogas from biomass gasification then for each utilization...

Identification of a Methane Oxidation Intermediate on Solid Oxide Fuel Cell Anode Surfaces with Fourier Transform Infrared Emission.

Science.gov (United States)

Pomfret, Michael B; Steinhurst, Daniel A; Owrutsky, Jeffrey C

2013-04-18

Fuel interactions on solid oxide fuel cell (SOFC) anodes are studied with in situ Fourier transform infrared emission spectroscopy (FTIRES). SOFCs are operated at 800 °C with CH4 as a representative hydrocarbon fuel. IR signatures of gas-phase oxidation products, CO2(g) and CO(g), are observed while cells are under load. A broad feature at 2295 cm(-1) is assigned to CO2 adsorbed on Ni as a CH4 oxidation intermediate during cell operation and while carbon deposits are electrochemically oxidized after CH4 operation. Electrochemical control provides confirmation of the assignment of adsorbed CO2. FTIRES has been demonstrated as a viable technique for the identification of fuel oxidation intermediates and products in working SOFCs, allowing for the elucidation of the mechanisms of fuel chemistry.

Pt20RuxSny nanoparticles dispersed on mesoporous carbon CMK-3 and their application in the oxidation of 2-carbon alcohols and fermentation effluent

International Nuclear Information System (INIS)

Lo, An-Ya; Chung, Yi-Chen; Hung, Wei-Hsuan; Hsu, Yun-Chi; Tseng, Chuan-Ming; Zhang, Wei-Lun; Wang, Fu-Kai; Lin, Chiu-Yue

2017-01-01

Highlights: • Pt 20 Ru x Sn y @C catalysts are formed by dispersing Pt-Sn and Pt-Ru-Sn NPs on CMK-3. • They are tested in fuel cells using ethanol, ethylene glycol, and CFHPE as fuels. • Higher Sn contents improve catalytic efficiency of Pt 20 Ru x Sn y when x = 0 or x = 10. • Role of Sn in C−C bond cleavage and improving poisoning tolerance is explained. • Pt 20 Ru 10 Sn 15 @C is used to show feasibility of using bioalcohol from CFHPE as fuel. - Abstract: We report the synthesis of Pt-Sn binary and Pt-Ru-Sn ternary alloy nanoparticles (NPs) dispersed on mesoporous carbon CMK-3 for bioalcohol fuel cell applications where ethanol, ethylene glycol, and fermentative hydrogen production effluent were used as the fuels. The proposed alloy electrocatalysts, denoted as Pt 20 Ru x Sn y @C (where 20, x, and y represent the weight fractions of Pt, Ru, and Sn, respectively), were examined using scanning electron microscopy, energy-dispersive X-ray spectroscopy mapping, transmission electron microscopy, Brunauer-Emmett-Teller measurements, X-ray diffraction analysis, and electrochemical measurements, in order to determine their morphologies, microstructures, compositions, phase structures, and electrochemical characteristics. The effects of the Sn content on the following factors were examined: 1) average particle size of the alloy NPs, 2) mesoporosity, 3) electrochemically active surfaces of Pt 20 Ru x Sn y @C, and 4) ethanol oxidation reaction and ethylene glycol oxidation reaction activities. Higher Sn contents improved the catalytic efficiency of Pt 20 Ru x Sn y when x = 0 or x = 10, with the optimized compositions being Pt 20 Sn 30 and Pt 20 Ru 10 Sn 15 for the binary and ternary alloys, respectively. Based on the ethanol and ethylene glycol oxidation reactions, we explain the role of Sn in promoting C−C bond cleavage and in improving catalyst tolerance against poisoning. Overall, for both the ethanol system and the ethylene glycol system, the catalytic

Highly-dispersed Ta-oxide catalysts prepared by electrodeposition in a non-aqueous plating bath for polymer electrolyte fuel cell cathodes

KAUST Repository

Seo, Jeongsuk; Cha, Dong Kyu; Takanabe, Kazuhiro; Kubota, Jun; Domen, Kazunari

2012-01-01

The Ta-oxide cathode catalysts were prepared by electrodeposition in a non-aqueous solution. These catalysts showed excellent catalytic activity and have an onset potential of 0.92 V RHE for the oxygen reduction reaction (ORR). The highly-dispersed Ta species at the nanometer scale on the carbon black was an important contributor to the high activity. © 2012 The Royal Society of Chemistry.

PLACA/DPLACA: a code to simulate the behavior of a monolithic/dispersed plate type fuel

International Nuclear Information System (INIS)

Denis, Alicia; Soba, Alejandro

2005-01-01

The PLACA code was originally built to simulate monolithic plate fuels contained in a metallic cladding, with a gap in between. The international program of high density fuels was recently oriented to the development of a plate-type fuel of a uranium rich alloy with a molybdenum content between 6 to 10 w %, without gap and with a Zircaloy cladding. To give account of these fuels, the DPLACA code was elaborated as a modification of the original code. The extension of the calculation tool to disperse fuels involves a detailed study of the properties and models (still in progress). Of special interest is the material formed by U Mo particles dispersed in an Al matrix. This material has appeared as a candidate fuel for high flux research reactors. However, the interaction layer that grows around the particles has a deleterious effect on the material performance in operation conditions and may represent a limit for its applicability. A number of recent experiments carried out on this material provide abundant information that allows testing of the numerical models. (author)

Comparison of Core Performance with Various Oxide fuels on Sodium Cooled Fast Reactor

Energy Technology Data Exchange (ETDEWEB)

Choi, Jin Ha; Kim, Myung Hyun [Kyung Hee University, Yongin (Korea, Republic of)

2016-05-15

The system is called Prototype GenIV Sodium-cooled Fast Reactor (PGSFR). Ultimate goal of PGSFR is test for capability of TRU transmutation. Purpose of this study is test for evaluation of in-core performance and TRU transmutation performance by applying various oxide fuel loaded TRU. Fuel type of reference core is changed to uranium-based oxide fuel. Oxide fuel has a lot of experience through fuel fabrication and reactor operation. This study performed by compared and analyzed a core performance of various oxide fuels. (U,Pu)O{sub 2} and (U,TRU)O{sub 2} which various oxide fuel types are selected as extreme case for comparison with core performance and transmutation capability of TRU isotopes. Thorium-based fuel is known that it has good performance for burner reactor due to low proliferation characteristic. To check the performance of TRU incineration for comparison with uranium-based fuel on prototype SFR, Thorium-based fuel, (Th,U)O{sub 2}, (Th,Pu)O{sub 2} and (Th,TRU)O{sub 2}, is selected. Calculations of core performance for various oxide fuel are performed using the fast calculation tool, TRANSX / DANTSTS / REBUS-3. In this study, comparison of core performance and transmutation performance is conducted with various fuel types in a sodium-cooled fast reactor. Mixed oxide fuel with TRU can produce the energy with small amount of fissile material. However, the TRU fuel is confirmed to bring a potential decline of the safety parameters. In case of (Th,U)O2 fuel, the flux level in thermal neutron region becomes lower because of higher capture cross-section of Th-232 than U-238. However, Th-232 has difficulty in converting to TRU isotopes. Therefore, the TRU consumption mass is relatively high in mixed oxide fuel with thorium and TRU.

Comparison of Core Performance with Various Oxide fuels on Sodium Cooled Fast Reactor

International Nuclear Information System (INIS)

Choi, Jin Ha; Kim, Myung Hyun

2016-01-01

The system is called Prototype GenIV Sodium-cooled Fast Reactor (PGSFR). Ultimate goal of PGSFR is test for capability of TRU transmutation. Purpose of this study is test for evaluation of in-core performance and TRU transmutation performance by applying various oxide fuel loaded TRU. Fuel type of reference core is changed to uranium-based oxide fuel. Oxide fuel has a lot of experience through fuel fabrication and reactor operation. This study performed by compared and analyzed a core performance of various oxide fuels. (U,Pu)O_2 and (U,TRU)O_2 which various oxide fuel types are selected as extreme case for comparison with core performance and transmutation capability of TRU isotopes. Thorium-based fuel is known that it has good performance for burner reactor due to low proliferation characteristic. To check the performance of TRU incineration for comparison with uranium-based fuel on prototype SFR, Thorium-based fuel, (Th,U)O_2, (Th,Pu)O_2 and (Th,TRU)O_2, is selected. Calculations of core performance for various oxide fuel are performed using the fast calculation tool, TRANSX / DANTSTS / REBUS-3. In this study, comparison of core performance and transmutation performance is conducted with various fuel types in a sodium-cooled fast reactor. Mixed oxide fuel with TRU can produce the energy with small amount of fissile material. However, the TRU fuel is confirmed to bring a potential decline of the safety parameters. In case of (Th,U)O2 fuel, the flux level in thermal neutron region becomes lower because of higher capture cross-section of Th-232 than U-238. However, Th-232 has difficulty in converting to TRU isotopes. Therefore, the TRU consumption mass is relatively high in mixed oxide fuel with thorium and TRU.

Visualization Study of Melt Dispersion Behavior for SFR with a Metallic Fuel under Severe Accidents

Energy Technology Data Exchange (ETDEWEB)

Heo, Hyo Heo; Park, Seong Dae; Bang, In Cheol [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of); Jerng, Dong Wook [Jungang Univ., Seoul (Korea, Republic of)

2015-05-15

The safety strategy provides negative reactivity driven by the melt dispersal, so it could reduce the possibility of the recriticality event under a severe triple or more fault scenario for SFR. Since the behavior of the melt dispersion is unpredictable, it depends on the accident condition, particularly core region. While the voided coolant channel region is usually developed in the inner core, the unvoided coolant channel region is formed in the outer core. It is important to confirm the fuel dispersion with the core region, but there are not sufficient existing studies for them. From the existing studies, the coolant vapor pressure is considered as one of driving force to move the melt towards outside of the core. There is a complexity of the phenomena during intermixing of the melt with the coolant after the horizontal melt injections. It is too difficult to understand the several combined mechanisms related to the melt dispersion and the fragmentation. The specific conditions to be well dispersed for the molten metallic fuel were discussed in the experiments with the simulant materials. The each melt behavior was compared to evaluate the melt dispersion under the coolant void condition and the boiling condition.

Analysis of fuel cladding chemical interaction in mixed oxide fuel pins

International Nuclear Information System (INIS)

Weber, J.W.; Dutt, D.S.

1976-01-01

An analysis is presented of the observed interaction between mixed oxide 75 wt percent UO 2 --25 wt percent PuO 2 fuel and 316--20 percent CW stainless steel cladding in LMFBR type fuel pins irradiated in EBR-II. A description is given of the test pins and their operating conditions together with, metallographic observations and measurements of the fuel/cladding reaction, and a correlation equation is developed relating depth of cladding attack to temperature and burnup. Some recent data on cladding reaction in fuel pins with low initial O/M in the fuel are given and compared with the correlation equation curves

The neutronic and fuel cycle performance of interchangeable 3500 MWth metal and oxide fueled LMRs

International Nuclear Information System (INIS)

Fujita, E.K.; Wade, D.C.

1990-01-01

This study summarizes the neutronic and fuel cycle analysis performed at Argonne National Laboratory for an oxide and a metal fueled 3500 MWth LMR. These reactor designs formed the basis for a joint US/European study of LMR ATWS events. The oxide and metal core designs were developed to meet reactor performance specifications that are constrained by requirements for core loading interchangeability and for a small burnup reactivity swing. Differences in the computed performance parameters of the oxide and metal cores, arising from basic differences in their neutronic characteristics, are identified and discussed. It is shown that metal and oxide cores designed to the same ground rules exhibit many similar performance characteristics; however, they differ substantially in reactivity coefficients, control strategies, and fuel cycle options. 12 refs., 2 figs., 12 tabs

A global-scale dispersion analysis of iodine-129 from nuclear fuel reprocessing plants

International Nuclear Information System (INIS)

Nishizawa, Masato; Suzuki, Takashi; Nagai, Haruyasu; Togawa, Orihiko

2010-01-01

A three-dimensional global chemical transport model, MOZART-2, is applied to investigate the global-sale dispersion of Iodine-129 from nuclear fuel reprocessing plants. The concentration and deposition of 129 I obtained by MOZART-2 are dispersed all over the Northern Hemisphere. The emission of 129 I to the atmosphere is thus important in considering the transport of 129 I to remote sites. (author)

Oxide thickness measurement for monitoring fuel performance at high burnup

International Nuclear Information System (INIS)

Jaeger, M.A.; Van Swam, L.F.P.; Brueck-Neufeld, K.

1991-01-01

For on-site monitoring of the fuel performance at high burnup, Advanced Nuclear Fuels uses the linear scan eddy current method to determine the oxide thickness of irradiated Zircaloy fuel cans. Direct digital data acquisition methods are employed to collect the data on magnetic storage media. This field-proven methodology allows oxide thickness measurements and rapid interpretation of the data during the reactor outages and makes it possible to immediately reinsert the assemblies for the next operating cycle. The accuracy of the poolside measurements and data acquisition/interpretation techniques have been verified through hot cell metallographic measurements of rods previously measured in the fuel pool. The accumulated data provide a valuable database against which oxide growth models have been benchmarked and allow for effective monitoring of fuel performance. (orig.) [de

Critical experiments with mixed oxide fuel

International Nuclear Information System (INIS)

Harris, D.R.

1997-01-01

This paper very briefly outlines technical considerations in performing critical experiments on weapons-grade plutonium mixed oxide fuel assemblies. The experiments proposed would use weapons-grade plutonium and Er 2 O 3 at various dissolved boron levels, and for specific fuel assemblies such as the ABBCE fuel assembly with five large water holes. Technical considerations described include the core, the measurements, safety, security, radiological matters, and licensing. It is concluded that the experiments are feasible at the Rensselaer Polytechnic Institute Reactor Critical Facility. 9 refs

Low temperature spent fuel oxidation under tuff repository conditions

International Nuclear Information System (INIS)

Einziger, R.E.; Woodley, R.E.

1985-01-01

The Nevada Nuclear Waste Storage Investigations Project is studying the suitability of tuffaceous rocks at Yucca Mountain, Nye County, Nevada, for high level waste disposal. The oxidation state of LWR spent fuel in a tuff repository may be a significant factor in determining its ability to inhibit radionuclide migration. Long term exposure at low temperatures to the moist air expected in a tuff repository is expected to increase the oxidation state of the fuel. A program is underway to determine the spent fuel oxidation mechanisms which might be active in a tuff repository. Initial work involves a series of TGA experiments to determine the effectiveness of the technique and to obtain preliminary oxidation data. Tests were run at 200 0 C and 225 0 C for as long as 720 hours. Grain boundary diffusion appears to open up a greater surface area for oxidation prior to onset of bulk diffusion. Temperature strongly influences the oxidation rates. The effect of moisture is small but readily measurable. 25 refs., 7 figs., 4 tabs

Dissolution of mixed oxide fuel as a function of fabrication variables

International Nuclear Information System (INIS)

Lerch, R.E.

1979-08-01

Dissolution properties of mechanically blended mixed oxide fuel were very dependent on the six fuel fabrication variables studied. Fuel sintering temperature, source of PuO 2 and PuO 2 content of the fuel had major effects: (1) as the sintering temperature was increased from 1400 to 1700 0 C, pellet dissolution was more complete; (2) pellets made from burned metal derived PuO 2 were more completely dissolved than pellets made from calcined nitrate derived PuO 2 which in turn were more completely dissolved than pellets made from calcined nitrate derived PuO 2 ; (3) as the PuO 2 content decreased from 25 to 15 wt % PuO 2 , pellet dissolution was more complete. Preferential dissolution of uranium occurred in all the mechanically blended mixed oxide. Unirradiated mixed oxide fuel pellets made by the Sol Gel process were generally quite soluble in nitric acid. Unirradiated mixed oxide fuel pellets made by the coprecipitation process dissolved completely and rapidly in nitric acid. Fuel made by the coprecipitation process was more completely dissolved than fuel made by the Sol Gel process which, in turn, was more completely dissolved than fuel made by mechanically blending UO 2 and PuO 2 as shown below. Addition of uncomplexed fluoride to nitric acid during fuel dissolution generally rendered all fuel samples completely dissolvable. In boiling 12M nitric acid, 95 to 99% of the plutonium which was going to dissolve did so in the first hour. Irradiated mechanically blended mixed oxide fuel with known fuel fabrication conditions was also subjected to fuel dissolution tests. While irradiation was shown to increase completeness of plutonium dissolution, poor dissolubility due to adverse fabrication conditions (e.g., low sintering temperature) remained after irradiation

Reversible solid oxide fuel cells (R-SOFCs) with chemically stable proton-conducting oxides

KAUST Repository

Bi, Lei

2015-07-01

Proton-conducting oxides offer a promising way of lowering the working temperature of solid oxide cells to the intermediate temperate range (500 to 700. °C) due to their better ionic conductivity. In addition, the application of proton-conducting oxides in both solid oxide fuel cells (SOFCs) and sold oxide electrolysis cells (SOECs) provides unique advantages compared with the use of conventional oxygen-ion conducting conductors, including the formation of water at the air electrode site. Since the discovery of proton conduction in some oxides about 30. years ago, the development of proton-conducting oxides in SOFCs and SOECs (the reverse mode of SOFCs) has gained increased attention. This paper briefly summarizes the development in the recent years of R-SOFCs with proton-conducting electrolytes, focusing on discussing the importance of adopting chemically stable materials in both fuel cell and electrolysis modes. The development of electrode materials for proton-conducting R-SOFCs is also discussed. © 2015 Elsevier B.V.

Development and Characterization of Gas Diffusion Layer Using Carbon Slurry Dispersed by Ammonium Lauryl Sulfate for Proton Exchange Member Fuel Cells

Science.gov (United States)

Villacorta, Rashida

Gas diffusion layers (GDLs) are a critical and essential part of proton exchange membrane fuel cells (PEMFCs). They carry out various important functions such as transportation of reactants to and from the reaction sites. The material properties and structural characteristics of the substrate and the microporous layer strongly influence fuel cell performance. The microporous layer of the GDLs was fabricated with the carbon slurry dispersed in water containing ammonium lauryl sulfate (ALS) using the wire rod coating method. GDLs were fabricated with different materials to compose the microporous layer and evaluated the effects on PEMFC power output performance. The consistency of the carbon slurry was achieved by adding 25 wt. % of PTFE, a binding agent with a 75:25 ratio of carbon (Pureblack and vapor grown carbon fiber). The GDLs were investigated in PEMFC under various relative humidity (RH) conditions using H2/O2 and H2/Air. GDLs were also fabricated with the carbon slurry dispersed in water containing sodium dodecyl sulfate (SDS) and multiwalled carbon nanotubes (MWCNTs) with isopropyl alcohol (IPA) based for fuel cell performance comparison. MWCNTs and SDS exhibits the highest performance at 60% and 70% RH with a peak power density of 1100 mW.cm-2 and 850 mW.cm-2 using air and oxygen as an oxidant. This means that the gas diffusion characteristics of these two samples were optimum at 60 and 70 % RH with high limiting current density range. It was also found that the composition of the carbon slurry, specifically ALS concentration has the highest peak power density of 1300 and 500mW.cm-2 for both H2/O 2 and H2/Air at 100% RH. However, SDS and MWCNTs demonstrates the lowest power density using air and oxygen as an oxidants at 100% RH.

Nuclear fuel element

International Nuclear Information System (INIS)

Yamamoto, Seigoro.

1994-01-01

Ultrafine particles of a thermal neutron absorber showing ultraplasticity is dispersed in oxide ceramic fuels by more than 1% to 10% or lower. The ultrafine particles of the thermal neutron absorber showing ultrafine plasticity is selected from any one of ZrGd, HfEu, HfY, HfGd, ZrEu, and ZrY. The thermal neutron absorber is converted into ultrafine particles and solid-solubilized in a nuclear fuel pellet, so that the dispersion thereof into nuclear fuels is made uniform and an absorbing performance of the thermal neutrons is also made uniform. Moreover, the characteristics thereof, for example, physical properties such as expansion coefficient and thermal conductivity of the nuclear fuels are also improved. The neutron absorber, such as ZrGd or the like, can provide plasticity of nuclear fuels, if it is mixed into the nuclear fuels for showing the plasticity. The nuclear fuel pellets are deformed like an hour glass as burning, but, since the end portion thereof is deformed plastically within a range of a repulsive force of the cladding tube, there is no worry of damaging a portion of the cladding tube. (N.H.)

Experimental studies of thermal and chemical interactions between molten aluminum and nuclear dispersion fuels with water

International Nuclear Information System (INIS)

Farahani, A.A.

1997-01-01

Because of the possibility of rapid physical and chemical molten fuel-water interactions during a core melt accident in noncommercial or experimental reactors, it is important to understand the interactions that might occur if these materials were to contact water. An existing vertical 1-D shock tube facility was improved and a gas sampling device to measure the gaseous hydrogen in the upper chamber of the shock tube was designed and built to study the impact of a water column driven downward by a pressurized gas onto both molten aluminum (6061 alloy) and oxide and silicide depleted nuclear dispersion fuels in aluminum matrices. The experiments were carried out with melt temperatures initially at 750 to 1,000 C and water at room temperature and driving pressures of 0.5 and 1 MPa. Very high transient pressures, in many cases even larger than the thermodynamic critical pressure of the water (∼ 20 MPa), were generated due to the interactions between the water and the crucible and its contents. The molten aluminum always reacted chemically with the water but the reaction did not increase consistently with increasing melt temperature. An aluminum ignition occurred when water at room temperature impacted 28.48 grams of molten aluminum at 980.3 C causing transient pressures greater than 69 MPa. No signs of aluminum ignition were observed in any of the experiments with the depleted nuclear dispersion fuels, U 3 O 8 -Al and U 3 Si 2 -Al. The greater was the molten aluminum-water chemical reaction, the finer was the debris recovered for a given set of initial conditions. Larger coolant velocities (larger driving pressures) resulted in more melt fragmentation but did not result in more molten aluminum-water chemical reaction. Decreasing the water temperature also resulted in more melt fragmentation and did not suppress the molten aluminum-water chemical reaction

Oxidative dissolution of ADOPT compared to standard UO2 fuel

International Nuclear Information System (INIS)

Nilsson, Kristina; Roth, Olivia; Jonsson, Mats

2017-01-01

In this work we have studied oxidative dissolution of pure UO 2 and ADOPT (UO 2 doped with Al and Cr) pellets using H 2 O 2 and gammaradiolysis to induce the process. There is a small but significant difference in the oxidative dissolution rate of UO 2 and ADOPT pellets, respectively. However, the difference in oxidative dissolution yield is insignificant. Leaching experiments were also performed on in-reactor irradiated ADOPT and UO 2 pellets under oxidizing conditions. The results indicate that the U(VI) release is slightly slower from the ADOPT pellet compared to the UO 2. This could be attributed to differences in exposed surface area. However, fission products with low UO 2 solubility display a higher relative release from ADOPT fuel compared to standard UO 2 -fuel. This is attributed to a lower matrix solubility imposed by the dopants in ADOPT fuel. The release of Cs is higher from UO 2 which is attributed to the larger grain size of ADOPT. - Highlights: •Oxidative dissolution of ADOPT fuel is compared to standard UO 2 fuel. •Only marginal differences are observed. •The main difference observed is in the relative release rate of fission products. •Differences are claimed to be attributed to a lower matrix solubility imposed by the dopants in ADOPT fuel.

Texture evolution in Oxide Dispersion Strengthened (ODS) steel tubes during pilgering process

Science.gov (United States)

Vakhitova, E.; Sornin, D.; Barcelo, F.; François, M.

2017-10-01

Oxide Dispersion Strengthened (ODS) steels are foreseen as fuel cladding material in the coming generation of Sodium Fast Reactors (SFR). Cladding tubes are manufactured by hot extrusion and subsequent cold forming steps. In this study, a 9 wt% Cr ODS steel exhibiting α-γ phase transformation at high temperature is cold formed under industrial conditions with a large section reduction in two pilgering steps. The influence of pilgering process parameters and intermediate heat treatment on the microstructure evolution is studied experimentally using Electron Backscattering Diffraction (EBSD) and X-ray Diffraction (XRD) methods. Pilgered samples show elongated grains and a high texture formation with a preferential orientation along the rolling direction. During the heat treatment, grain morphology is recovered from elongated grains to almost equiaxed ones, while the well-known α-fiber texture presents an unexpected increase in intensity. The remarkable temperature stability of this fiber is attributed to a crystallographic structure memory effect during phase transformations.

Comparison of thermal compatibility between atomized and comminuted U3Si dispersion fuels

International Nuclear Information System (INIS)

Ryu, Woo-Seog; Park, Jong-Man; Kim, Chang-Kyu; Kuk, II-Hyun

1997-01-01

Thermal compatibility of atomized U 3 Si dispersion fuels were evaluated up to 2600 hours in the temperature range from 250 to 500 degrees C, and compared with that of comminuted U 3 Si. Atomized U 3 Si showed better performance in terms of volume expansion of fuel meats. The reaction zone of U 3 Si and Al occurred along the grain boundaries and deformation bands in U 3 Si particles. Pores around fuel particles appeared at high temperature or after long-term annealing tests to remain diffusion paths over the trench of the pores. The constraint effects of cladding on fuel rod suppressed the fuel meat, and reduced the volume expansion

Iron oxide redox chemistry and nuclear fuel disposal

International Nuclear Information System (INIS)

Jobe, D.J.; Lemire, R.J.; Taylor, P.

1997-04-01

Solubility and stability data for iron (III) oxides and aqueous Fe(II) and Fe(III) species are reviewed, and selected values are used to calculate potential-pH diagrams for the iron system at temperatures of 25 and 100 deg C, chloride activities {C1 - } = 10 -2 and 1 mol/kg, total carbonate activity {C T } = 10 -3 mol/kg, and iron(III) oxide/oxyhydroxide solubility products (25 deg C values) K sp = {Fe 3+ }{OH - } 3 = 10 -38.5 , 10 -40 and 10 -42 . The temperatures and anion concentrations bracket the range of conditions expected in a Canadian nuclear fuel waste disposal vault. The three solubility products represent a conservative upper limit, a most probable value, and a minimum credible value, respectively, for the iron oxides likely to be important in controlling redox conditions in a disposal vault for CANDU nuclear reactor fuel. Only in the first of these three cases do the calculated redox potentials significantly exceed values under which oxidative dissolution of the fuel may occur. (author)

Rapid synthesis of platinum-ruthenium bimetallic nanoparticles dispersed on carbon support as improved electrocatalysts for ethanol oxidation.

Science.gov (United States)

Gu, Zhulan; Li, Shumin; Xiong, Zhiping; Xu, Hui; Gao, Fei; Du, Yukou

2018-07-01

Bimetallic nanocatalysts with small particle size benefit from markedly enhanced electrocatalytic activity and stability during small molecule oxidation. Herein, we report a facile method to synthesize binary Pt-Ru nanoparticles dispersed on a carbon support at an optimum temperature. Because of its monodispersed nanostructure, synergistic effects were observed between Pt and Ru and the PtRu/C electrocatalysts showed remarkably enhanced electrocatalytic activity towards ethanol oxidation. The peak current density of the Pt 1 Ru 1 /C electrocatalyst is 3731 mA mg -1 , which is 9.3 times higher than that of commercial Pt/C (401 mA mg -1 ). Furthermore, the synthesized Pt 1 Ru 1 /C catalyst exhibited higher stability during ethanol oxidation in an alkaline medium and maintained a significantly higher current density after successive cyclic voltammograms (CVs) of 500 cycles than commercial Pt/C. Our work highlights the significance of the reaction temperature during electrocatalyst synthesis, leading to enhanced catalytic performance towards ethanol oxidation. The Pt 1 Ru 1 /C electrocatalyst has great potential for application in direct ethanol fuel cells. Copyright © 2018 Elsevier Inc. All rights reserved.

High temperature oxidation test of oxide dispersion strengthened (ODS) steel claddings

International Nuclear Information System (INIS)

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

2006-07-01

In a feasibility study of ODS steel cladding, its high temperature oxidation resistance was evaluated. Although addition of Cr is effective for preventing high temperature oxidation, excessively higher amount of Cr leads to embrittlement due to the Cr-rich α' precipitate formation. In the ODS steel developed by the Japan Atomic Energy Agency (JAEA), the Cr content is controlled in 9Cr-ODS martensite and 12Cr-ODS ferrite. In this study, high temperature oxidation test was conducted for ODS steels, and their results were compared with that of conventional austenitic stainless steel and ferritic-martensitic stainless steel. Following results were obtained in this study. (1) 9Cr-ODS martensitic and 12Cr-ODS ferritic steel have superior high temperature oxidation resistance compared to 11mass%Cr PNC-FMS and even 17mass% SUS430 and equivalent to austenitic PNC316. (2) The superior oxidation resistance of ODS steel was attributed to earlier formation of the protective alpha-Cr 2 O 3 layer at the matrix and inner oxide scale interface. The grain size of ODS steel is finer than that of PNC-FMS, so the superior oxidation resistance of ODS steel can be attributed to the enhanced Cr-supplying rate throughout the accelerated grain boundary diffusion. Finely dispersed Y 2 O 3 oxide particles in the ODS steel matrix may also stabilized the adherence between the protective alpha-Cr 2 O 3 layer and the matrix. (author)

Performance of advanced oxide fuel pins in EBR-II

International Nuclear Information System (INIS)

Lawrence, L.A.; Jensen, S.M.; Hales, J.W.; Karnesky, R.A.; Makenas, B.J.

1986-05-01

The effects of design and operating parameters on mixed-oxide fuel pin irradiation performance were established for the Hanford Engineering Development Laboratory (HEDL) advanced oxide EBR-II test series. Fourteen fuel pins breached in-reactor with reference 316 SS cladding. Seven of the breaches are attributed to FCMI. Of the remaining seven breached pins, three are attributed to local cladding over-temperatures similar to the breach mechanism for the reference oxide pins irradiated in EBR-II. FCCI was found to be a contributing factor in two high burnup, i.e., 11.7 at. % breaches. The remaining two breaches were attributed to mechanical interaction of UO 2 fuel and fission products accumulated in the lower cladding insulator gap, and a loss of cladding ductility possibly due to liquid metal embrittlement. Fuel smear density appears to have the most significant impact on lifetime. Quantitative evaluations of cladding diameter increases attributed to FCMI, established fuel smear density, burnup, and cladding thickness-to-diameter ratio as the major parameters influencing the extent of cladding strain

Aqueous dispersion of monodisperse magnetic iron oxide nanocrystals through phase transfer

International Nuclear Information System (INIS)

Yu, William W; Chang, Emmanuel; Sayes, Christie M; Drezek, Rebekah; Colvin, Vicki L

2006-01-01

A facile method was developed for completely transferring high quality monodisperse iron oxide nanocrystals from organic solvents to water. The as-prepared aqueous dispersions of iron oxide nanocrystals were extremely stable and could be functionalized for bioconjugation with biomolecules. These iron oxide nanocrystals showed negligible cytotoxicity to human breast cancer cells (SK-BR-3) and human dermal fibroblast cells. This method is general and versatile for many organic solvent-synthesized nanoparticles, including fluorescent semiconductor nanocrystals

Breaking up of pure and simulated 'burnt' mixed oxide fuel by chemical interaction with oxidized sodium

International Nuclear Information System (INIS)

Besnard, R.; Chaudat, J.P.

1983-01-01

A large experimental program have permitted to investigate the behaviour of mixed oxide fuel coming in contact with hot oxidized sodium. The kinetic of the reaction, the size and the chemical nature of the particules after interaction have been studied. The main part of experiments have been performed using mixed oxide fuel non irradiated at first and with simulated fission products afterwards. Complementary informations have been obtained with UO 2 fuel pellets. After description of the experimental devices, the results are discussed and the importance of the main parameters, like temperature and fission products effect, are pointed out. (orig.)

Interconnection of bundled solid oxide fuel cells

Science.gov (United States)

Brown, Michael; Bessette, II, Norman F; Litka, Anthony F; Schmidt, Douglas S

2014-01-14

A system and method for electrically interconnecting a plurality of fuel cells to provide dense packing of the fuel cells. Each one of the plurality of fuel cells has a plurality of discrete electrical connection points along an outer surface. Electrical connections are made directly between the discrete electrical connection points of adjacent fuel cells so that the fuel cells can be packed more densely. Fuel cells have at least one outer electrode and at least one discrete interconnection to an inner electrode, wherein the outer electrode is one of a cathode and and anode and wherein the inner electrode is the other of the cathode and the anode. In tubular solid oxide fuel cells the discrete electrical connection points are spaced along the length of the fuel cell.

Electrocatalytic oxidation of methanol: study with Pt:Mo dispersed catalysts

Directory of Open Access Journals (Sweden)

Oliveira Neto Almir

2000-01-01

Full Text Available The electrocatalytic oxidation of methanol on Pt:Mo dispersed on carbon prepared using an alternative method recently developed in this laboratory was investigated. The EDX analysis confirmed that the simultaneous reduction of the precursor salts of Pt and Mo leads to the presence of these materials at the nominal composition initially calculated. The addition of Mo to Pt causes an increase of the oxidation currents, but does not improve the catalytic effect for methanol oxidation. Tafel plots for various methanol concentrations showed the presence of two slopes. On line differential electrochemical mass spectrometry (DEMS was used to investigate the distribution of products and intermediates in methanol oxidation.

Ultrahighly Dispersed Titanium Oxide on Silica : Effect of Precursors on the Structure and Photocatalysis

OpenAIRE

Yoshida , S.; Takenaka , S.; Tanaka , T.; Funabiki , T.

1997-01-01

The effect of precursor on the dispersion and catalytic performance of titanium oxide supported on silica has ben investigated. The catalysts were prepared by a simple impregnation method with three kinds of titanium complexes of different ligands (bis(isopropyato)-bis(pivaroylmethanato) : DPM, acetylacetonato : ACAC, tetrakis(isopropylato) : IPRO) with the aim of preparing ultrahighly dispersed titanium oxide on silica. The XAFS study revealed that titanium species in the catalyst prepared f...

Pt nanoparticle-reduced graphene oxide nanohybrid for proton exchange membrane fuel cells.

Science.gov (United States)

Park, Dae-Hwan; Jeon, Yukwon; Ok, Jinhee; Park, Jooil; Yoon, Seong-Ho; Choy, Jin-Ho; Shul, Yong-Gun

2012-07-01

A platinum nanoparticle-reduced graphene oxide (Pt-RGO) nanohybrid for proton exchange membrane fuel cell (PEMFC) application was successfully prepared. The Pt nanoparticles (Pt NPs) were deposited onto chemically converted graphene nanosheets via ethylene glycol (EG) reduction. According to the powder X-ray diffraction (XRD) pattern and transmission electron microscopy (TEM) analysis, the face-centered cubic Pt NPs (3-5 nm in diameter) were homogeneously dispersed on the RGO nanosheets. The electrochemically active surface area and PEMFC power density of the Pt-RGO nanohybrid were determined to be 33.26 m2/g and 480 mW/cm2 (maximum values), respectively, at 75 degrees C and at a relative humidity (RH) of 100% in a single-cell test experiment.

Application of laser ablation inductivly coupled plasma mass spectrometry for characterization of U-7Mo/Al-55i dispersion fuels

Energy Technology Data Exchange (ETDEWEB)

Lee, Jeong Mook; Park, Jai Il; Youn, Young Sang; Ha, Yeong Keong; Kim, Jong Yun [Nuclear Chemistry Research Division, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2017-04-15

This technical note demonstrates the feasibility of using laser ablation inductively coupled plasma mass spectrometry for the characterization of U–7Mo/Al–5Si dispersion fuel. Our measurements show 5.0% Relative Standard Deviation (RSD) for the reproducibility of measured {sup 98}Mo/{sup 238}U ratios in fuel particles from spot analysis, and 3.4% RSD for {sup 98}Mo/{sup 238}U ratios in a NIST-SRM 612 glass standard. Line scanning allows for the distinction of U–7Mo fuel particles from the Al–5Si matrix. Each mass spectrum peak indicates the presence of U–7Mo fuel particles, and the time width of each peak corresponds to the size of that fuel particle. The size of the fuel particles is estimated from the time width of the mass spectrum peak for {sup 98}Mo by considering the scan rate used during the line scan. This preliminary application clearly demonstrates that laser ablation inductively coupled plasma mass spectrometry can directly identify isotope ratios and sizes of the fuel particles in U–Mo/Al dispersion fuel. Once optimized further, this instrument will be a powerful tool for investigating irradiated dispersion fuels in terms of fission product distributions in fuel matrices, and the changes in fuel particle size or shape after irradiation.

Postirradiation examination of high-density uranium alloy dispersion fuels

International Nuclear Information System (INIS)

Hayes, S.L.; Meyer, M.K.; Hofman, G.L.; Strain, R.V.

1998-01-01

Two irradiation test vehicles, designated RERTR-2, were inserted into the Advanced Test reactor in Idaho in August 1997. These tests were designed to obtain irradiation performance information on a variety of potential new, high-density uranium alloy dispersion fuels, including U-10Mo, U-8Mo, U-6Mo, U-4Mo, U-9Nb-3Zr, U-6Nb-4Zr, U-5Nb-3Zr, U-6Mo-1Pt, U-6Mo-0.6Ru and U-10Mo-0.05Sn: the intermetallic compounds U 2 Mo and U-10Mo-0.-5Sn; the intermetallic compounds U 2 Mo and U 3 Si 2 were also included in the fuel test matrix. These fuels are included in the experiments as microplates (76 mm x 22 mm x 1.3mm outer dimensions) with a nominal fuel volume loading of 25% and irradiated at relatively low temperature (∼100 deg C). RERTR-1 and RERTR-2 were discharged from the reactor in November 1997 and July 1998, respectively at calculated peak fuel burnups of 45 and 71 at %-U 235 Both experiments are currently under examination at the Alpha Gamma Hot Cell Facility at Argonne National Laboratory in Chicago. This paper presents the postirradiation examination results available to date from these experiments. (author)

Fabrication of CNT Dispersion Fluid by Wet-Jet Milling Method for Coating on Bipolar Plate of Fuel Cell

Directory of Open Access Journals (Sweden)

Anas Almowarai

2015-01-01

Full Text Available Water based carbon nanotube (CNT dispersion was produced by wet-jet milling method. Commercial CNT was originally agglomerated at the particle size of less than 1 mm. The wet-jet milling process exfoliated CNTs from the agglomerates and dispersed them into water. Sedimentation of the CNTs in the dispersion fluid was not observed for more than a month. The produced CNT dispersion was characterized by the SEM and the viscometer. CNT/PTFE composite film was formed with the CNT dispersion in this study. The electrical conductivity of the composite film increased to 10 times when the CNT dispersion, which was produced by the wet-jet milling method, was used as a constituent of the film. Moreover, the composite film was applied to bipolar plate of fuel cell and increased the output power of the fuel cell to 1.3 times.

High temperature transient deformation of mixed oxide fuels

International Nuclear Information System (INIS)

Slagle, O.D.

1986-01-01

The purpose of this paper is to present recent experimental results on fuel creep under transient conditions at high temperatures. The effect of temperature, stress, heating rate, density and grain size were considered. An empirical formulation is derived for the relationship between strain, stress, temperature and heating rate. This relationship provides a means for incorporating stress relief into the analysis of fuel-cladding interaction during an overpower transient. The effect of sample density and initial grain size is considered by varying the sample parameters. Previously derived steady-state creep relationships for the high temperature creep of mixed oxide fuel were combined with the time dependency of creep found for UO 2 to calculate a transient creep relationship for mixed oxide fuel. These calculated results were found to be in good agreement with the measured high temperature transient creep results

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

Energy Technology Data Exchange (ETDEWEB)

Boltax, A [Westinghouse Electric Corporation, Advanced Reactor Division, Madison, PA (United States); Biancheria, A

1977-04-01

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

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

International Nuclear Information System (INIS)

Boltax, A.; Biancheria, A.

1977-01-01

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

Pilot-scale equipment development for pyrochemical treatment of spent oxide fuel

International Nuclear Information System (INIS)

Herrmann, S. D.

1999-01-01

Fundamental objectives regarding spent nuclear fuel treatment technologies include, first, the effective distribution of spent fuel constituents among product and stable waste forms and, second, the minimization and standardization of waste form types and volumes. Argonne National Laboratory (ANL) has developed and is presently demonstrating the electrometallurgical treatment of sodium-bonded metal fuel from Experimental Breeder Reactor II, resulting in an uranium product and two stable waste forms, i.e. ceramic and metallic. Engineering efforts are underway at ANL to develop pilot-scale equipment which would precondition irradiated oxide fuel via pyrochemical processing and subsequently allow for electrometallurgical treatment of such non-metallic fuels into standard product and waste forms. This paper highlights the integration of proposed spent oxide fuel treatment with existing electrometallurgical processes. System designs and technical bases for development of pilot-scale oxide reduction equipment are also described

Transmutation of minor actinide using BWR fueled mixed oxide

International Nuclear Information System (INIS)

Susilo, Jati

2000-01-01

Nuclear spent fuel recycle has a strategic importance in the aspect of nuclear fuel economy and prevention of its spread-out. One among other application of recycle is to produce mixed oxide fuel (Mo) namely mixed Plutonium and uranium oxide. As for decreasing the burden of nuclear high level waste (HLW) treatment, transmutation of minor actinide (MA) that has very long half life will be carried out by conversion technique in nuclear reactor. The purpose of this study was to know influence of transition fuel cell regarding the percent weight of transmutation MA in the BWR fueled MOX. Calculation of cell BWR was used SRAC computer code, with assume that the reactor in equilibrium. The percent weight of transmutation MA to be optimum by increasing the discharge burn-up of nuclear fuel, raising ratio of moderator to fuel volume (Vm/Vf), and loading MA with percent weight about 3%-6% and also reducing amount of percent weight Pu in MOX fuel. For mixed fuel standard reactor, reactivity value were obtained between about -50pcm ∼ -230pcm for void coefficient and -1.8pcm ∼ -2.6pcm for fuel temperature coefficient

Trends for Methane Oxidation at Solid Oxide Fuel Cell Conditions

DEFF Research Database (Denmark)

Kleis, Jesper; Jones, Glenn; Abild-Pedersen, Frank

2009-01-01

First-principles calculations are used to predict a plausible reaction pathway for the methane oxidation reaction. In turn, this pathway is used to obtain trends in methane oxidation activity at solid oxide fuel cell (SOFC) anode materials. Reaction energetics and barriers for the elementary...... the Ni surfaces to other metals of interest. This allows the reactivity over the different metals to be understood in terms of two reactivity descriptors, namely, the carbon and oxygen adsorption energies. By combining a simple free-energy analysis with microkinetic modeling, activity landscapes of anode...

Fabrication of cermet fuel for fast reactor

International Nuclear Information System (INIS)

Mishra, Sudhir; Kumar, Arun; Kutty, T.R.G.; Kamath, H.S.

2011-01-01

Mixed oxide (MOX) (U,Pu)O 2 , and metallic (U,Pu ,Zr) fuels are considered promising fuels for the fast reactor. The fuel cycle of MOX is well established. The advantages of the oxide fuel are its easy fabricability, good performance in the reactor and a well established reprocessing technology. However the problems lie in low thermal conductivity , low density of the fuel leading to low breeding ratio and consequently longer doubling time. The metallic fuel has the advantages of high thermal conductivity, higher metal density and higher coefficient of linear expansion. The higher coefficient of linear expansion is good from the safety consideration (negative reactivity factor). Because of higher metal density it offers highest breeding ratio and shortest doubling time. Metallic fuel disadvantages comprise large swelling at high burnup, fuel cladding interaction and lower margin between operating and melting temperature. The optimal solution may lie in cermet fuel (U, PuO 2 ), where PuO 2 is dispersed in U metal matrix and combines the favorable features of both the fuel types. The advantages of this fuel include high thermal conductivity, larger margin between melting and operating temperature, ability to retain fission product etc. The matrix being of high density metal the advantage of high breeding ratio is also maintained. In this report some results of fabrication of cermet pellet comprising of UO 2 /PuO 2 dispersed in U metal powder through classical powder metallurgy route and characterization are presented. (author)

Reirradiation of mixed-oxide fuel pins at increased temperatures

International Nuclear Information System (INIS)

Lawrence, L.A.; Weber, E.T.

1976-05-01

Mixed-oxide fuel pins from EBR-II irradiations were reirradiated in the General Electric Test Reactor (GETR) at higher temperatures than experienced in EBR-II to study effects of the increased operating temperatures on thermal/mechanical and chemical behavior. The response of a mixed-oxide fuel pin to a power increase after having operated at a lower power for a significant portion of its life-time is an area of performance evaluation where little information currently exists. Results show that the cladding diameter changes resulting from the reirradiation are strongly dependent upon both prior burnup level and the magnitude of the temperature increase. Results provide the initial rough outlines of boundaries within which mixed-oxide fuel pins can or cannot tolerate power increases after substantial prior burnup at lower powers

Plasticity of oxide dispersion strengthened ferritic alloys; Plasticite des alliages ferritiques renforces par dispersion d`oxydes

Energy Technology Data Exchange (ETDEWEB)

Zakine, C

1994-07-05

The object of this work is to study the plasticity mechanisms of two oxide dispersion strengthened ferritic alloys, DT and DY. Microstructural characterisation has been performed on DT and DY alloys by optical, scanning and transmission electron microscopy. These materials, strengthened by an oxide dispersion, contain an intermetallic {chi} phase precipitated on grain boundaries. The {chi} phase, stable up to 900 deg, can be dissolved into the matrix by heat treatment beyond 1 000 deg. Between 20 and 700 deg, according to tensile tests, the DY alloy which is strengthened by a fine dispersion of yttria particles is more resistant and less ductile than DT alloy, strengthened by titanium oxides. Tensile tests performed at room temperature, in the chamber of a SEM, have shown that micro-cracking of the {chi} phase coincides with the first stage of the macroscopic yielding. The cavities initiated by the {chi} phase micro-cracking induce a ductile fracture of the matrix. A dynamic strain ageing mechanism has been observed around 400 deg, which is attributed to the Mo contribution. Between 20 and 700 deg, comparison of tensile properties of alloys with or without {chi} phase has shown that the intermetallic phase has a detrimental effect on the ductility, but has no influence on the mechanical strength. Creep tests have been performed between 500 and 700 deg. Thermally activated plasticity mechanisms are observed in this temperature range. The {chi} phase, which is always micro-cracked after tensile testing, is not damaged after creep testing below a critical stress. This behaviour is explained by the influence of strain rate through the competition between strain hardening and relaxation of the matrix. (author).

A structure study of copper oxide for monolayer dispersion of anatase supported

International Nuclear Information System (INIS)

Zi Fenlan; Yu Xiaofeng; Guo Hongyou; Cai Xiaohai; Yang Pengcheng; Wu Nianzu; Xie Yaning; Zang Jing; Hu Tiandou

2002-01-01

The monolayer dispersion of copper oxide on the surface of anatase and its effect on the properties have been studied by X-ray photoelectron spectroscopy (XPS) and X-ray extended absorption fine structure (EXAFS). XPS results give an utmost dispersion capacity of 7.2 mg/gTiO 2 . Strong interactions between copper oxide and anatase can be seen from EXAFS results. The structure of the supported CuO species is strongly dependent on the amount of CuO loading. When the content of CuO loading is below the utmost dispersion capacity, the surface of CuO/TiO 2 is dominated by the highly dispersed CuO species having no -Cu-O-Cu- chains. The copper ion is located in an octahedral coordination environment, and the Cu-O coordination distance is much longer than that in pure crystalline CuO. When CuO loading is exceeds the utmost dispersion capacity, crystalline CuO is formed on the surface of CuO/TiO 2 . From the result of the structure study, it is Cu-O octahedral coordination and coordination distance change in comparison with pure crystalline CuO on the surface CuO/TiO 2 that have catalytic activity

Stability under irradiation of a fine dispersion of oxides in a ferritic matrix

International Nuclear Information System (INIS)

Monnet, I.

1999-01-01

Oxide dispersion strengthened (ODS) ferritic-martensitic steels are being considered for high temperature, high fluence nuclear applications, like fuel pin cladding in Fast Breeder Reactors. ODS alloys offer improved out of pile strength characteristics at temperature above 550 deg.C and ferritic-martensitic matrix is highly swelling resistant. A clad in an ODS ferritic steel, call DY (Fe-13Cr-1,5Mo+TiO 2 +Y 2 O 3 ) has been irradiated in the experimental reactor Phenix. Under irradiation oxide dissolution occurs. Microstructural observations indicated that oxide evolution is correlated with the dose and consist in four phenomena: the interfaces of oxide particles with the matrix become irregular, the uniform distribution of the finest oxide ( 2 O 3 , Y 2 O 3 , MgO or MgAl 2 O 4 . These materials were irradiated with charged particles in order to gain a better understanding of the mechanisms of dissolution. Irradiation with 1 MeV Helium does not induce any modification, neither in the chemical modification of the particles nor in their spatial and size distribution. Since most of the energy of helium ions is lost by inelastic interaction, this result proves that this kind of interaction does not induce oxide dissolution. Irradiation with 1 MeV or 1.2 MeV electrons leads to a significant dissolution with a radius decrease proportional to the dose. These experiments prove that oxide dissolution can be induced by Frenkel pairs alone, provided that metallic atoms are displaced. The comparison between irradiation with ions (displacements cascades) and electrons (Frenkel pairs only) shows the importance of free point defects in the dissolution phenomena. For all the irradiations (ions or electrons) the spinel MgAl 2 O 4 seems more resistant than Y 2 O 3 to dissolution, and MgO and Al 2 O 3 are even less resistant. This is the order of stability under irradiation of bulk oxides. (author)

Analyses of Interaction Phases of U Mo Dispersion Fuel by Synchrotron X ray Diffraction

Energy Technology Data Exchange (ETDEWEB)

Kim, Woo Jeong; Nam, Ji Min; Ryu, Ho Jin; Park, Jong Man [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Herve, Palancher; Charollais, Francois [Saint Paul Lez Durance Cedex, Rhone (France); Bonnin, Anne; Honkimaeki, Veijo [Grenoble Cedex, Grenoble (France); Patrick Lemoined [Gif sur Yvette, Paris (France)

2012-10-15

Gamma phase U Mo alloys are one of the promising candidates to be used as advanced high uranium density fuel for high power research reactors due to their excellent irradiation performance. However, formation of interaction layers between the U Mo particles and Al matrix degrades the irradiation performance of U Mo dispersion fuel. One of the remedies to the interaction problem is a Si addition to the Al matrix. Recent irradiation tests have shown that the use of Al (2{approx}5wt%)Si matrices retarded the growth of interaction layers effectively during irradiation. Recently, KAERI has proposed silicide or nitride coated U Mo fuel for the minimization of the interaction layer growth. The silicide or nitride coatings are expected to act as interdiffusion barriers and their out of pile tests showed the improved diffusion barrier performances of the silicide and nitride layers. In order to characterize constituent phases in the coated layers on U Mo particles and the interaction layers of coated U Mo particle dispersed fuel, synchrotron X ray diffraction experiments have been performed at the ESRF (European Synchrotron Radiation Facility), France as a KAERI CEA cooperation program.

LG Solid Oxide Fuel Cell (SOFC) Model Development

Energy Technology Data Exchange (ETDEWEB)

Haberman, Ben [LG Fuel Cell Systems Inc., North Canton, OH (United States); Martinez-Baca, Carlos [LG Fuel Cell Systems Inc., North Canton, OH (United States); Rush, Greg [LG Fuel Cell Systems Inc., North Canton, OH (United States)

2013-05-31

This report presents a summary of the work performed by LG Fuel Cell Systems Inc. during the project LG Solid Oxide Fuel Cell (SOFC) Model Development (DOE Award Number: DE-FE0000773) which commenced on October 1, 2009 and was completed on March 31, 2013. The aim of this project is for LG Fuel Cell Systems Inc. (formerly known as Rolls-Royce Fuel Cell Systems (US) Inc.) (LGFCS) to develop a multi-physics solid oxide fuel cell (SOFC) computer code (MPC) for performance calculations of the LGFCS fuel cell structure to support fuel cell product design and development. A summary of the initial stages of the project is provided which describes the MPC requirements that were developed and the selection of a candidate code, STAR-CCM+ (CD-adapco). This is followed by a detailed description of the subsequent work program including code enhancement and model verification and validation activities. Details of the code enhancements that were implemented to facilitate MPC SOFC simulations are provided along with a description of the models that were built using the MPC and validated against experimental data. The modeling work described in this report represents a level of calculation detail that has not been previously available within LGFCS.

Oxidative dissolution of ADOPT compared to standard UO{sub 2} fuel

Energy Technology Data Exchange (ETDEWEB)

Nilsson, Kristina [School of Chemical Science and Engineering, Applied Physical Chemistry, KTH Royal Institute of Technology, SE-100 44 Stockholm (Sweden); Roth, Olivia [Studsvik Nuclear AB, SE-611 82 Nyköping (Sweden); Jonsson, Mats, E-mail: matsj@kth.se [School of Chemical Science and Engineering, Applied Physical Chemistry, KTH Royal Institute of Technology, SE-100 44 Stockholm (Sweden)

2017-05-15

In this work we have studied oxidative dissolution of pure UO{sub 2} and ADOPT (UO{sub 2} doped with Al and Cr) pellets using H{sub 2}O{sub 2} and gammaradiolysis to induce the process. There is a small but significant difference in the oxidative dissolution rate of UO{sub 2} and ADOPT pellets, respectively. However, the difference in oxidative dissolution yield is insignificant. Leaching experiments were also performed on in-reactor irradiated ADOPT and UO{sub 2} pellets under oxidizing conditions. The results indicate that the U(VI) release is slightly slower from the ADOPT pellet compared to the UO{sub 2.} This could be attributed to differences in exposed surface area. However, fission products with low UO{sub 2} solubility display a higher relative release from ADOPT fuel compared to standard UO{sub 2}-fuel. This is attributed to a lower matrix solubility imposed by the dopants in ADOPT fuel. The release of Cs is higher from UO{sub 2} which is attributed to the larger grain size of ADOPT. - Highlights: •Oxidative dissolution of ADOPT fuel is compared to standard UO{sub 2} fuel. •Only marginal differences are observed. •The main difference observed is in the relative release rate of fission products. •Differences are claimed to be attributed to a lower matrix solubility imposed by the dopants in ADOPT fuel.

Comparison of thermal compatibility between atomized and comminuted U{sub 3}Si dispersion fuels

Energy Technology Data Exchange (ETDEWEB)

Ryu, Woo-Seog; Park, Jong-Man; Kim, Chang-Kyu; Kuk, II-Hyun [Korea Atomic Research Institute, Taejon (Korea, Republic of)

1997-08-01

Thermal compatibility of atomized U{sub 3}Si dispersion fuels were evaluated up to 2600 hours in the temperature range from 250 to 500{degrees}C, and compared with that of comminuted U{sub 3}Si. Atomized U{sub 3}Si showed better performance in terms of volume expansion of fuel meats. The reaction zone of U{sub 3}Si and Al occurred along the grain boundaries and deformation bands in U{sub 3}Si particles. Pores around fuel particles appeared at high temperature or after long-term annealing tests to remain diffusion paths over the trench of the pores. The constraint effects of cladding on fuel rod suppressed the fuel meat, and reduced the volume expansion.

Industrial use of SPring-8 in fuel cell development

International Nuclear Information System (INIS)

Sugiura, Masahiro

2007-01-01

The study of fuel cells by using synchrotron radiation from SPring-8 was reviewed for polymer electrolyte fuel cells (PEFCs; also called proton exchange membrane fuel cells), solid oxide fuel cells (SOFCs), and fuel cell-related materials. PEFCs use a solid polymer as an electrolyte and porous carbon electrodes containing a platinum catalyst. Measurements of Pt particles in the cathode catalyst by x-ray absorption near-edge structure (XANES) showed that the decrease of the particle size of Pt particles caused an increased of the number of 5d orbital holes of Pt atoms. Oxidization processes of Pt particles were measured by time-resolved dispersive x-ray absorption fine structure (XAFS). Measurements by time-gating quick scan XAFS together with dispersive XAFS revealed the reduction-oxidation process of Pt nanoparticles under the operation condition of PEFCs. SOFCs use a hard, non-porous ceramic compound as the electrolyte. SOFCs are operated at very high temperatures - around 1,000degC. Oxides having perovskite and fluorite structures are one of the most promising materials for electrolyte of SOFCs operated at reduced temperatures. The local structures of doped ceria and lanthanum gallate compounds were studied by extended XAFS. It was indicated from the measurements of ionic conductivity of these compounds that the local structure seriously affected oxide ionic conduction. Residual thermal stresses in the electrolyte of the anode-supported planar SOFCs were measured by high-energy x-rays. The crystal structure and electron density distribution of hydrogen storage alloys were measured by x-ray diffraction. (Y.K.)

Pilot-scale equipment development for lithium-based reduction of spent oxide fuel

International Nuclear Information System (INIS)

Herrmann, S. D.

1998-01-01

An integral function of the electrometallurgical conditioning of DOE spent nuclear fuel is the standardization of waste forms. Argonne National Laboratory (ANL) has developed and is presently demonstrating the electrometallurgical conditioning of sodium-bonded metal fuel from Experimental Breeder Reactor II, resulting in uranium, ceramic waste, and metal waste forms. Engineering studies are underway at ANL in support of pilot-scale equipment development, which would precondition irradiated oxide fuel and likewise demonstrate the application of electrometallurgical conditioning to such non-metallic fuels. This paper highlights the integration of proposed spent oxide fuel conditioning with existing electrometallurgical processes. Additionally, technical bases for engineering activities to support a scale up of an oxide reduction process are described

DART-TM: A thermomechanical version of DART for LEU VHD dispersed and monolithic fuel analysis

International Nuclear Information System (INIS)

Saliba, Roberto; Taboada, Horacio; Moscarda, Ma.Virginia; Rest, Jeff

2003-01-01

A collaboration agreement between ANL/USDOE and CNEA Argentina, in the area of Low Enriched Uranium Advanced Fuels has been in place since October 16, 1997 under the 'Implementation Arrangement for Technical Exchange and Cooperation in the Area of Peaceful Uses of Nuclear Energy'. An annex concerning DART code optimization has been operative since February 8, 1999. Previously, as a part of this annex a visual thermal FASTDART version was developed that includes mechanistic models for the calculation of the fission-gas-bubble and fuel particle size distribution, reaction layer thickness, and meat thermal conductivity. FASTDART was presented at the last RERTR Meeting that included validation against RERTR 3 irradiation data. The thermal FASTDART version was assessed as an adequate tool for modeling the behavior of LEU U-Mo dispersed fuels under irradiation against PIE RERTR irradiation data. During this past year the development of a 3-D thermo-mechanical version of the code for modeling the irradiation behavior of LEU U-Mo monolithic and dispersion fuel was initiated. Some preliminary results of this work will be shown during RERTR-2003 meeting. (author)

Dry air oxidation kinetics of K-Basin spent nuclear fuel

International Nuclear Information System (INIS)

Abrefah, J.; Buchanan, H.C.; Gerry, W.M.; Gray, W.J.; Marschman, S.C.

1998-06-01

The safety and process analyses of the proposed Integrated Process Strategy (IPS) to move the N-Reactor spent nuclear fuel (SNF) stored at K-Basin to an interim storage facility require information about the oxidation behavior of the metallic uranium. Limited experiments have been performed on the oxidation reaction of SNF samples taken from an N-Reactor outer fuel element in various atmospheres. This report discusses studies on the oxidation behavior of SNF using two independent experimental systems: (1) a tube furnace with a flowing gas mixture of 2% oxygen/98% argon; and (2) a thermogravimetric system for dry air oxidation

Progress in irradiation performance of experimental uranium - Molybdenum dispersion fuel

International Nuclear Information System (INIS)

Hofman, Gerard L.; Meyer, Mitchell K.

2002-01-01

High-density dispersion fuel experiment, RERTR-4, was removed from the Advanced Test Reactor (ATR) after reaching a peak U-235 burnup of ∼80% and is presently undergoing postirradiation examination at the ANL alpha-gamma hot cells. This test consists of 32 mini fuel plates of which 27 were fabricated with nominally 6 and 8 g cm -3 atomized and machined uranium alloy powders containing 7 wt% and 10 wt% molybdenum. In addition, two miniplates containing solid U-10 wt% Mo foils and three containing 6 g cm -3 U 3 Si 2 are part of the test. The results of the postirradiation examination and analysis of RERTR-4 in conjunction with data from previous tests performed to lower burnup will be presented. (author)

Steam and partial oxidation reforming options for hydrogen production from fossil fuels for PEM fuel cells

Directory of Open Access Journals (Sweden)

Yousri M.A. Welaya

2012-06-01

Full Text Available Proton exchange membrane fuel cell (PEM generates electrical power from air and from hydrogen or hydrogen rich gas mixtures. Therefore, there is an increasing interest in converting current hydrocarbon based marine fuels such as natural gas, gasoline, and diesel into hydrogen rich gases acceptable to the PEM fuel cells on board ships. Using chemical flow sheeting software, the total system efficiency has been calculated. Natural gas appears to be the best fuel for hydrogen rich gas production due to its favorable composition of lower molecular weight compounds. This paper presents a study for a 250 kW net electrical power PEM fuel cell system utilizing a partial oxidation in one case study and steam reformers in the second. This study has shown that steam-reforming process is the most competitive fuel processing option in terms of fuel processing efficiency. Partial oxidation process has proved to posses the lowest fuel processing efficiency. Among the options studied, the highest fuel processing efficiency is achieved with natural gas steam reforming system.

Fracture toughness of oxide-dispersion strengthened copper

Energy Technology Data Exchange (ETDEWEB)

Alexander, D.J. [Oak Ridge National Lab., TN (United States)

1996-10-01

The fracture toughness of an oxide-dispersion strengthened copper alloy AL-15 has been examined at room temperature and 250{degrees}C, in air and in vacuum (< 10{sup {minus}6} torr). Increasing test temperature causes a significant decrease in the fracture toughness of this material, in either air or vacuum environments. In addition, specimens oriented in the T-L orientation (crack growth parallel to the extrusion direction) show significantly lower toughness than those in the L-T orientation (crack growth perpendicular to the extrusion direction).

Operando Raman Micro Spectroscopy of Polymer Electrolyte Fuel Cells

Science.gov (United States)

2016-01-16

the cathode , transitions ion exchange sites from the sulfonic acid to the dissociated sulfonate form. Visualization of density functional theory...catalysts dispersed in an alcoholic dispersion of solubilized ionomer (e.g., Nafion). Teflon dispersion is included in cathode inks to lower the surface...tolerant of condensed water, is complementary to FTIR. Operando Raman spectroscopy of solid oxide fuel cells has been reported.28–30 Although there are

Durable and Robust Solid Oxide Fuel Cells

DEFF Research Database (Denmark)

Hjalmarsson, Per; Knibbe, Ruth; Hauch, Anne

project had as one of its’ overarching goals to improve durability and robustness of the Danish solid oxide fuel cells. The project focus was on cells and cell components suitable for SOFC operation in the temperature range 600 – 750 °C. The cells developed and/or studied in this project are intended......The solid oxide fuel cell (SOFC) is an attractive technology for the generation of electricity with high efficiency and low emissions. Risø DTU (now DTU Energy Conversion) works closely together with Topsoe Fuel Cell A/S in their effort to bring competitive SOFC systems to the market. This 2-year...... for use within the CHP (Combined Heat and Power) market segment with stationary power plants in the range 1 – 250 kWe in mind. Lowered operation temperature is considered a good way to improve the stack durability since corrosion of the interconnect plates in a stack is lifetime limiting at T > 750 °C...

Fabrication of simulated DUPIC fuel

Science.gov (United States)

Kang, Kweon Ho; Song, Ki Chan; Park, Hee Sung; Moon, Je Sun; Yang, Myung Seung

2000-12-01

Simulated DUPIC fuel provides a convenient way to investigate the DUPIC fuel properties and behavior such as thermal conductivity, thermal expansion, fission gas release, leaching, and so on without the complications of handling radioactive materials. Several pellets simulating the composition and microstructure of DUPIC fuel are fabricated by resintering the powder, which was treated through OREOX process of simulated spent PWR fuel pellets, which had been prepared from a mixture of UO2 and stable forms of constituent nuclides. The key issues for producing simulated pellets that replicate the phases and microstructure of irradiated fuel are to achieve a submicrometre dispersion during mixing and diffusional homogeneity during sintering. This study describes the powder treatment, OREOX, compaction and sintering to fabricate simulated DUPIC fuel using the simulated spent PWR fuel. The homogeneity of additives in the powder was observed after attrition milling. The microstructure of the simulated spent PWR fuel agrees well with the other studies. The leading structural features observed are as follows: rare earth and other oxides dissolved in the UO2 matrix, small metallic precipitates distributed throughout the matrix, and a perovskite phase finely dispersed on grain boundaries.

Elaboration and characterisation of functionally graded cathodes for solid oxide fuel cells

Energy Technology Data Exchange (ETDEWEB)

Simonet, J.; Kapelski, G.; Bouvard, D. [Laboratoire de Genie Physique et Mecanique des Materiaux, Institut National Polytechnique de Grenoble, CNRS UMR 5010, BP 46, 38042 Saint Martin d' Heres cedex (France)

2005-07-01

The industrial development of solid oxide fuel cells (SOFC) requires decreasing their operating temperature from 1000 deg. C to 700 deg. C while keeping acceptable mechanical and electrochemical performances. A solution consists in designing composite bulk cathodes with numerous electro-chemical reaction sites. The fabrication of such cathodes has been investigated with classical materials as lanthanum strontium manganese (LSM) and yttrium stabilized zirconia (YSZ), which is also the constitutive material of the electrolyte. A composite cathode with continuous composition gradient has been obtained by co-sedimentation of the powders in a liquid and subsequent firing. The obtained composition is investigated with Scanning Electron Microscope (SEM) and Electron Dispersive Spectrometry (EDS). It is found to be in good agreement with the prediction of a numerical model of the sedimentation process. (authors)

Performance Evaluation of Solid Oxide Fuel Cell by Computer ...

African Journals Online (AJOL)

The search for sustainable energy source that can compete with the existing one led to the discovery and acceptance of fuel cell technologies as a perfect replacement for fossil fuel. The ability of Solid Oxide Fuel Cells (SOFC) to capture the heat generation during the process of energy generation from electrochemical ...

Experimental and thermodynamic evaluation of the melting behavior of irradiated oxide fuels

International Nuclear Information System (INIS)

Adamson, M.G.; Aitken, E.A.; Caputi, R.W.

1985-01-01

Onset of melting is an important performance limit for irradiated UO 2 and UO 2 -based nuclear reactor fuels. Melting (solidus) temperatures are reasonably well known for starting fuel materials such as UO 2 and (U,PU)O 2 , however the influence of burnup on oxide fuel melting behavior continues to represent an area of considerable uncertainty. In this paper we report the results of a variety of melting temperature measurements on pseudo-binary fuel-fissia mixtures such as UO 2 -PUO 2 , UO 2 -CeO 2 , UO 2 -BaO, UO 2 -SrO, UO 2 -BaZrO 3 and UO 2 -SrZrO 3 . These measurements were performed using the thermal arrest technique on tungsten-encapsulated specimens. Several low melting eutectics, the existence of which had previously been inferred from post-irradiation examinations of high burnup mixed oxide fuels, were characterized in the course of the investigation. Also, an assessment of melting temperature changes in irradiated oxide fuels due to the production and incorporation of soluble oxidic fission products was performed by application of solution theory to the available pseudo-binary phase diagram data. The results of this assessment suggest that depression of oxide fuel solidus temperatures by dissolved fission products is substantially less than that indicated by earlier experimental studies. (orig.)

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.

Evaluation of the oxide and silicide fuels reactivity in the RSG-GAS core

International Nuclear Information System (INIS)

S, Tukiran; M S, Tagor; S, Lily; Pinem, S.

2000-01-01

Fuel exchange of The RSG-GAS reactor core from uranium oxide to uranium silicide in the same loading, density, and enrichment, that is, 250 gr, 2.98 gr/cm 3 , and 19.75 % respectively, will be performed in-step wise. In every cycle of exchange with 5/l mode, it is needed to evaluate the parameter of reactor core operation. One of the important operation parameters is fuel reactivity that gives effect to the core reactivity. The experiment was performed at core no. 36, BOC, low power which exist 2 silicide fuels. The evaluation was done based on the RSG-GAS control rod calibration consisting of 40 fuels and 8 control rod.s. From 40 fuels in the core, there are 2 silicide fuels, RI-225/A-9 and RI-224/C-3. For inserting 2 silicide fuels, the reactivity effect to the core must be know. To know this effect , it was performed fuels reactivity experiment, which based on control rod calibration. But in this case the RSG-GAS has no other fresh oxide fuel so that configuration of the RSG-GAS core was rearranged by taking out the both silicide fuels and this configuration is used as reference core. Then silicide fuel RI-224 was inserted to position F-3 replacing the fresh oxide fuel RI-260 so the different reactivity of the fuels is obtained. The experiment result showed that the fuel reactivity change is in amount of 12.85 cent (0.098 % ) The experiment result was compared to the calculation result, using IAFUEL code which amount to 13.49 cent (0.103 %) The result showed that the reactivity change of oxide to silicide fuel is small so that the fuel exchange from uranium oxide to uranium silicide in the first step can be done without any significant change of the operation parameter

Introduction of mixed oxide fuel elements in the belgian cores

International Nuclear Information System (INIS)

Charlier, A.F.; Hollasky, N.A.

1994-01-01

The important amount of plutonium recovered from the reprocessing of spent fuel on the one hand, the national and international experience of the use of mixed oxide UO 2 -PuO 2 fuel in power reactors on the other hand, have led Belgian utilities to decide the introduction of Mixed-Oxide fuel in Doel unit 3 and Tihange unit 2 cores. The 'MOX' project has shown that it was possible without reducing safety or requiring modifications of the plant equipment. It has been approved by the Belgian 'Nuclear Safety Commission'. (authors). 1 tab., 2 figs

Design study and evaluation of fuel fabrication systems for FR fuel cycle

International Nuclear Information System (INIS)

Namekawa, Takashi; Tanaka, Kenya; Kawaguchi, Koichi; Koike, Kazuhiro; Shimuta, Hiroshi; Suzuki, Yoshihiro

2004-01-01

The plant concept for each FBR fuel fabrication system has been constructed and evaluated, which achieves economical improvement, decrease in the environmental burden, better resource utilization, and proliferation resistance by the various innovative techniques employed. The results are as follows: (1) For oxide fuels, the simplified pelletizing method has a high technical feasibility, and it is possible to apply this method to practical process at early stage, because this method is based on wealth results of a conventional method. (2) For oxide fuels, the sphere packing fuel fabrication system by gelation and vibro-compaction processes has the advantage of lesser dispersion of the fine powder due to the use of solution and granule in the process. However this system shoulders additional cost for the liquid waste treatment process to dispose a large bulk of process liquid waste. (3) For the metal fuel, the casting system is generally expected to have high economical efficiency even for small-scale facilities, although verification for fabrication of the TRU alloy slug is required. (author)

COMPARATIVE ANALYSIS OF STRUCTURAL CHANGES IN U-MO DISPERSED FUEL OF FULL-SIZE FUEL ELEMENTS AND MINI-RODS IRRADIATED IN THE MIR REACTOR

OpenAIRE

ALEKSEY. L. IZHUTOV; VALERIY. V. IAKOVLEV; ANDREY. E. NOVOSELOV; VLADIMIR. A. STARKOV; ALEKSEY. A. SHELDYAKOV; VALERIY. YU. SHISHIN; VLADIMIR. M. KOSENKOV; ALEKSANDR. V. VATULIN; IRINA. V. DOBRIKOVA; VLADIMIR. B. SUPRUN; GENNADIY. V. KULAKOV

2013-01-01

The paper summarizes the irradiation test and post-irradiation examination (PIE) data for the U-Mo low-enriched fuel that was irradiated in the MIR reactor under the RERTR Program. The PIE data were analyzed for both full-size fuel rods and mini-rods with atomized powder dispersed in Al matrix as well as with additions of 2%, 5% and 13% of silicon in the matrix and ZrN protective coating on the fuel particles. The full-size fuel rods were irradiated up to an average burnup of ∼ 60%235U; th...

TEM characterization of irradiated U-7Mo/Mg dispersion fuel

Science.gov (United States)

Gan, J.; Keiser, D. D.; Miller, B. D.; Jue, J. F.; Robinson, A. B.; Madden, J.

2017-10-01

This paper presents the results of transmission electron microscopy (TEM) characterization on neutron-irradiated samples taken from the low-flux and high-flux sides of the same fuel plate with U-7Mo fuel particles dispersed in Mg matrix with aluminum alloy Al6061 as cladding material that was irradiated edge-on to the core in the Advanced Test Reactor. The corresponding local fission density and fission rate of the fuel particles and the average fuel-plate centerline temperature for the low-flux and high-flux samples are estimated to be 3.7 × 1021 f/cm3, 7.4 × 1014 f/cm3/s and 123 °C, and 5.5 × 1021 f/cm3, 11.0 × 1014 f/cm3/s and 158 °C, respectively. Complex interaction layers developed at the Al-Mg interface, consisting of Al3Mg2 and Al12Mg17 along with precipitates of MgO, Mg2Si and FeAl5.3. No interaction between Mg matrix and U-Mo fuel particle was identified. For the U-Mo fuel particles, at low fission density, small elongated bubbles wrapped around the clean areas with a fission gas bubble superlattice, which suggests that bubble coalescence is an important mechanism for converting the fission gas bubble superlattice to large bubbles. At high fission density, no bubbles or porosity were observed in the Mg matrix, and pockets of residual fission gas bubble superlattice were observed in the U-Mo fuel particle interior.

Ductility Enhancement of Molybdenum Phase by Nano-sizedd Oxide Dispersions

Energy Technology Data Exchange (ETDEWEB)

Bruce Kang

2008-07-31

The present research is focused on ductility enhancement of molybdenum (Mo) alloys by adding nano-sized oxide particles to the alloy system. The research approach includes: (1) determination of microscopic mechanisms responsible for the macroscopic ductility enhancement effects through atomistic modeling of the metal-ceramic interface; (2) subsequent computer simulation-aided optimization of composition and nanoparticle size of the dispersion for improved performance; (3) synthesis and characterization of nanoparticle dispersion following the guidance from atomistic computational modeling analyses (e.g., by processing a small sample of Mo alloy for evaluation); and (4) experimental testing of the mechanical properties to determine optimal ductility enhancement.Through atomistic modeling and electronic structure analysis using full-potential linearized muffin-tin orbital (FP-LMTO) techniques, research to date has been performed on a number of selected chromium (Cr) systems containing nitrogen (N) and/or magnesium oxide (MgO) impurities. The emphasis has been on determining the properties of the valence electrons and the characteristics of the chemical bonds they formed. It was found that the brittle/ductile behavior of this transitional metal system is controlled by the relative population of valence charges: bonds formed by s valence electrons yield metallic, ductile behavior, whereas bonds formed by d valence electrons lead to covalent, brittle behavior. The presence of valence bands from impurities also affects the metal bonding, thereby explaining the detrimental and beneficial effects induced by the inclusion of N impurities and MgO dispersions. These understandings are useful for optimizing ductility enhancement effects on the dispersion materials.

Corrosion-resistant coating technique for oxide-dispersion-strengthened ferritic/martensitic steel

International Nuclear Information System (INIS)

Sakasegawa, Hideo; Tanigawa, Hiroyasu; Ando, Masami

2014-01-01

Oxide-dispersion-strengthened (ODS) steels are attractive materials for application as fuel cladding in fast reactors and first-wall material of fusion blanket. Recent studies have focused more on high-chromium ferritic (12-18 wt% Cr) ODS steels with attractive corrosion resistance properties. However, they have poor material workability, require complicated heat treatments for recrystallization, and possess anisotropic microstructures and mechanical properties. On the other hand, low-chromium ferritic/martensitic (8-9 wt% Cr) ODS steels have no such limitations; nonetheless, they have poor corrosion resistance properties. In our work, we developed a corrosion-resistant coating technique for a low-chromium ferritic/martensitic ODS steel. The ODS steel was coated with the 304 or 430 stainless steel, which has better corrosion resistances than the low-chromium ferritic/martensitic ODS steels. The 304 or 430 stainless steel was coated by changing the canning material from mild steel to stainless steel in the conventional material processing procedure for ODS steels. Microstructural observations and micro-hardness tests proved that the stainless steels were successfully coated without causing a deterioration in the mechanical property of the low-chromium ferritic/martensitic ODS steel. (author)

Fast reactors with axial arrangement of oxide and metal fuels in the core

International Nuclear Information System (INIS)

Troyanov, M.F.; Ilyunin, V.G.; Matveev, V.I.; Murogov, V.M.; Proshkin, A.A.; Rudneva, V.Ya.; Shmelev, A.N.

1980-01-01

Problems of using metal fuel in fast reactor (FR) core are discussed Results are given of the calculation of two-dimentional (R-Z) FR version having a composed core with the combined usage of oxide and metal fuels having parameters close to optimal from the point of view of fuel breeding rate, an oxide subzone having increased enrichment and a decreased proper conversion ratio. A reactor is considered where metallic fuel elements are placed from the side of ''cold'' coolant inlet (400-480 deg C), and oxide fuel elements - in the region where the coolant has a higher temperature (500-560 deg C). It is shown that the new fuel breeding rate in such a reactor can be increased by 20-30% as compared with an oxide fuel reactor. Growth of the total conversion ratio is mainly stipulated with the increase of the inner conversion ratio of the core (CRC) which is important not only from the point of view of nuclear fuel breeding rate but also the optimization of the mode of powerful fast reactor operation with provision for the change in reactivity in the process of its continuous operation. The fact, that the core version under investigation has a CRC value slightly exceeding unit, stipulates considerably less reactivity change as compared with the oxide version in the process of the reactor operation and permits at a constant reactor control system power to significantly increase the time between reloadings and, therefore, to increase the NPP load factor which is of great importance both from the point of view of economy and the improvement of operation conditions as well as of reactor operation reliability. It is concluded on the base of the analysis of the results obtained that FRs with the combined usage of oxide and metal fuels having an increased specific load and increased conversion ratio as compared with the oxide fuel FRs provide a higher rate of development of the whole nuclear power balanced with respect to the fuel [ru

Modeling of thermal expansion coefficient of perovskite oxide for solid oxide fuel cell cathode

Science.gov (United States)

Heydari, F.; Maghsoudipour, A.; Alizadeh, M.; Khakpour, Z.; Javaheri, M.

2015-09-01

Artificial intelligence models have the capacity to eliminate the need for expensive experimental investigation in various areas of manufacturing processes, including the material science. This study investigates the applicability of adaptive neuro-fuzzy inference system (ANFIS) approach for modeling the performance parameters of thermal expansion coefficient (TEC) of perovskite oxide for solid oxide fuel cell cathode. Oxides (Ln = La, Nd, Sm and M = Fe, Ni, Mn) have been prepared and characterized to study the influence of the different cations on TEC. Experimental results have shown TEC decreases favorably with substitution of Nd3+ and Mn3+ ions in the lattice. Structural parameters of compounds have been determined by X-ray diffraction, and field emission scanning electron microscopy has been used for the morphological study. Comparison results indicated that the ANFIS technique could be employed successfully in modeling thermal expansion coefficient of perovskite oxide for solid oxide fuel cell cathode, and considerable savings in terms of cost and time could be obtained by using ANFIS technique.

Degradation of solid oxide fuel cell metallic interconnects in fuels containing sulfur

Energy Technology Data Exchange (ETDEWEB)

Ziomek-Moroz, M.; Hawk, Jeffrey A.

2005-01-01

Hydrogen is the main fuel for all types of fuel cells except direct methanol fuel cells. Hydrogen can be generated from all manner of fossil fuels, including coal, natural gas, diesel, gasoline, other hydrocarbons, and oxygenates (e.g., methanol, ethanol, butanol, etc.). Impurities in the fuel can cause significant performance problems and sulfur, in particular, can decrease the cell performance of fuel cells, including solid oxide fuel cells (SOFC). In the SOFC, the high (800-1000°C) operating temperature yields advantages (e.g., internal fuel reforming) and disadvantages (e.g., material selection and degradation problems). Significant progress in reducing the operating temperature of the SOFC from ~1000 ºC to ~750 ºC may allow less expensive metallic materials to be used for interconnects and as balance of plant (BOP) materials. This paper provides insight on the material performance of nickel, ferritic steels, and nickel-based alloys in fuels containing sulfur, primarily in the form of H2S, and seeks to quantify the extent of possible degradation due to sulfur in the gas stream.

Computational sensitivity study of spray dispersion and mixing on the fuel properties in a gas turbine combustor

Energy Technology Data Exchange (ETDEWEB)

Grosshans, Holger; Szász, Robert-Zoltán [Division of Fluid Mechanics, Lund University (Sweden); Cao, Le [Key Laboratory for Aerosol-Cloud-Precipitation of China Meteorological Administration, Nanjing University of Information Science and Technology, Nanjing (China); Fuchs, Laszlo, E-mail: holger.grosshans@uclouvain.be [Department of Mechanics, KTH, Stockholm (Sweden)

2017-04-15

A swirl stabilized gas turbine burner has been simulated in order to assess the effects of the fuel properties on spray dispersion and fuel–air mixing. The properties under consideration include fuel surface tension, viscosity and density. The turbulence of the gas phase is modeled applying the methodology of large eddy simulation whereas the dispersed liquid phase is described by Lagrangian particle tracking. The exchange of mass, momentum and energy between the two phases is accounted for by two-way coupling. Bag and stripping breakup regimes are considered for secondary droplet breakup, using the Reitz–Diwakar and the Taylor analogy breakup models. Moreover, a model for droplet evaporation is included. The results reveal a high sensitivity of the spray structure to variations of all investigated parameters. In particular, a decrease in the surface tension or the fuel viscosity, or an increase in the fuel density, lead to less stable liquid structures. As a consequence, smaller droplets are generated and the overall spray surface area increases, leading to faster evaporation and mixing. Furthermore, with the trajectories of the small droplets being strongly influenced by aerodynamic forces (and less by their own inertia), the spray is more affected by the turbulent structures of the gaseous phase and the spray dispersion is enhanced. (paper)

Catalytic oxidative desulfurization of liquid hydrocarbon fuels using air

Science.gov (United States)

Sundararaman, Ramanathan

Conventional approaches to oxidative desulfurization of liquid hydrocarbons involve use of high-purity, expensive water soluble peroxide for oxidation of sulfur compounds followed by post-treatment for removal of oxidized sulfones by extraction. Both are associated with higher cost due to handling, storage of oxidants and yield loss with extraction and water separation, making the whole process more expensive. This thesis explores an oxidative desulfurization process using air as an oxidant followed by catalytic decomposition of sulfones thereby eliminating the aforementioned issues. Oxidation of sulfur compounds was realized by a two step process in which peroxides were first generated in-situ by catalytic air oxidation, followed by catalytic oxidation of S compounds using the peroxides generated in-situ completing the two step approach. By this technique it was feasible to oxidize over 90% of sulfur compounds present in real jet (520 ppmw S) and diesel (41 ppmw S) fuels. Screening of bulk and supported CuO based catalysts for peroxide generation using model aromatic compound representing diesel fuel showed that bulk CuO catalyst was more effective in producing peroxides with high yield and selectivity. Testing of three real diesel fuels obtained from different sources for air oxidation over bulk CuO catalyst showed different level of effectiveness for generating peroxides in-situ which was consistent with air oxidation of representative model aromatic compounds. Peroxides generated in-situ was then used as an oxidant to oxidize sulfur compounds present in the fuel over MoO3/SiO2 catalyst. 81% selectivity of peroxides for oxidation of sulfur compounds was observed on MoO3/SiO2 catalyst at 40 °C and under similar conditions MoO3/Al2O3 gave only 41% selectivity. This difference in selectivity might be related to the difference in the nature of active sites of MoO3 on SiO2 and Al2O 3 supports as suggested by H2-TPR and XRD analyses. Testing of supported and bulk Mg

The interfacial orientation relationship of oxide nanoparticles in a hafnium-containing oxide dispersion-strengthened austenitic stainless steel

Energy Technology Data Exchange (ETDEWEB)

Miao, Yinbin, E-mail: miao2@illinois.edu [Department of Nuclear, Plasma and Radiological Engineering, University of Illinois at Urbana-Champaign, Urbana, IL 61801 (United States); Mo, Kun [Nuclear Engineering Division, Argonne National Laboratory, Lemont, IL 60493 (United States); Cui, Bai [Department of Mechanical and Materials Engineering, University of Nebraska-Lincoln, Lincoln, NE 68588 (United States); Department of Material Science and Engineering, University of Illinois at Urbana-Champaign, Urbana, IL 61801 (United States); Chen, Wei-Ying [Department of Nuclear, Plasma and Radiological Engineering, University of Illinois at Urbana-Champaign, Urbana, IL 61801 (United States); Miller, Michael K.; Powers, Kathy A. [Center for Nanophase Materials Sciences, Oak Ridge National Laboratory, Oak Ridge, TN 37830 (United States); McCreary, Virginia; Gross, David [Department of Material Science and Engineering, University of Illinois at Urbana-Champaign, Urbana, IL 61801 (United States); Almer, Jonathan [X-ray Science Division, Argonne National Laboratory, Lemont, IL 60493 (United States); Robertson, Ian M. [Department of Material Science and Engineering, University of Wisconsin-Madison, Madison, WA 53706 (United States); Department of Material Science and Engineering, University of Illinois at Urbana-Champaign, Urbana, IL 61801 (United States); Stubbins, James F. [Department of Nuclear, Plasma and Radiological Engineering, University of Illinois at Urbana-Champaign, Urbana, IL 61801 (United States)

2015-03-15

This work reports comprehensive investigations on the orientation relationship of the oxide nanoparticles in a hafnium-containing austenitic oxide dispersion-strengthened 316 stainless steel. The phases of the oxide nanoparticles were determined by a combination of scanning transmission electron microscopy–electron dispersive X-ray spectroscopy, atom probe tomography and synchrotron X-ray diffraction to be complex Y–Ti–Hf–O compounds with similar crystal structures, including bixbyite Y{sub 2}O{sub 3}, fluorite Y{sub 2}O{sub 3}–HfO{sub 2} solid solution and pyrochlore (or fluorite) Y{sub 2}(Ti,Hf){sub 2−x}O{sub 7−x}. High resolution transmission electron microscopy was used to characterize the particle–matrix interfaces. Two different coherency relationships along with one axis-parallel relation between the oxide nanoparticles and the steel matrix were found. The size of the nanoparticles significantly influences the orientation relationship. The results provide insight into the relationship of these nanoparticles with the matrix, which has implications for interpreting material properties as well as responses to radiation. - Highlights: • The oxide nanoparticles in a hafnium-containing austenitic ODS were characterized. • The nanoparticles are Y–Hf–Ti–O enriched phases according to APT and STEM–EDS. • Two coherency and an axis-parallel orientation relationships were found by HR-TEM. • Particle size has a prominent effect on the orientation relationship (OR). • Formation mechanism of the oxide nanoparticles was discussed based on the ORs.

Study of catalysis for solid oxide fuel cells and direct methanol fuel cells

Science.gov (United States)

Jiang, Xirong

Fuel cells offer the enticing promise of cleaner electricity with lower environmental impact than traditional energy conversion technologies. Driven by the interest in power sources for portable electronics, and distributed generation and automotive propulsion markets, active development efforts in the technologies of both solid oxide fuel cell (SOFC) and direct methanol fuel cell (DMFC) devices have achieved significant progress. However, current catalysts for fuel cells are either of low catalytic activity or extremely expensive, presenting a key barrier toward the widespread commercialization of fuel cell devices. In this thesis work, atomic layer deposition (ALD), a novel thin film deposition technique, was employed to apply catalytic Pt to SOFC, and investigate both Pt skin catalysts and Pt-Ru catalysts for methanol oxidation, a very important reaction for DMFC, to increase the activity and utilization levels of the catalysts while simultaneously reducing the catalyst loading. For SOFCs, we explored the use of ALD for the fabrication of electrode components, including an ultra-thin Pt film for use as the electrocatalyst, and a Pt mesh structure for a current collector for SOFCs, aiming for precise control over the catalyst loading and catalyst geometry, and enhancement in the current collect efficiency. We choose Pt since it has high chemical stability and excellent catalytic activity for the O2 reduction reaction and the H2 oxidation reaction even at low operating temperatures. Working SOFC fuel cells were fabricated with ALD-deposited Pt thin films as an electrode/catalyst layer. The measured fuel cell performance reveals that comparable peak power densities were achieved for ALD-deposited Pt anodes with only one-fifth of the Pt loading relative to a DC-sputtered counterpart. In addition to the continuous electrocatalyst layer, a micro-patterned Pt structure was developed via the technique of area selective ALD. By coating yttria-stabilized zirconia, a

Spent oxide fuel regeneration by crystallization in molybdate melts

International Nuclear Information System (INIS)

Ustinov, O.A.; Sukhanov, L.P.; Yakunin, S.A.

2006-01-01

Paper describes a procedure to regenerate spent oxide fuel by its crystallization in molybdate melts. Paper presents the process procedures to regenerate spent fuel of both fast and thermal neutron reactors. One analyzes the advantages of the elaborated procedure [ru

Stability of solid oxide fuel cell materials

Energy Technology Data Exchange (ETDEWEB)

Armstrong, T.R.; Bates, J.L.; Chick, L.A. [Pacific Northwest Lab., Richland, WA (United States)

1996-04-01

Interconnection materials in a solid oxide fuel cell are exposed to both highly oxidizing conditions at the cathode and to highly reducing conditions at the anode. The thermal expansion characteristics of substituted lanthanum and yttrium chromite interconnect materials were evaluated by dilatometry as a function of oxygen partial pressures from 1 atm to 10{sup -18} atm, controlled using a carbon dioxide/hydrogen buffer.

Preparation of Highly Dispersed Reduced Graphene Oxide Decorated with Chitosan Oligosaccharide as Electrode Material for Enhancing the Direct Electron Transfer of Escherichia coli.

Science.gov (United States)

Luo, Zhimin; Yang, Dongliang; Qi, Guangqin; Yuwen, Lihui; Zhang, Yuqian; Weng, Lixing; Wang, Lianhui; Huang, Wei

2015-04-29

Water-dispersed reduced graphene oxide/chitosan oligosaccharide (RGO-CTSO) was prepared by chemical reduction of graphene oxide and synchronous functionalization with biocompatible chitosan oligosaccharide (CTSO). ζ potential measurement indicated that RGO-CTSO was highly stable in the acidic aqueous solution. RGO-CTSO was used to modify glassy carbon electrode (GCE) as the growth template of Escherichia coli (E. coli). The enhanced direct electron transfer of E. coli on the RGO-CTSO-modified GCE was studied by cyclic voltammetry. Compared with GCE or RGO-modified GCE, RGO-CTSO-modified GCE was more suitable for the adhesion growth of E. coli to improve direct electron transfer. The biocompatibility and versatility of RGO-CTSO made it promising for use as an anode material in microbial fuel cells.

Conversion of hydrocarbons in solid oxide fuel cells

DEFF Research Database (Denmark)

Mogensen, Mogens Bjerg; Kammer Hansen, K.

2003-01-01

Recently, a number of papers about direct oxidation of methane and hydrocarbon in solid oxide fuel cells (SOFC) at relatively low temperatures (about 700degreesC) have been published. Even though the conversion of almost dry CH4 at 1000degreesC on ceramic anodes was demonstrated more than 10 years...

Characteristics of SME biodiesel-fueled diesel particle emissions and the kinetics of oxidation.

Science.gov (United States)

Jung, Heejung; Kittelson, David B; Zachariah, Michael R

2006-08-15

Biodiesel is one of the most promising alternative diesel fuels. As diesel emission regulations have become more stringent, the diesel particulate filter (DPF) has become an essential part of the aftertreatment system. Knowledge of kinetics of exhaust particle oxidation for alternative diesel fuels is useful in estimating the change in regeneration behavior of a DPF with such fuels. This study examines the characteristics of diesel particulate emissions as well as kinetics of particle oxidation using a 1996 John Deere T04045TF250 off-highway engine and 100% soy methyl ester (SME) biodiesel (B100) as fuel. Compared to standard D2 fuel, this B100 reduced particle size, number, and volume in the accumulation mode where most of the particle mass is found. At 75% load, number decreased by 38%, DGN decreased from 80 to 62 nm, and volume decreased by 82%. Part of this decrease is likely associated with the fact that the particles were more easily oxidized. Arrhenius parameters for the biodiesel fuel showed a 2-3times greater frequency factor and approximately 6 times higher oxidation rate compared to regular diesel fuel in the range of 700-825 degrees C. The faster oxidation kinetics should facilitate regeneration when used with a DPF.

Enhanced performance of a novel anodic PdAu/VGCNF catalyst for electro-oxidation in a glycerol fuel cell.

Science.gov (United States)

Yahya, N; Kamarudin, S K; Karim, N A; Masdar, M S; Loh, K S

2017-11-25

This study presents a novel anodic PdAu/VGCNF catalyst for electro-oxidation in a glycerol fuel cell. The reaction conditions are critical issues affecting the glycerol electro-oxidation performance. This study presents the effects of catalyst loading, temperature, and electrolyte concentration. The glycerol oxidation performance of the PdAu/VGCNF catalyst on the anode side is tested via cyclic voltammetry with a 3 mm 2 active area. The morphology and physical properties of the catalyst are examined using X-ray diffraction (XRD), field emission scanning electron microscopy (SEM) and energy dispersive X-ray (EDX) spectroscopy. Then, optimization is carried out using the response surface method with central composite experimental design. The current density is experimentally obtained as a response variable from a set of experimental laboratory tests. The catalyst loading, temperature, and NaOH concentration are taken as independent parameters, which were evaluated previously in the screening experiments. The highest current density of 158.34 mAcm -2 is obtained under the optimal conditions of 3.0 M NaOH concentration, 60 °C temperature and 12 wt.% catalyst loading. These results prove that PdAu-VGCNF is a potential anodic catalyst for glycerol fuel cells.

On the fission gas release from oxide fuels during normal grain growth

International Nuclear Information System (INIS)

Paraschiv, M.C.; Paraschiv, A.; Glodeanu, F.

1997-01-01

A mathematical formalism for calculating the fission gas release from oxide fuels considering an arbitrary distribution of fuel grain size with only zero boundary condition for gas diffusion at the grain boundary is proposed. It has also been proved that it becomes unnecessary to consider the grain volume distribution function for fission products diffusion when the grain boundary gas resolution is considered, if thermodynamic forces on grain boundaries are only time dependent. In order to highlight the effect of the normal grain growth on fission gas release from oxide fuels Hillert's and Lifshitz and Slyozov's theories have been selected. The last one was used to give an adequate treatment of normal grain growth for the diffusion-controlled grain boundary movement in oxide fuels. It has been shown that during the fuel irradiation, the asymptotic form of the grain volume distribution functions given by Hillert and Lifshitz and Slyozov models can be maintained but the grain growth rate constant becomes time dependent itself. Experimental results have been used to correlate the two theoretical models of normal grain growth to the fission gas release from oxide fuels. (orig.)

Mathematical modeling of solid oxide fuel cells

Science.gov (United States)

Lu, Cheng-Yi; Maloney, Thomas M.

1988-01-01

Development of predictive techniques, with regard to cell behavior, under various operating conditions is needed to improve cell performance, increase energy density, reduce manufacturing cost, and to broaden utilization of various fuels. Such technology would be especially beneficial for the solid oxide fuel cells (SOFC) at it early demonstration stage. The development of computer models to calculate the temperature, CD, reactant distributions in the tubular and monolithic SOFCs. Results indicate that problems of nonuniform heat generation and fuel gas depletion in the tubular cell module, and of size limitions in the monolithic (MOD 0) design may be encountered during FC operation.

Solid oxide fuel cell having a monolithic core

International Nuclear Information System (INIS)

Ackerman, J.P.; Young, J.E.

1984-01-01

A solid oxide fuel cell for electrochemically combining fuel and oxidant for generating galvanic output, wherein the cell core has an array of electrolyte and interconnect walls that are substantially devoid of any composite inert materials for support. Instead, the core is monolithic, where each electrolyte wall consists of thin layers of cathode and anode materials sandwiching a thin layer of electrolyte material therebetween, and each interconnect wall consists of thin layers of the cathode and anode materials sandwiching a thin layer of interconnect material therebetween. The electrolyte walls are arranged and backfolded between adjacent interconnect walls operable to define a plurality of core passageways alternately arranged where the inside faces thereof have only the anode material or only the cathode material exposed. Means direct the fuel to the anode-exposed core passageways and means direct the oxidant to the cathode-exposed core passageway; and means also direct the galvanic output to an exterior circuit. Each layer of the electrolyte and interconnect materials is of the order of 0.002-0.01 cm thick; and each layer of the cathode and anode materials is of the order of 0.002-0.05 cm thick

Thermal conductivity of U–Mo/Al dispersion fuel. Effects of particle shape and size, stereography, and heat generation

International Nuclear Information System (INIS)

Cho, Tae Won; Sohn, Dong-Seong; Kim, Yeon Soo

2015-01-01

This paper describes the effects of particle sphericity, interfacial thermal resistance, stereography, and heat generation on the thermal conductivity of U–Mo/Al dispersion fuel. The ABAQUS finite element method (FEM) tool was used to calculate the effective thermal conductivity of U–Mo/Al dispersion fuel by implementing fuel particles. For U–Mo/Al, the particle sphericity effect was insignificant. However, if the effect of the interfacial thermal resistance between the fuel particles and Al matrix was considered, the thermal conductivity of U–Mo/Al was increased as the particle size increases. To examine the effect of stereography, we compared the two-dimensional modeling and three-dimensional modeling. The results showed that the two-dimensional modeling predicted lower than the three-dimensional modeling. We also examined the effect of the presence of heat sources in the fuel particles and found a decrease in thermal conductivity of U–Mo/Al from that of the typical homogeneous heat generation modeling. (author)

Research on the interfacial behaviors of plate-type dispersion nuclear fuel elements

Energy Technology Data Exchange (ETDEWEB)

Wang Qiming; Yan Xiaoqing [Department of Mechanics and Engineering Science, Fudan University, Shanghai 200433 (China); Ding Shurong, E-mail: dsr1971@163.co [Department of Mechanics and Engineering Science, Fudan University, Shanghai 200433 (China); Huo Yongzhong [Department of Mechanics and Engineering Science, Fudan University, Shanghai 200433 (China)

2010-04-01

The three-dimensional constitutive relations are constructed, respectively, for the fuel particles, the metal matrix and the cladding of dispersion nuclear fuel elements, allowing for the effects of large deformation and thermal-elastoplasticity. According to the constitutive relations, the method of modeling their irradiation behaviors in ABAQUS is developed and validated. Numerical simulations of the interfacial performances between the fuel meat and the cladding are implemented with the developed finite element models for different micro-structures of the fuel meat. The research results indicate that: (1) the interfacial tensile stresses and shear stresses for some cases will increase with burnup, but the relative stresses will decrease with burnup for some micro-structures; (2) at the lower burnups, the interfacial stresses increase with the particle sizes and the particle volume fractions; however, it is not the case at the higher burnups; (3) the particle distribution characteristics distinctly affect the interfacial stresses, and the face-centered cubic case has the best interfacial performance of the three considered cases.

Oxidation behaviour and electrical properties of cobalt/cerium oxide composite coatings for solid oxide fuel cell interconnects

DEFF Research Database (Denmark)

Harthøj, Anders; Holt, Tobias; Møller, Per

2015-01-01

This work evaluates the performance of cobalt/cerium oxide (Co/CeO2) composite coatings and pure Co coatings to be used for solid oxide fuel cell (SOFC) interconnects. The coatings are electroplated on the ferritic stainless steels Crofer 22 APU and Crofer 22H. Coated and uncoated samples...

Highly dispersed Pt-Ni nanoparticles on nitrogen-doped carbon nanotubes for application in direct methanol fuel cells.

Science.gov (United States)

Jiang, Shujuan; Ma, Yanwen; Tao, Haisheng; Jian, Guoqiang; Wang, Xizhang; Fan, Yining; Zhu, Jianmin; Hu, Zheng

2010-06-01

Binary Pt-Ni alloyed nanoparticles supported on nitrogen-doped carbon nanotubes (NCNTs) have been facilely constructed without pre-modification by making use of the active sites in NCNTs due to the N-participation. So-obtained binary Pt-Ni alloyed nanoparticles have been highly dispersed on the outer surface of the support with the size of about 3-4 nm. The electrochemical properties of the catalysts for methanol oxidation have been systematically evaluated. Binary Pt-Ni alloyed composites with molar ratio (Pt:Ni) of 3:2 and 3:1 present enhanced electrocatalytic activities and improved tolerance to CO poisoning as well as the similar stability, in comparison with the commercial Pt/C catalyst and the monometallic Pt/NCNTs catalysts. These results imply that so-constructed nanocomposite catalysts have the potential for applications in direct methanol fuel cells.

Development of a solid oxide fuel cell (SOFC) automotive auxiliary power unit (APU) fueled by gasoline

International Nuclear Information System (INIS)

DeMinco, C.; Mukerjee, S.; Grieve, J.; Faville, M.; Noetzel, J.; Perry, M.; Horvath, A.; Prediger, D.; Pastula, M.; Boersma, R.; Ghosh, D.

2000-01-01

This paper describes the design and the development progress of a 3 to 5 auxiliary power unit (APU) based on a gasoline fueled solid oxide fuel cell (SOFC). This fuel cell was supplied reformate gas (reactant) by a partial oxidation (POx) catalytic reformer utilizing liquid gasoline and designed by Delphi Automotive Systems. This reformate gas consists mainly of hydrogen, carbon monoxide and nitrogen and was fed directly in to the SOFC stack without any additional fuel reformer processing. The SOFC stack was developed by Global Thermoelectric and operates around 700 o C. This automotive APU produces power to support future 42 volt vehicle electrical architectures and loads. The balance of the APU, designed by Delphi Automotive Systems, employs a packaging and insulation design to facilitate installation and operation on-board automobiles. (author)

Investigation of aluminosilicate refractory for solid oxide fuel cell applications

Science.gov (United States)

Gentile, Paul Steven

Stationary solid oxide fuel cells (SOFCs) have been demonstrated to provide clean and reliable electricity through electro-chemical conversion of various fuel sources (CH4 and other light hydrocarbons). To become a competitive conversion technology the costs of SOFCs must be reduced to less than $400/kW. Aluminosilicate represents a potential low cost alternative to high purity alumina for SOFC refractory applications. The objectives of this investigation are to: (1) study changes of aluminosilicate chemistry and morphology under SOFC conditions, (2) identify volatile silicon species released by aluminosilicates, (3) identify the mechanisms of aluminosilicate vapor deposition on SOFC materials, and (4) determine the effects of aluminosilicate vapors on SOFC electrochemical performance. It is shown thermodynamically and empirically that low cost aluminosilicate refractory remains chemically and thermally unstable under SOFC operating conditions between 800°C and 1000°C. Energy dispersive spectroscopy (EDS) and X-ray photoelectron spectroscopy (XPS) of the aluminosilicate bulk and surface identified increased concentrations of silicon at the surface after exposure to SOFC gases at 1000°C for 100 hours. The presence of water vapor accelerated surface diffusion of silicon, creating a more uniform distribution. Thermodynamic equilibrium modeling showed aluminosilicate remains stable in dry air, but the introduction of water vapor indicative of actual SOFC gas streams creates low temperature (active anode interface.

Poly(vinylpyrrolidone) as dispersing agent for cerium-gadolinium oxide (CGO) suspensions

DEFF Research Database (Denmark)

Marani, Debora; Sudireddy, Bhaskar Reddy; Nielsen, Lotte

2016-01-01

The behaviour of selected poly(vinylpyrrolidone) grades to act as dispersant for ethanol-based ceriumgadolinium oxide suspensions was investigated and related to the molecular weight characteristics. The number, weight, and z-average molecular weights Mn, Mw, and Mz were determined by gel...

Microstructure characterizaton of advanced oxide fuel

International Nuclear Information System (INIS)

Rasmussen, D.E.; Gerber, E.W.; McCord, R.B.

1977-01-01

Preirradiation porosity, grain size, and microcomposition characteristics are presented for selected advanced oxide (PuO 2 -UO 2 ) LMFBR developmental fuels fabricated for irradiation testing in EBR-II. Quantitative microscopy, electron microprobe analysis, and a recently developed quantitative autoradiographic technique are utilized to relate microstructure characteristics to fabrication parameters

On the intrinsic transient capability and limitations of solid oxide fuel cell systems

OpenAIRE

Mueller, F; Jabbari, F; Brouwer, J

2009-01-01

The intrinsic transient performance capability and limitation of integrated solid oxide fuel cell (SOFC) systems is evaluated based on the system balance-of-plant response and fuel cell operating requirements (i.e., allowable deviation from nominal operation). Specifically, non-dimensional relations are derived from conservation principles that quantify the maximum instantaneous current increase that a solid oxide fuel cell system can safely manage based on (1) the desired fuel cell operating...

Jet-stirred reactor oxidation of alkane-rich FACE gasoline fuels

KAUST Repository

Chen, Bingjie

2016-06-23

Understanding species evolution upon gasoline fuel oxidation can aid in mitigating harmful emissions and improving combustion efficiency. Experimentally measured speciation profiles are also important targets for surrogate fuel kinetic models. This work presents the low- and high-temperature oxidation of two alkane-rich FACE gasolines (A and C, Fuels for Advanced Combustion Engines) in a jet-stirred reactor at 10. bar and equivalence ratios from 0.5 to 2 by probe sampling combined with gas chromatography and Fourier Transformed Infrared Spectrometry analysis. Detailed speciation profiles as a function of temperature are presented and compared to understand the combustion chemistry of these two real fuels. Simulations were conducted using three surrogates (i.e., FGA2, FGC2, and FRF 84), which have similar physical and chemical properties as the two gasolines. The experimental results reveal that the reactivity and major product distributions of these two alkane-rich FACE fuels are very similar, indicating that they have similar global reactivity despite their different compositions. The simulation results using all the surrogates capture the two-stage oxidation behavior of the two FACE gasolines, but the extent of low temperature reactivity is over-predicted. The simulations were analyzed, with a focus on the n-heptane and n-butane sub-mechanisms, to help direct the future model development and surrogate fuel formulation strategies.

Jet-stirred reactor oxidation of alkane-rich FACE gasoline fuels

KAUST Repository

Chen, Bingjie; Togbé , Casimir; Wang, Zhandong; Dagaut, Philippe; Sarathy, Mani

2016-01-01

Understanding species evolution upon gasoline fuel oxidation can aid in mitigating harmful emissions and improving combustion efficiency. Experimentally measured speciation profiles are also important targets for surrogate fuel kinetic models. This work presents the low- and high-temperature oxidation of two alkane-rich FACE gasolines (A and C, Fuels for Advanced Combustion Engines) in a jet-stirred reactor at 10. bar and equivalence ratios from 0.5 to 2 by probe sampling combined with gas chromatography and Fourier Transformed Infrared Spectrometry analysis. Detailed speciation profiles as a function of temperature are presented and compared to understand the combustion chemistry of these two real fuels. Simulations were conducted using three surrogates (i.e., FGA2, FGC2, and FRF 84), which have similar physical and chemical properties as the two gasolines. The experimental results reveal that the reactivity and major product distributions of these two alkane-rich FACE fuels are very similar, indicating that they have similar global reactivity despite their different compositions. The simulation results using all the surrogates capture the two-stage oxidation behavior of the two FACE gasolines, but the extent of low temperature reactivity is over-predicted. The simulations were analyzed, with a focus on the n-heptane and n-butane sub-mechanisms, to help direct the future model development and surrogate fuel formulation strategies.

Solid Oxide Fuel Cell Systems PVL Line

International Nuclear Information System (INIS)

Shearer, Susan; Rush, Gregory

2012-01-01

In July 2010, Stark State College (SSC), received Grant DE-EE0003229 from the U.S. Department of Energy (DOE), Golden Field Office, for the development of the electrical and control systems, and mechanical commissioning of a unique 20kW scale high-pressure, high temperature, natural gas fueled Stack Block Test System (SBTS). SSC worked closely with subcontractor, Rolls-Royce Fuel Cell Systems (US) Inc. (RRFCS) over a 13 month period to successfully complete the project activities. This system will be utilized by RRFCS for pre-commercial technology development and training of SSC student interns. In the longer term, when RRFCS is producing commercial products, SSC will utilize the equipment for workforce training. In addition to DOE Hydrogen, Fuel Cells, and Infrastructure Technologies program funding, RRFCS internal funds, funds from the state of Ohio, and funding from the DOE Solid State Energy Conversion Alliance (SECA) program have been utilized to design, develop and commission this equipment. Construction of the SBTS (mechanical components) was performed under a Grant from the State of Ohio through Ohio's Third Frontier program (Grant TECH 08-053). This Ohio program supported development of a system that uses natural gas as a fuel. Funding was provided under the Department of Energy (DOE) Solid-state Energy Conversion Alliance (SECA) program for modifications required to test on coal synthesis gas. The subject DOE program provided funding for the electrical build, control system development and mechanical commissioning. Performance testing, which includes electrical commissioning, was subsequently performed under the DOE SECA program. Rolls-Royce Fuel Cell Systems is developing a megawatt-scale solid oxide fuel cell (SOFC) stationary power generation system. This system, based on RRFCS proprietary technology, is fueled with natural gas, and operates at elevated pressure. A critical success factor for development of the full scale system is the capability to

The influence of deformation, annealing and recrystallisation on oxide nanofeatures in oxide dispersion strengthened steel

Energy Technology Data Exchange (ETDEWEB)

Dawson, Karl, E-mail: k.dawson@liverpool.ac.uk; Tatlock, Gordon J.

2017-04-01

This work demonstrates that Y-Ti oxide nanofeatures, observed in as-extruded oxide dispersion strengthened steel, are structurally modified by cold forging. A 950 °C heat treatment promoted restructuring of the deformed particles and partial recrystallisation of the cold forged alloy. Transmission electron microscopy revealed that cuboid shaped nanofeatures were deformed during forging, which resulted in high number densities of lens shaped yttrium-titanium oxide particles. Annealing the forged alloy promoted partial recrystallisation of the ferritic matrix. Particle morphology reverted from lens shaped, as witnessed in the deformed material, to cuboid shaped oxide nanofeatures, identical to those observed in as-extruded material. Precipitation distributions evaluated in both recrystallised and recovering grains were indistinguishable from those first measured in the as-extruded alloy. TEM images revealed a widespread orientation relationship between the oxide precipitates and the recrystallised grains; registration with the ferrite lattice was omnipresent in both recovering and recrystallised grains.

Uranium Oxide Rate Summary for the Spent Nuclear Fuel (SNF) Project (OCRWM)

Energy Technology Data Exchange (ETDEWEB)

PAJUNEN, A.L.

2000-09-20

The purpose of this document is to summarize the uranium oxidation reaction rate information developed by the Hanford Spent Nuclear Fuel (SNF) Project and describe the basis for selecting reaction rate correlations used in system design. The selection basis considers the conditions of practical interest to the fuel removal processes and the reaction rate application during design studies. Since the reaction rate correlations are potentially used over a range of conditions, depending of the type of evaluation being performed, a method for transitioning between oxidation reactions is also documented. The document scope is limited to uranium oxidation reactions of primary interest to the SNF Project processes. The reactions influencing fuel removal processes, and supporting accident analyses, are: uranium-water vapor, uranium-liquid water, uranium-moist air, and uranium-dry air. The correlation selection basis will consider input from all available sources that indicate the oxidation rate of uranium fuel, including the literature data, confirmatory experimental studies, and fuel element observations. Trimble (2000) summarizes literature data and the results of laboratory scale experimental studies. This document combines the information in Trimble (2000) with larger scale reaction observations to describe uranium oxidation rate correlations applicable to conditions of interest to the SNF Project.

Uranium Oxide Rate Summary for the Spent Nuclear Fuel (SNF) Project (OCRWM)

International Nuclear Information System (INIS)

PAJUNEN, A.L.

2000-01-01

The purpose of this document is to summarize the uranium oxidation reaction rate information developed by the Hanford Spent Nuclear Fuel (SNF) Project and describe the basis for selecting reaction rate correlations used in system design. The selection basis considers the conditions of practical interest to the fuel removal processes and the reaction rate application during design studies. Since the reaction rate correlations are potentially used over a range of conditions, depending of the type of evaluation being performed, a method for transitioning between oxidation reactions is also documented. The document scope is limited to uranium oxidation reactions of primary interest to the SNF Project processes. The reactions influencing fuel removal processes, and supporting accident analyses, are: uranium-water vapor, uranium-liquid water, uranium-moist air, and uranium-dry air. The correlation selection basis will consider input from all available sources that indicate the oxidation rate of uranium fuel, including the literature data, confirmatory experimental studies, and fuel element observations. Trimble (2000) summarizes literature data and the results of laboratory scale experimental studies. This document combines the information in Trimble (2000) with larger scale reaction observations to describe uranium oxidation rate correlations applicable to conditions of interest to the SNF Project

Influence of oxygen-metal ratio on mixed-oxide fuel performance

International Nuclear Information System (INIS)

Lawrence, L.A.; Leggett, R.D.

1979-04-01

The fuel oxygen-to-metal ratio (O/M) is recognized as an important consideration for performance of uranium--plutonium oxide fuels. An overview of the effects of differing O/M's on the irradiation performance of reference design mixed-oxide fuel in the areas of chemical and mechanical behavior, thermal performance, and fission gas behavior is presented. The pellet fuel has a nominal composition of 75 wt% UO 2 + 25 wt% PuO 2 at a pellet density of approx. 90% TD. for nominal conditions this results in a smeared density of approx. 85%. The cladding in all cases is 20% CW type 316 stainless steel with an outer diameter of 5.84 to 6.35 mm. O/M has been found to significantly influence fuel pin chemistry, mainly FCCI and fission product and fuel migration. It has little effect on thermal performance and overall mechanical behavior or fission gas release. The effects of O/M (ranging from 1.938 to 1.984) in the areas of fuel pin chemistry, to date, have not resulted in any reduction in fuel pin performance capability to goal burnups of approx. 8 atom% or more

Formation of Lamellar Structured Oxide Dispersion Strengthening Layers in Zircaloy-4

Energy Technology Data Exchange (ETDEWEB)

Jung, Yang-Il; Park, Jung-Hwan; Park, Dong-Jun; Kim, Hyun-Gil; Yang, Jae-Ho; Koo, Yang-Hyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Lim, Yoon-Soo [Hanbat National University, Daejeon (Korea, Republic of)

2016-10-15

Korea Atomic Energy Research Institute (KAERI) is one of the leading organizations for developing ATF claddings. One concept is to form an oxidation-resistant layer on Zr cladding surface. The other is to increase high-temperature mechanical strength of Zr tube. The oxide dispersion strengthened (ODS) zirconium was proposed to increase the strength of the Zr-based alloy up to high temperatures. According to our previous investigations, the tensile strength of Zircaloy-4 was increased by up to 20% with the formation of a thin dispersed oxide layer with a thickness less than 10% of that of the Zircaloy-4 substrate. However, the tensile elongation of the samples decreased drastically. The brittle fracture was a major concern in development of the ODS Zircaloy-4. In this study, a lamellar structure of ODS layer was formed to increase ductility of the ODS Zircaloy-4. The mechanical properties were varied depending on the structure of ODS layer. For example, the partial formation of ODS layer with the thickness of 10% to the substrate thickness induced the increase in tensile strength up to about 20% than fresh Zircaloy-4.

Application of the DART Code for the Assessment of Advanced Fuel Behavior

International Nuclear Information System (INIS)

Rest, J.; Totev, T.

2007-01-01

The Dispersion Analysis Research Tool (DART) code is a dispersion fuel analysis code that contains mechanistically-based fuel and reaction-product swelling models, a one dimensional heat transfer analysis, and mechanical deformation models. DART has been used to simulate the irradiation behavior of uranium oxide, uranium silicide, and uranium molybdenum aluminum dispersion fuels, as well as their monolithic counterparts. The thermal-mechanical DART code has been validated against RERTR tests performed in the ATR for irradiation data on interaction thickness, fuel, matrix, and reaction product volume fractions, and plate thickness changes. The DART fission gas behavior model has been validated against UO 2 fission gas release data as well as measured fission gas-bubble size distributions. Here DART is utilized to analyze various aspects of the observed bubble growth in U-Mo/Al interaction product. (authors)

Post-irradiation analysis of low enriched U-Mo/Al dispersions fuel miniplate tests, RERTR 4 and 5

International Nuclear Information System (INIS)

Hofman, G.L.; Finlay, M.R.; Kim, Y.S.

2005-01-01

Interpretation of the post irradiation data of U-Mo/Al dispersion fuel mini plates irradiated in the Advanced Test Reactor to a maximum U-235 burn up of 80% are presented. The analyses addresses fuel swelling and porosity formation as these fuel performance issues relate to fuel fabrication and irradiation parameters. Specifically, mechanisms involved in the formation of porosity observed in the U-Mo/Al interaction phase are discussed and, means of mitigating or eliminating this irradiation phenomenon are offered. (author)

Initiation of depleted uranium oxide and spent fuel testing for the spent fuel sabotage aerosol ratio program

Energy Technology Data Exchange (ETDEWEB)

Molecke, M.A.; Gregson, M.W.; Sorenson, K.B. [Sandia National Labs. (United States); Billone, M.C.; Tsai, H. [Argonne National Lab. (United States); Koch, W.; Nolte, O. [Fraunhofer Inst. fuer Toxikologie und Experimentelle Medizin (Germany); Pretzsch, G.; Lange, F. [Gesellschaft fuer Anlagen- und Reaktorsicherheit (Germany); Autrusson, B.; Loiseau, O. [Inst. de Radioprotection et de Surete Nucleaire (France); Thompson, N.S.; Hibbs, R.S. [U.S. Dept. of Energy (United States); Young, F.I.; Mo, T. [U.S. Nuclear Regulatory Commission (United States)

2004-07-01

We provide a detailed overview of an ongoing, multinational test program that is developing aerosol data for some spent fuel sabotage scenarios on spent fuel transport and storage casks. Experiments are being performed to quantify the aerosolized materials plus volatilized fission products generated from actual spent fuel and surrogate material test rods, due to impact by a high energy density device, HEDD. The program participants in the U.S. plus Germany, France, and the U.K., part of the international Working Group for Sabotage Concerns of Transport and Storage Casks, WGSTSC have strongly supported and coordinated this research program. Sandia National Laboratories, SNL, has the lead role for conducting this research program; test program support is provided by both the U.S. Department of Energy and Nuclear Regulatory Commission. WGSTSC partners need this research to better understand potential radiological impacts from sabotage of nuclear material shipments and storage casks, and to support subsequent risk assessments, modeling, and preventative measures. We provide a summary of the overall, multi-phase test design and a description of all explosive containment and aerosol collection test components used. We focus on the recently initiated tests on ''surrogate'' spent fuel, unirradiated depleted uranium oxide, and forthcoming actual spent fuel tests. The depleted uranium oxide test rodlets were prepared by the Institut de Radioprotection et de Surete Nucleaire, in France. These surrogate test rodlets closely match the diameter of the test rodlets of actual spent fuel from the H.B. Robinson reactor (high burnup PWR fuel) and the Surry reactor (lower, medium burnup PWR fuel), generated from U.S. reactors. The characterization of the spent fuels and fabrication into short, pressurized rodlets has been performed by Argonne National Laboratory, for testing at SNL. The ratio of the aerosol and respirable particles released from HEDD-impacted spent

Effect of Co deposition on oxidation behavior and electrical properties of ferritic steel for solid oxide fuel cell interconnects

Energy Technology Data Exchange (ETDEWEB)

Kruk, A.; Adamczyk, A.; Gil, A. [AGH University of Science and Technology, Faculty of Materials Science and Ceramics, al. Mickiewicza 30, 30-059 Krakow (Poland); Kąc, S. [AGH University of Science and Technology, Faculty of Metals Engineering and Industrial Computer Science, al. Mickiewicza 30, 30-059 Krakow (Poland); Dąbek, J.; Ziąbka, M. [AGH University of Science and Technology, Faculty of Materials Science and Ceramics, al. Mickiewicza 30, 30-059 Krakow (Poland); Brylewski, T., E-mail: brylew@agh.edu.pl [AGH University of Science and Technology, Faculty of Materials Science and Ceramics, al. Mickiewicza 30, 30-059 Krakow (Poland)

2015-09-01

In this work, a Co layer deposited on DIN 50049 steel by means of pulsed laser deposition was applied for the protection of solid oxide fuel cell (SOFC) interconnects operating on the cathode side. The coated and uncoated steel samples were oxidized in air at 1073 K for 500 h, and their microstructures as well as electrical resistances were evaluated using X-ray diffraction, atomic force microscopy, scanning electron microscopy coupled with energy-dispersive X-ray spectroscopy, and the 2-probe 4-point direct current method. It was demonstrated that the Co coating had reduced the oxidation rate of the steel by nearly a half. The area-specific resistance value of the coated steel was 5 × 10{sup −6} Ω·m{sup 2}, which was significantly lower than that of bare steel after 350 h of oxidation at 1073 K. Cr vaporization tests showed that the Co coating was efficient at blocking the outward diffusion of Cr. The obtained results prove that steel coated with a thin film of cobalt was suitable for use as metallic interconnect material in SOFCs operating at intermediate temperatures. - Highlights: • Co layer was deposited on ferritic steel by means of pulsed laser deposition. • Coated and bare ferritic steel samples were exposed to air at 1073 K for 500 h. • Scale growth rate on bare steel is higher than that on coated steel. • Electrical resistance for oxidized coated steel was lower than for bare steel. • Co-coated steel effectively reduced the formation of volatile Cr species.

A development of solid oxide fuel cell technology

Energy Technology Data Exchange (ETDEWEB)

Lim, Hee Chun; Lee, Chang Woo [Korea Electric Power Corp. (KEPCO), Taejon (Korea, Republic of). Research Center; Kim, Kwy Youl; Yoon, Moon Soo; Kim, Ho Ki; Kim, Young Sik; Mun, Sung In; Eom, Sung Wuk [Korea Electrotechnology Research Inst., Changwon (Korea, Republic of)

1996-12-31

Solid oxide fuel cell which was consisted of ceramics has high power density and is very simple in shape. The project named A development of SOFC(Solid Oxide Fuel Cell) technology is to develop the unit cell fabrication processing and to evaluate the unit cell of solid oxide full cell. In this project, a manufacturing process of cathode by citrate method and polymeric precursor methods were established. By using tape casting method, high density thin electrolyte was manufactured and has high performance. Unit cell composed with La{sub 17}Sr{sub 13}Mn{sub 3} as cathode, 8YSZ electrolyte and 50% NiYSZ anode had a performance of O.85 W/cm{sup 2} and recorded 510 hours operation time. On the basis of these results. 100 cm{sup 2} class unit cell will be fabricated and tests in next program (author). 59 refs., 120 figs.

A development of solid oxide fuel cell technology

Energy Technology Data Exchange (ETDEWEB)

Lim, Hee Chun; Lee, Chang Woo [Korea Electric Power Corp. (KEPCO), Taejon (Korea, Republic of). Research Center; Kim, Kwy Youl; Yoon, Moon Soo; Kim, Ho Ki; Kim, Young Sik; Mun, Sung In; Eom, Sung Wuk [Korea Electrotechnology Research Inst., Changwon (Korea, Republic of)

1995-12-31

Solid oxide fuel cell which was consisted of ceramics has high power density and is very simple in shape. The project named A development of SOFC(Solid Oxide Fuel Cell) technology is to develop the unit cell fabrication processing and to evaluate the unit cell of solid oxide full cell. In this project, a manufacturing process of cathode by citrate method and polymeric precursor methods were established. By using tape casting method, high density thin electrolyte was manufactured and has high performance. Unit cell composed with La{sub 17}Sr{sub 13}Mn{sub 3} as cathode, 8YSZ electrolyte and 50% NiYSZ anode had a performance of O.85 W/cm{sup 2} and recorded 510 hours operation time. On the basis of these results. 100 cm{sup 2} class unit cell will be fabricated and tests in next program (author). 59 refs., 120 figs.

Mathematical modeling of synthesis gas fueled electrochemistry and transport including H2/CO co-oxidation and surface diffusion in solid oxide fuel cell

Science.gov (United States)

Bao, Cheng; Jiang, Zeyi; Zhang, Xinxin

2015-10-01

Fuel flexibility is a significant advantage of solid oxide fuel cell (SOFC). A comprehensive macroscopic framework is proposed for synthesis gas (syngas) fueled electrochemistry and transport in SOFC anode with two main novelties, i.e. analytical H2/CO electrochemical co-oxidation, and correction of gas species concentration at triple phase boundary considering competitive absorption and surface diffusion. Staring from analytical approximation of the decoupled charge and mass transfer, we present analytical solutions of two defined variables, i.e. hydrogen current fraction and enhancement factor. Giving explicit answer (rather than case-by-case numerical calculation) on how many percent of the current output contributed by H2 or CO and on how great the water gas shift reaction plays role on, this approach establishes at the first time an adaptive superposition mechanism of H2-fuel and CO-fuel electrochemistry for syngas fuel. Based on the diffusion equivalent circuit model, assuming series-connected resistances of surface diffusion and bulk diffusion, the model predicts well at high fuel utilization by keeping fixed porosity/tortuosity ratio. The model has been validated by experimental polarization behaviors in a wide range of operation on a button cell for H2-H2O-CO-CO2-N2 fuel systems. The framework could be helpful to narrow the gap between macro-scale and meso-scale SOFC modeling.

Fabrication procedures for manufacturing high uranium concentration dispersion fuel elements; Procedimentos de fabricacao de elementos combustiveis a base de dispersoes com alta concentracao de uranio

Energy Technology Data Exchange (ETDEWEB)

Souza, Jose Antonio Batista de

2011-07-01

IPEN-CNEN/SP developed the technology to produce the dispersion type fuel elements for research reactors and made it available for routine production. Today, the fuel produced in IPEN-CNEN/SP is limited to the uranium concentration of 3.0 gU/cm{sup 3} for U{sub 3}Si{sub 2}-Al dispersion-based and 2.3 gU/cm{sup 3} for U{sub 3}O{sub 8}-Al dispersion. The increase of uranium concentration in fuel plates enables the reactivity of the reactor core reactivity to be higher and extends the fuel life. Concerning technology, it is possible to increase the uranium concentration in the fuel meat up to the limit of 4.8 gU/cm{sup 3} in U{sub 3}Si{sub 2}-Al dispersion and 3.2 gU/cm{sup 3} U{sub 3}O{sub 8}-Al dispersion. These dispersions are well qualified worldwide. This work aims to develop the manufacturing process of both fuel meats with high uranium concentrations, by redefining the manufacturing procedures currently adopted in the Nuclear Fuel Center of IPEN-CNEN/SP. Based on the results, it was concluded that to achieve the desired concentration, it is necessary to make some changes in the established procedures, such as in the particle size of the fuel powder and in the feeding process inside the matrix, before briquette pressing. These studies have also shown that the fuel plates, with a high concentration of U{sub 3}Si{sub 2}-Al, met the used specifications. On the other hand, the appearance of the microstructure obtained from U{sub 3}O{sub 8}-Al dispersion fuel plates with 3.2 gU/cm{sup 3} showed to be unsatisfactory, due to the considerably significant porosity observed. The developed fabrication procedure was applied to U{sub 3}Si{sub 2} production at 4.8 gU/cm{sup 3}, with enriched uranium. The produced plates were used to assemble the fuel element IEA-228, which was irradiated in order to check its performance in the IEA-R1 reactor at IPEN-CNEN/SP. These new fuels have potential to be used in the new Brazilian Multipurpose Reactor - RMB. (author)

Electrode Reaction Pathway in Oxide Anode for Solid Oxide Fuel Cells

Science.gov (United States)

Li, Wenyuan

Oxide anodes for solid oxide fuel cells (SOFC) with the advantage of fuel flexibility, resistance to coarsening, small chemical expansion and etc. have been attracting increasing interest. Good performance has been reported with a few of perovskite structure anodes, such as (LaSr)(CrMn)O3. However, more improvements need to be made before meeting the application requirement. Understanding the oxidation mechanism is crucial for a directed optimization, but it is still on the early stage of investigation. In this study, reaction mechanism of oxide anodes is investigated on doped YCrO 3 with H2 fuel, in terms of the origin of electrochemical activity, rate-determining steps (RDS), extension of reactive zone, and the impact from overpotential under service condition to those properties. H2 oxidation on the YCs anodes is found to be limited by charge transfer and H surface diffusion. A model is presented to describe the elementary steps in H2 oxidation. From the reaction order results, it is suggested that any models without taking H into the charge transfer step are invalid. The nature of B site element determines the H2 oxidation kinetics primarily. Ni displays better adsorption ability than Co. However, H adsorption ability of such oxide anode is inferior to that of Ni metal anode. In addition, the charge transfer step is directly associated with the activity of electrons in the anode; therefore it can be significantly promoted by enhancement of the electron activity. It is found that A site Ca doping improves the polarization resistance about 10 times, by increasing the activity of electrons to promote the charge transfer process. For the active area in the oxide anode, besides the traditional three-phase boundary (3PB), the internal anode surface as two-phase boundary (2PB) is proven to be capable of catalytically oxidizing the H2 fuel also when the bulk lattice is activated depending on the B site elements. The contribution from each part is estimated by switching

Water dispersal and functionalization of hydrophobic Iron oxide nanoparticles with lipid-modified poly(amidoamine) dendrimers

NARCIS (Netherlands)

Boni, A; Albertazzi, L.; Innocenti, C; Gemmi, M; Bifone, A

2013-01-01

A novel and facile method for water dispersal of hydrophobic iron oxide nanoparticles based on the amphiphilic PAMAM-C-12 dendrimer is described. Stable and highly concentrated water dispersions of multifunctional, magnetic nanoparticles were obtained with this single-step approach, and showed

Application of the monolithic solid oxide fuel cell to space power systems

International Nuclear Information System (INIS)

Myles, K.M.; Bhattacharyya, S.K.

1991-01-01

The monolithic solid-oxide fuel cell (MSOFC) is a promising electrochemical power generation device that is currently under development at Argonne National Laboratory. The extremely high power density of the MSOFC leads to MSOFC systems that have sufficiently high energy densities that they are excellent candidates for a number of space missions. The fuel cell can also be operated in reverse, if it can be coupled to an external power source, to regenerate the fuel and oxidant from the water product. This feature further enhances the potential mission applications of the MSOFC. In this paper, the current status of the fuel cell development is presented---the focus being on fabrication and currently achievable performance. In addition, a specific example of a space power system, featuring a liquid metal cooled fast spectrum nuclear reactor and a monolithic solid oxide fuel cell, is presented to demonstrate the features of an integrated system

Application of the monolithic solid oxide fuel cell to space power systems

Science.gov (United States)

Myles, Kevin M.; Bhattacharyya, Samit K.

1991-01-01

The monolithic solid-oxide fuel cell (MSOFC) is a promising electrochemical power generation device that is currently under development at Argonne National Laboratory. The extremely high power density of the MSOFC leads to MSOFC systems that have sufficiently high energy densities that they are excellent candidates for a number of space missions. The fuel cell can also be operated in reverse, if it can be coupled to an external power source, to regenerate the fuel and oxidant from the water product. This feature further enhances the potential mission applications of the MSOFC. In this paper, the current status of the fuel cell development is presented—the focus being on fabrication and currently achievable performance. In addition, a specific example of a space power system, featuring a liquid metal cooled fast spectrum nuclear reactor and a monolithic solid oxide fuel cell, is presented to demonstrate the features of an integrated system.

Modern methods of material accounting for mixed oxide fuel fabrication facility

International Nuclear Information System (INIS)

Eggers, R.F.; Pindak, J.L.; Brouns, R.J.; Williams, R.C.; Brite, D.W.; Kinnison, R.R.; Fager, J.E.

1981-01-01

The generic requirements loss detection, and response to alarms of a contemporary material control and accounting (MCandA) philosophy have been applied to a mixed oxide fuel fabrication plant to produce a detailed preliminary MCandA system design that is generally applicable to facilities of this type. This paper summarizes and discusses detailed results of the mixed oxide fuel fabrication plant study

Modern methods of material accounting for mixed-oxide fuel-fabrication facility

International Nuclear Information System (INIS)

Eggers, R.F.; Brouns, R.J.; Brite, D.W.; Pindak, J.L.

1981-07-01

The generic requirements loss detection, and response to alarms of a contemporary material control and accounting (MC and A) philosophy have been applied to a mixed-oxide fuel-fabrication plant to produce a detailed preliminary MC and A system design that is generally applicable to facilities of this type. This paper summarizes and discusses detailed results of the mixed-oxide fuel-fabrication plant study. Topics covered in this paper include: mixed-oxide fuel-fabrication process description, process disaggregation into MC and A system control units, quantitative results of analysis of control units for abrupt and recurring loss-detection capability, impact of short- and long-term holdup on loss-detection capability, response to alarms for abrupt loss, and response to alarms for recurring loss

Computer Simulations of Composite Electrodes in Solid-Oxide Fuel-Cells

Energy Technology Data Exchange (ETDEWEB)

Sunde, Svein

1999-07-01

Fuel cells are devices for converting the combined chemical (free) energy of fuels and oxygen (air) directly to electrical energy without relying on the dynamic action of steam heated by reacting fuel-oxygen mixtures, like in steam turbines, or of the reacting gas mixtures themselves, like in gas turbines. The basic rationale for fuel cells is their high efficiencies as compared to indirect-conversion methods. Fuel cells are currently being considered for a number of applications, among them de-centralised power supply. Fuel cells come in five basic types and are usually classified according to the type of electrolyte used, which in turn to a significant degree limits the options for anode and cathode materials. The solid-oxide fuel-cell (SOFC) , with which this thesis is concerned, is thus named after its oxide electrolyte, typically the oxide-ion conducting material yttria-stabilised zirconia (YSZ). While the cathode of an SOFC is often uniform in chemical composition (or at least intended to be), various problems of delamination, cracking etc. associated with the use of metallic anode electrocatalysts led to the development of composite SOFC anodes. Porous anodes consisting of Ni and YSZ particles in roughly 50/50 wt-% mixtures are now almost standard with any SOFC-development programme. The designer of composite SOFC electrodes is faced with at least three, interrelated questions: (1) What will be the optimum microstructure and composition of the composite electrode? (2) If the structure changes during operation, as is often observed, what will be the consequences for the internal losses in the cell? (3) How do we interpret electrochemical and conductivity measurements with regard to structure and composition? It is the primary purpose of this thesis to provide a framework for modelling the electrochemical and transport properties of composite electrodes for SOFC, and to arrive at some new insights that cannot be offered by experiment alone. Emphasis is put on

Hydrogen peroxide oxidant fuel cell systems for ultra-portable applications

Science.gov (United States)

Valdez, T. I.; Narayanan, S. R.

2001-01-01

This paper will address the issues of using hydrogen peroxide as an oxidant fuel in a miniature DMFC system. Cell performance for DMFC based fuel cells operating on hydrogen peroxide will be presented and discussed.

Materials for high temperature solid oxide fuel cells

International Nuclear Information System (INIS)

Singhal, S.C.

1987-01-01

High temperature solid oxide fuel cells show great promise for economical production of electricity. These cells are based upon the ability of stabilized zirconia to operate as an oxygen ion conductor at elevated temperatures. The design of the tubular solid oxide fuel cell being pursued at Westinghouse is illustrated. The cell uses a calcia-stabilized zironcia porous support tube, which acts both as a structural member onto which the other cell components are fabricated in the form of thin layers, and as a functional member to allow the passage, via its porosity, of air (or oxygen) to the air electrode. This paper summarizes the materials and fabrication processes for the various cell components

Corrosion mechanisms of spent fuel under oxidizing conditions

International Nuclear Information System (INIS)

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

1997-01-01

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

Microstructure evolution of the oxide dispersion strengthened CLAM steel during mechanical alloying process

Energy Technology Data Exchange (ETDEWEB)

Song, Liangliang [Key Laboratory of Neutronics and Radiation Safety, Institute of Nuclear Energy Safety Technology, Chinese Academy of Science, Hefei, Anhui, 230031 (China); University of Science and Technology of China, Hefei, Anhui, 230031 (China); Liu, Shaojun, E-mail: shaojun.liu@fds.org.cn [Key Laboratory of Neutronics and Radiation Safety, Institute of Nuclear Energy Safety Technology, Chinese Academy of Science, Hefei, Anhui, 230031 (China); Mao, Xiaodong [Key Laboratory of Neutronics and Radiation Safety, Institute of Nuclear Energy Safety Technology, Chinese Academy of Science, Hefei, Anhui, 230031 (China)

2016-11-15

Highlights: • A nano-sized oxides dispersed ODS-CLAM steel was obtained by MA and HIP. • A minimum saturated grain size of down to 30 nm was achieved by varying the milling time from 0 to 100 h. • Solution of W in the MA powder could be significantly improved by increasing MA rotation speed. - Abstracts: Oxide dispersion strengthened Ferritic/Martensitic steel is considered as one of the most potential structural material for future fusion reactor, owing to its high mechanical properties and good irradiation resistance. The oxide dispersion strengthened China Low Activation Martensitic (ODS-CLAM) steel was fabricated by mechanical alloying (MA) and hot isostatic pressing (HIP). The microstructural evolutions during the process of ball milling and subsequent consolidation were investigated by SEM, XRD and TEM. The results showed that increasing the milling time during the first 36 h milling could effectively decrease the grain size to a value of around 30 nm, over which grain sized remained nearly constant. Increasing the rotation speed promoted the solution of tungsten (W) element obviously and decreased the grain size to a certain degree. Observation on the consolidated and further heat-treated ODS-CLAM steel samples indicated that a martensite microstructure with a high density of nano-particles was achieved.

Process for recovery of plutonium from fabrication residues of mixed fuels consisting of uranium oxide and plutonium oxide

International Nuclear Information System (INIS)

Heremanns, R.H.; Vandersteene, J.J.

1983-01-01

The invention concerns a process for recovery of plutonium from fabrication residues of mixed fuels consisting of uranium oxide and plutonium oxide in the form of PuO 2 . Mixed fuels consisting of uranium oxide and plutonium oxide are being used more and more. The plants which prepare these mixed fuels have around 5% of the total mass of fuels as fabrication residue, either as waste or scrap. In view of the high cost of plutonium, it has been attempted to recover this plutonium from the fabrication residues by a process having a purchase price lower than the price of plutonium. The problem is essentially to separate the plutonium, the uranium and the impurities. The residues are fluorinated, the UF 6 and PuF 6 obtained are separated by selective absorption of the PuF 6 on NaF at a temperature of at least 400 0 C, the complex obtained by this absorption is dissolved in nitric acid solution, the plutonium is precipitated in the form of plutonium oxalate by adding oxalic acid, and the precipitated plutonium oxalate is calcined

Grain and burnup dependence of spent fuel oxidation: geological repository impact

International Nuclear Information System (INIS)

Hanson, B. D.; Kansa, E. J.; Stoot, R.B.

1998-01-01

Further refinements to the oxidation model of Stout et al. have been made. The present model incorporates the burnup dependence of the oxidation rate in addition to an allowance for a distribution of grain sizes. The model was tested by comparing the model results with the oxidation histories of spent fuel samples oxidized in Thermogravimetric Analysis (TGA) or Oven Dry-Bath (ODB) experiments. The comparison between the experimental and model results are remarkably close and confirm the assumption that grain-size distributions and activation energies are the important parameters to predicting oxidation behavior. The burnup dependence of the activation energy was shown to have a greater effect than decreasing the effective grain size in suppressing the rate of the reaction U 4 O 9 (rightwards arrow)U 3 O 4 . Model results predict that U 3 O 8 formation of spent fuels exposed to oxygen will be suppressed even for high burnup fuels that have undergone restructuring in the rim region, provided the repository temperature is kept sufficient

Radiation Stability of Nanoclusters in Nano-structured Oxide Dispersion Strengthened (ODS) Steels

International Nuclear Information System (INIS)

Certain, Alicia G.; Kuchibhatla, Satyanarayana; Shutthanandan, V.; Allen, T. R.

2013-01-01

Nanostructured oxide dispersion strengthened (ODS) steels are considered candidates for nuclear fission and fusion applications at high temperature and dose. The complex oxide nanoclusters in these alloys provide high-temperature strength and are expected to afford better radiation resistance. Proton, heavy ion, and neutron irradiations have been performed to evaluate cluster stability in 14YWT and 9CrODS steel under a range of irradiation conditions. Energy-filtered transmission electron microscopy and atom probe tomography were used in this work to analyze the evolution of the oxide population.

Oxide nanoparticles in an Al-alloyed oxide dispersion strengthened steel: crystallographic structure and interface with ferrite matrix

Science.gov (United States)

Zhang, Zhenbo; Pantleon, Wolfgang

2017-07-01

Oxide nanoparticles are quintessential for ensuring the extraordinary properties of oxide dispersion strengthened (ODS) steels. In this study, the crystallographic structure of oxide nanoparticles, and their interface with the ferritic steel matrix in an Al-alloyed ODS steel, i.e. PM2000, were systematically investigated by high-resolution transmission electron microscopy. The majority of oxide nanoparticles were identified to be orthorhombic YAlO3. During hot consolidation and extrusion, they develop a coherent interface and a near cuboid-on-cube orientation relationship with the ferrite matrix in the material. After annealing at 1200 °C for 1 h, however, the orientation relationship between the oxide nanoparticles and the matrix becomes arbitrary, and their interface mostly incoherent. Annealing at 1300 °C leads to considerable coarsening of oxide nanoparticles, and a new orientation relationship of pseudo-cube-on-cube between oxide nanoparticles and ferrite matrix develops. The reason for the developing interfaces and orientation relationships between oxide nanoparticles and ferrite matrix under different conditions is discussed.

Fission gas behavior in mixed-oxide fuel during transient overpower

International Nuclear Information System (INIS)

Randklev, E.H.; Treibs, H.A.; Mastel, B.; Baldwin, D.L.

1979-01-01

Fission gas behavior can be important in determining fuel pin and core performance during a reactor transient. The results are presented of examinations characterizing the changes in microstructural distribution and retention of fission gas in fuel for a series of transient overpower (50 cents/s) tested mixed-oxide fuel pins and their steady state siblings

Development of MTR fuel plate with U-Al dispersion core constituents

International Nuclear Information System (INIS)

Bressiani, Jose Carlos

1979-01-01

This work is a contribution to the development of fuel plates for Research Nuclear Reaction Materials Test Reactors. The plates have the core constituted by dispersions of metallic uranium in aluminum. The main topics of this work are: 1) The preparation of uranium powder with particle sizes in the 53-105μm diameter range; 2) The mixture and cold-pressing of uranium and aluminum powders for different uranium concentrations; 3) The behavior of the dispersions in the roll milling conditions; 4) Blister, radiographic, metallographic and irradiation tests for quality control of the plates. The irradiation test was performed in the IEA-R1 swimming-pool reactor using a prototype with a dispersion of aluminum and natural uranium (45 w/o ), reaching an integrated neutron flux of 8.663 X 10 18 n/cm 2 , no visual changes being noticed after the completion of the experiment. The behavior of the uranium-aluminum reaction for dispersions with 45% w/o uranium also studied. X-ray diffraction experiments showed the formation of UAl 2 UAl 3 and UAl 4 , while energy dispersive analysis of X-rays(EDAX) demonstrated that the diffusion of aluminum in uranium is the mechanism responsible for that reaction. The activation energy for the U-Al reaction was determined by dilatometric experiments yielding 20.2 kcal/mol.The aluminum-uranium reaction reaches an end when extended to 96 h at 600 deg C, namely, when all the uranium is found in the UAl 4 composition. (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.

Highly dispersed TaOx nanoparticles prepared by electrodeposition as oxygen reduction electrocatalysts for polymer electrolyte fuel cells

KAUST Repository

Seo, Jeongsuk

2013-06-06

Based on the chemical stability of group IV and V elements in acidic solutions, TaOx nanoparticles prepared by electrodeposition in an ethanol-based Ta plating bath at room temperature were investigated as novel nonplatinum electrocatalysts for the oxygen reduction reaction (ORR) in polymer electrolyte fuel cells (PEFCs). Electrodeposition conditions of Ta complexes and subsequent various heat treatments for the deposited TaOx were examined for the best performance of the ORR. TaOx particles on carbon black (CB), electrodeposited at a constant potential of -0.5 V Ag/AgCl for 10 s and then heat-treated by pure H2 flow at 523 K for 1 h, showed excellent catalytic activity with an onset potential of 0.93 VRHE (for 2 μA cm-2) for the ORR. Surface characterizations of the catalysts were performed by scanning transmission electron microscopy (STEM), transmission electron microscopy (TEM), and energy dispersive X-ray spectroscopy (EDS). The loading amounts of the electrodeposited material on the CB were determined by inductively coupled plasma atomic emission spectroscopy (ICP-AES). All the physical results suggested that high dispersion of TaOx particles on the CB surface with 2-3 nm size was critical and key for high activity. The chemical identity and modified surface structure for the deposited TaOx catalysts before and after H 2 heat treatment were analyzed by X-ray photoelectron spectroscopy (XPS). The formation of more exposed active sites on the electrode surface and enhanced electroconductivity of the tantalum oxide promoted from the H 2 treatment greatly improved the ORR performance of the electrodeposited TaOx nanoparticles on CB. Finally, the highly retained ORR activity after an accelerated durability test in an acidic solution confirmed and proved the chemical stability of the oxide nanoparticles. The high utilization of the electrodeposited TaOx nanoparticles uniformly dispersed on CB for the ORR was comparable to that of commercial Pt/CB catalysts

Solid oxide fuel cells and hydrogen production

International Nuclear Information System (INIS)

Dogan, F.

2009-01-01

'Full text': A single-chamber solid oxide fuel cell (SC-SOFC), operating in a mixture of fuel and oxidant gases, provides several advantages over the conventional SOFC such as simplified cell structure (no sealing required). SC-SOFC allows using a variety of fuels without carbon deposition by selecting appropriate electrode materials and cell operating conditions. The operating conditions of single chamber SOFC was studied using hydrocarbon-air gas mixtures for a cell composed of NiO-YSZ / YSZ / LSCF-Ag. The cell performance and catalytic activity of the anode was measured at various gas flow rates. The results showed that the open-circuit voltage and the power density increased as the gas flow rate increased. Relatively high power densities up to 660 mW/cm 2 were obtained in a SC-SOFC using porous YSZ electrolytes instead of dense electrolytes required for operation of a double chamber SOFC. In addition to propane- or methane-air mixtures as a fuel source, the cells were also tested in a double chamber configuration using hydrogen-air mixtures by controlling the hydrogen/air ratio at the cathode and the anode. Simulation of single chamber conditions in double chamber configurations allows distinguishing and better understanding of the electrode reactions in the presence of mixed gases. Recent research efforts; the effect of hydrogen-air mixtures as a fuel source on the performance of anode and cathode materials in single-chamber and double-chamber SOFC configurations,will be presented. The presentation will address a review on hydrogen production by utilizing of reversible SOFC systems. (author)

Evaluation of oxide dispersion strengthened (ODS) molybdenum and molybdenum-rhenium alloys

International Nuclear Information System (INIS)

Mueller, A.J.; Bianco, R.; Buckman, R.W. Jr.

1999-01-01

Oxide dispersion strengthened (ODS) molybdenum alloys being developed for high temperature applications possess excellent high temperature strength and creep resistance. In addition they exhibit a ductile-to-brittle transition temperature (DBIT) in the worked and stress-relieved condition under longitudinal tensile load well below room temperature. However, in the recrystallized condition, the DBTT maybe near or above room temperature, depending on the volume fraction of oxide dispersion and the amount of prior work. Dilute rhenium additions (7 and 14 wt.%) to ODS molybdenum were evaluated to determine their effect on low temperature ductility. The addition of 7 wt.% rhenium to the ODS molybdenum did not significantly enhance the mechanical properties. However, the addition of 14 wt.% rhenium to the ODS molybdenum resulted in a DBTT well below room temperature in both the stress-relieved and recrystallized condition. Additionally, the tensile strength of ODS Mo-14Re is greater than the base ODS molybdenum at 1,000 to 1,250 C

Storage of LWR spent fuel in air: Volume 1: Design and operation of a spent fuel oxidation test facility

International Nuclear Information System (INIS)

Thornhill, C.K.; Campbell, T.K.; Thornhill, R.E.

1988-12-01

This report describes the design and operation and technical accomplishments of a spent-fuel oxidation test facility at the Pacific Northwest Laboratory. The objective of the experiments conducted in this facility was to develop a data base for determining spent-fuel dry storage temperature limits by characterizing the oxidation behavior of light-water reactor (LWR) spent fuels in air. These data are needed to support licensing of dry storage in air as an alternative to spent-fuel storage in water pools. They are to be used to develop and validate predictive models of spent-fuel behavior during dry air storage in an Independent Spent Fuel Storage Installation (ISFSI). The present licensed alternative to pool storage of spent fuel is dry storage in an inert gas environment, which is called inerted dry storage (IDS). Licensed air storage, however, would not require monitoring for maintenance of an inert-gas environment (which IDS requires) but does require the development of allowable temperature limits below which UO 2 oxidation in breached fuel rods would not become a problem. Scoping tests at PNL with nonirradiated UO 2 pellets and spent-fuel fragment specimens identified the need for a statistically designed test matrix with test temperatures bounding anticipated maximum acceptable air-storage temperatures. This facility was designed and operated to satisfy that need. 7 refs

Ultrasound-assisted oxidative desulfurization process of liquid fuel by phosphotungstic acid encapsulated in a interpenetrating amine-functionalized Zn(II)-based MOF as catalyst.

Science.gov (United States)

Afzalinia, Ahmad; Mirzaie, Abbas; Nikseresht, Ahmad; Musabeygi, Tahereh

2017-01-01

In this work, ultrasound-assisted oxidative desulfurization (UAOD) of liquid fuels performed with a novel heterogeneous highly dispersed Keggin-type phosphotungstic acid (H 3 PW 12 O 40 , PTA) catalyst that encapsulated into an amino-functionalized MOF (TMU-17-NH 2 ). The prepared composite exhibits high catalytic activity and reusability in oxidative desulfurization of model fuel. Ultrasound-assisted oxidative desulfurization (UAOD) is a new way to performed oxidation reaction of sulfur-contain compounds rapidly, economically, environment-friendly and safely, under mild conditions. Ultrasound waves can be apply as an efficient tool to decrease the reaction time and improves oxidative desulfurization system performance. PTA@TMU-17-NH 2 could be completely performed desulfurization of the model oil by 20mg of catalyst, O/S molar ratio of 1:1 in presence of MeCN as extraction solvent. The obtained results indicated that the conversions of DBT to DBTO 2 achieve 98% after 15min in ambient temperature. In this work, we prepared TMU-17-NH 2 and PTA/TMU-17-NH 2 composite by ultrasound irradiation for first time and employed in UAOD process. Prepared catalyst exhibit an excellent reusability without PTA leaching and loss of activity. Copyright © 2016 The Authors. Published by Elsevier B.V. All rights reserved.

Fission product release by fuel oxidation after water ingress

International Nuclear Information System (INIS)

Schreiber.

1990-01-01

On the basis of data obtained by a literature search, a computer code has been established for the calculation of the degree of oxidation of the fuel in the damaged fuel particles, and hence of the fission product release as a function of the time period of steam ingress. (orig.) [de

MARMOT update for oxide fuel modeling

Energy Technology Data Exchange (ETDEWEB)

Zhang, Yongfeng [Idaho National Lab. (INL), Idaho Falls, ID (United States); Schwen, Daniel [Idaho National Lab. (INL), Idaho Falls, ID (United States); Chakraborty, Pritam [Idaho National Lab. (INL), Idaho Falls, ID (United States); Jiang, Chao [Idaho National Lab. (INL), Idaho Falls, ID (United States); Aagesen, Larry [Idaho National Lab. (INL), Idaho Falls, ID (United States); Ahmed, Karim [Idaho National Lab. (INL), Idaho Falls, ID (United States); Jiang, Wen [Idaho National Lab. (INL), Idaho Falls, ID (United States); Biner, Bulent [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bai, Xianming [Virginia Polytechnic Inst. and State Univ. (Virginia Tech), Blacksburg, VA (United States); Tonks, Michael [Pennsylvania State Univ., University Park, PA (United States); Millett, Paul [Univ. of Arkansas, Fayetteville, AR (United States)

2016-09-01

This report summarizes the lower-length-scale research and development progresses in FY16 at Idaho National Laboratory in developing mechanistic materials models for oxide fuels, in parallel to the development of the MARMOT code which will be summarized in a separate report. This effort is a critical component of the microstructure based fuel performance modeling approach, supported by the Fuels Product Line in the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. The progresses can be classified into three categories: 1) development of materials models to be used in engineering scale fuel performance modeling regarding the effect of lattice defects on thermal conductivity, 2) development of modeling capabilities for mesoscale fuel behaviors including stage-3 gas release, grain growth, high burn-up structure, fracture and creep, and 3) improved understanding in material science by calculating the anisotropic grain boundary energies in UO$_2$ and obtaining thermodynamic data for solid fission products. Many of these topics are still under active development. They are updated in the report with proper amount of details. For some topics, separate reports are generated in parallel and so stated in the text. The accomplishments have led to better understanding of fuel behaviors and enhance capability of the MOOSE-BISON-MARMOT toolkit.

Modelling of Zircaloy-steam-oxidation under severe fuel damage conditions

International Nuclear Information System (INIS)

Malang, S.; Neitzel, H.J.

1983-01-01

Small break loss-of-coolant accidents and special transients in an LWR, in combination with loss of required safety systems, may lead to an uncovered core for an extended period of time. As a consequence, the cladding temperature could rise up to the melting point due to the decay heat, resulting in severely damaged fuel rods. During heat-up the claddings oxidize due to oxygen uptake from the steam atmosphere in the core. The modeling and assessment of the Zircaloy-steam oxidation under such conditions is important, mainly for two reasons: The oxidation of the cladding influences the temperature transients due to the exothermic heat of reaction; the amount of liquified fuel depends on the oxide layer thickness and the oxygen content of the remaining Zircaloy metal when the melting point is reached. (author)

Prediction of U-Mo dispersion nuclear fuels with Al-Si alloy using artificial neural network